Applications of SYS TH codes to nuclear reactor design and accident analysis

Applications of SYS TH codes to nuclear reactor design and accident analysis

15

F. D’Auria, G.M. Galassi University of Pisa, Pisa, Italy

Abstract Envisaged Nuclear Power Plant transient performances are translated via experiments and expert judgement into lists of accidents. Accident scenarios are at the origin of phenomena. Phenomena are characterized by parameters or variables. The parameter values can be translated into requirements for predictive capabilities in thermal-hydraulics. The safety acceptance criteria consist of threshold parameter values. These are established by logical processes typically independent upon thermal-hydraulics knowledge. A global vision of nuclear reactor thermal-hydraulics is provided in the chapter from the side of phenomena and accident scenarios: 12 reactor types are considered for the characterization of 47 accident scenarios cross-linked with 113 phenomena and 15 sets of “homogeneous” variable time trends.

Acronyms ABWR AC ACC ADS AFW AM AMP ANSI AOO AP1000 APR1400 ATLAS ATWS BAF BC BD BD-SAT BD-SBC BIC BL

advanced boiling water reactor alternate current accumulator automatic depressurization system auxiliary feedwater accident management accident management procedure American National Standards Institute anticipated operational occurrence advanced pressurized water reactor (Westinghouse design) advanced pressurized water reactor (KEPCO design) integral simulator for PWR in Korea anticipated transient without scram bottom of active fuel bubble condenser (containment type) facility in Russia blowdown saturated blowdown subcooled blowdown boundary and initial conditions broken loop

Thermal Hydraulics in Water-Cooled Nuclear Reactors. http://dx.doi.org/10.1016/B978-0-08-100662-7.00015-4 © 2017 Elsevier Ltd. All rights reserved.

952

BOC BOL BOP BWR CANDU CCF CCFL CCVM CET CHF CL CLL CMT CONT CRE CRGT CSNI CVCS DBA DC DCC DEGB DEHL DG DiD DNB DWI ECCI ECCS EOC EOP EPR ERVCS ESF FA FC FCB FCO FP FWLB GF HEBR HFP HL HOSG HPCS HT HTC HZP

Thermal Hydraulics in Water-Cooled Nuclear Reactors

beginning of cycle and bottom of core (elevation of FA hydraulic entrance) beginning of life balance of plant boiling water reactor Canadian Deuterium Uranium Reactor designed in Canada counter current flow counter current flow limitation CSNI code validation matrix core exit thermocouples critical heat flux cold leg collapsed liquid level core make-up tank(s) containment control rod ejection control rod guide tube Committee on the Safety of Nuclear Installations Chemical and Volume Control System design basis accident downcomer direct contact condensation double-ended guillotine break double-ended hot leg (break) diesel generator defense in depth departure from nucleate boiling density wave instability emergency core cooling injection emergency core cooling system end of cycle emergency operating procedures European Pressurized Water Reactor (AREVA design) External Reactor Vessel Cooling System engineered safety features fuel assembly fuel channel fuel channel blockage forced convection fission products or fission power feed water line break gravity feed header break hot full power hot leg horizontal steam generator high-pressure core spray heat transfer heat transfer coefficient hot zero power

Accident scenarios and phenomena in WCNR

I I&C IAC IAFB IBLOCA IC IHE ILOCA IRWST ISP ITF JP KWU LBLOCA LCC/SW LOBI LOCA LOFA LOFW LOOSP LP LPCI LPCS LPIS LS LSTF LWR MBLOCA MCP MI MOC MOX MPTR MSIV MSLB MSSV nc NC NCO NEA NPP NPSH OECD ONB OTSG PCCS PCEI PCT PD

imposed (related to TSE) instrumentation and control interim acceptance criteria inverted annular film boiling intermediate break LOCA, see also MBLOCA isolation condenser intermediate heat exchanger interfacing LOCA in containment reactor water storage tank international standard problem integral test facility jet pump KraftWerk Union (NPP designer) large break LOCA loss of component cooling service water integral facility for PWR in Italy loss of coolant accident loss of flow accident (also MCP-trip) loss of feedwater loss of on- and off-site power lower plenum low-pressure coolant injection low-pressure core spray low-pressure injection system loop seal integral facility for PWR in Japan light water reactor medium break LOCA, see also IBLOCA main coolant pump mass inventory middle of cycle Mixed Uranium Plutonium nuclear fuel multiple pressure tube rupture main steam isolation valve main steam line break main steam safety valves noncondensable natural circulation natural convection Nuclear Energy Agency Nuclear Power Plant Net Positive Suction Head Organization for Economic Cooperation and Development onset of nucleate boiling once through steam generator passive containment cooling system parallel channel effects and instabilities peak cladding temperature pressure drop

953

954

PDGD PH PHW PHWR PIE PIUS PKL PORV PRHR PRISE PRZ PS PSA PSB PSD PSP PT PTS PWR PWR-O PWR-V QF R RBMK RC RCS RE RF RHR RIA RL RPV RST RTA SAMP SANB SBLOCA SBO SETF SG SGTR SH-D SIS SIT SMR SNB SRV SS SSC

Thermal Hydraulics in Water-Cooled Nuclear Reactors

pressure drop at geometric discontinuity phenomenon phenomenological window pressurized heavy water reactor designed by KWU (G) postulated initiating event process intrinsic ultimate safety (reactor design concept) integral facility for PWR in Germany pilot operated relief valve passive residual heat removal primary to secondary leakage (in VVER) pressurizer primary side (of RCS) or primary system Probabilistic Safety Assessment integral facility for VVER in Russia power spectral density pressure suppression pool pressure tube pressurized thermal shock pressurized water reactor (equipped with UTSG) PWR equipped with OTSG PWR equipped with HOSG (VVER type) quench front resulting (related to TSE) boiling water cooled and graphite moderated reactor designed in Russia reactor cavity reactor coolant system (includes PS and SS in PWR) reflood refill residual heat removal reactivity initiated accident recirculation loop reactor pressure vessel rod surface temperature relevant thermal-hydraulic aspect severe accident management procedure saturated nucleate boiling small break LOCA station blackout (or loss of on-site power) separate effect test facility steam generator steam generator tube rupture shutdown (accident conditions) safety injection system safety injection tank (see also accumulator) small modular reactor subcooled nucleate boiling steam relief valves secondary side (or secondary system) structures, systems, and components

Accident scenarios and phenomena in WCNR

SYS TH TAF TMI-2 TOC TPCF TSE TT UCSP UH UP USAEC USNRC UTP UTSG VVER WDVB WWER

955

system thermal-hydraulics top of active fuel Three Mile Island (unit 2) top of core (top of core upper plate, above TAF) two-phase critical flow time sequence of events turbine trip upper core support plate (see also UTP) upper head upper plenum United States Atomic Energy Commission United States Nuclear Regulatory Commission upper tie plate (see also UCSP) U-tubes steam generator water-cooled water-moderated energy reactors designed in Russia wetwell-to-drywell vacuum breakers see VVER

Chapter foreword The connection between thermal-hydraulics and accident analysis constitutes the framework for this chapter. In order to connect the fundamental thermal-hydraulic concepts, noticeably the models and the equations, with the transient performance of Nuclear Power Plants (NPP), a virtual top-down or bottom-up approach can be pursued. Within the former approach envisaged NPP, accident performances, hereafter referred as accident scenarios, are translated via experiments and expert judgment into lists of accidents, then into phenomena applicable to each accident, then into parameters or variables characterizing physical quantities, and finally into requirements for predictive capabilities in thermal-hydraulics. The safety acceptance criteria, established by logical processes independent upon thermal-hydraulics knowledge, constitute a necessary element to achieve the goals of the approach. Within the latter approach competences developed in thermal-hydraulics, also derived from NPP design and safety technology, are applied to estimate the transient performances of reactors. In this case the knowledge of fundamental physics triggers the process for characterizing the thermal-hydraulic evolutions of two-phase fluids during accidents. The safety acceptance criteria need proper consideration in this approach, too. The first approach provides the roadmap for this chapter. However, an interaction with the second approach cannot be avoided. The process of connecting calculated variables and phenomena could be done using a couple of accident scenarios (i.e., not 47 scenarios as used in the text) with full calculation details. This does not include the description of scenarios and does not give an idea of the interest toward accident scenarios by the scientific community.

956

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Very many acronyms are used throughout the text (see also Section 15.5.1). In a number of cases, the acronym is not identified the first time it is used (this is mostly the case of tables and figures). Although an attempt is made to use acronyms recognized in the literature, the reader is recommended to become familiar with the acronyms listed at the end of the chapter. The present chapter constitutes the tip of the virtual pyramid of knowledge developed in nuclear thermal-hydraulics. Therefore all chapters of the book have some connection with the present one. This is specifically true for Chapter 6 where phenomena are characterized and for Chapter 11 where the features of system codes (i.e., needed to predict accident scenarios) are discussed. The bases for code application, noticeably V&V, scaling, and uncertainty discussed in Chapter 13 are also important here. However, uncertainty of predictions for the transient scenarios presented in this chapter is not under consideration.

15.1

Introduction

The key motivation for nuclear thermal-hydraulics is the design of nuclear reactors and the safety demonstration. Water cooled reactors are of concern in the book. Thermal-hydraulic design of nuclear reactors implies, or has implied, for thermal-hydraulics support (to other disciplines) in (a) selecting the components, e.g., core, pumps, steam generators, turbine, condenser, needed to transform the fission power into electrical power; (b) fixing the layout of the whole system, e.g., mutual position between core (heat source) and steam generator (heat sink), pressurizer location also to allow for spray line, pump location to maximize the Net Positive Suction Head (NPSH) from the loop, condenser location to optimize the thermal efficiency; (c) searching for the optimum configuration of individual components, e.g., the core constituted by thin fuel pins, the helicoidally shaped steam separators to facilitate steam liquid separation by inertia forces; (d) fixing the normal operational parameters, like working pressure, maximum and average linear heat flux for the fuel pins in the core, pressure drops, level in the pressurizer, in the down-comer of BWR ad of steam generators in PWR; (e) designing each component, e.g., the fuel pin, the number of tubes in steam generators, the size of hot and cold legs, the flow paths inside the pressure vessels, as well as the size and the number so headers (in pressure tubes reactor types like CANDU and RBMK); and (f ) determining expected loads, e.g., pressure and pressure drops including pressure wave propagation originated by opening/closure of valves, thermal stresses originated by cooling/heating rates and thermal fatigue.

Safety demonstration of nuclear reactors implies, among the other things: (g) the calculation of the transient performance of reactors; (h) namely, the proof of suitable design/construction for emergency core cooling systems (ECCS), e.g., the volume and the number of accumulators, and, more in general of engineered safety features (ESF), e.g., the pressure set point for the operation of steam relief valves (SRV), the closure time for main steam isolation valves (MSIV); (i) the design of emergency operating procedure (EOP) and the accident management procedure (AMP);

Accident scenarios and phenomena in WCNR

957

(j) an interaction with safety and licensing processes to fix requirements and procedures for the related analyses, e.g., list of design basis accidents (DBA), acceptable probability for failures and demonstration that consequences of failure are within the acceptability domain.

The above list of thermal-hydraulic applications, incomplete and/or not comprehensive, may be taken as another demonstration, or even one origin, for the universe in nuclear thermal-hydraulics introduced in the Chapter 1. The impossibility to describe all applications of nuclear thermal-hydraulics in one chapter (even though in one book) is also a consequence from the list. The complexity of some topics, namely those which require the analysis of accident performances of nuclear reactors, imposed the development of system thermal-hydraulic (SYS TH) codes as outlined, for instance, in Chapters 2 and 10. The scope for the present chapter is restricted to the transient analyses in Light Water Reactors (LWR), i.e., BWR and PWR including evolutionary and similar design like EPR, AP1000, ABWR, AP1400, and VVER, partly accomplishing the content of item (g), as listed previously; in a few cases transient analyses of CANDU, RBMK, and PHWR (Pressurized Heavy Water Reactor equipped with vertical channels core and pressure vessel) are presented. The scope can be further specified as “addressing the capabilities of the SYS TH codes (discussed in Chapter 11), constituting a cross-connection among accident scenarios (DBA area), and results of those codes and phenomena (discussed in Chapter 6, see also Section 15.1.1).” The consideration of Beyond DBA (BDBA) conditions is within the scope for the chapter till the situation where irreversible degradation of core or the loss of geometric integrity occurs. Therefore, proofs of qualification of the discussed SYS TH code calculated results and the evaluation of uncertainty of calculations are outside the scope for the present chapter.

15.1.1 The connection with phenomena Thermal-hydraulic phenomena expected to occur in LWR (and VVER) in DBA conditions have been classified in the documents OECD/NEA/CSNI (1993) and OECD/ NEA/CSNI (1996a), respectively, related to separate effect test facilities (SETF) and integral test facilities (ITF), see also OECD/NEA/CSNI (2001), related to VVER (as also discussed Chapter 2). Containment-related phenomena can be found in the report OECD/NEA/CSNI (1999). So-called computer code validation matrices (CCVM) have been created based on the correspondence between phenomena and experiments. Therefore, the concerned phenomena have been the object of experimental investigations and are considered in various reports, papers, or textbooks, see for instance, the report (USNRC, 1988), the CSNI documents dealing with International Standard Problems (ISP) (e.g., OECD/NEA/CSNI-ISP, 2000, and more recently OECD/ NEA/CSNI-ISP-50, 2012), the textbook (Levy, 1999), and the papers (like D’Auria and Galassi, 1990a; D’Auria and Ingegneri, 1997; Del Nevo et al., 2012a,b; Mascari et al., 2012; Reventos et al., 2012). The idea here is to create a correspondence between thermal-hydraulic phenomena, variables to characterize the phenomena, and results of code calculations related to accident analyses.

958

Thermal Hydraulics in Water-Cooled Nuclear Reactors

15.1.2 Objective and structure The objective for this chapter is to close a logical circle, the starting point of the circle being the definitions and the fundamental concepts given in Chapter 3. These are followed by needs in nuclear reactor design and safety (Chapter 4), expected phenomena for DBA in LWR and advanced water cooled designs (Chapter 6), equations and models (Chapters 5, 7–10), and by computational tools which make use of the equations and models (e.g., Chapters 11 and 12 where applicable). In this connection Chapters 1 and 2 provide a general framework suitable for the understanding of the nuclear thermal-hydraulics discipline; the remaining two chapters discuss the conditions for the applicability of those codes to the accident analysis, noticeably, the needs for qualification and the need to supplement the calculation results by uncertainty (Chapter 13), and the connection with the licensing process (Chapter 14). Additional objective is to provide essential information to the reader in relation to the links among phenomena, the predicted accident evolutions in nuclear reactors and the variables, or time trends or parameters, which are the direct outcomes of SYS TH code calculations and are suitable for the characterization of phenomena. Although a complex process is pursued to establish those links, it seems worthwhile to emphasize that only a narrow sector of system thermal-hydraulics is involved in this chapter. This is the item (g) from the list above (i.e., safety analysis), supplemented by one noticeable example from the item (d) (i.e., design analysis) (e.g., see Section 15.4.12.1). Four main tables (plus one) have been created as bases for the document: Table 15.1 deals with a minimum-reasonable list of accidents or transient scenarios resulting from NPP calculations performed by system thermal-hydraulic codes.

l

List of NPP accident scenarios calculated by system thermal-hydraulics codes

Table 15.1

Section No 1 2

Identification

1 15.4.10

3

2

Type

Reactor

3

4

ATWS-1 ATWS-2

BDBA BDBA

ABWR BWR

ATWS-3

BDBA

PWR

DBA

PWR

4

15.4.12.8

BORON-D

5

15.4.12.1

CHF

6 7

15.4.12.2

CONT-1 CONT-2

N/A

AP1000 ABWR

Notes and references 5 Ferng et al. (2010) Munoz-Cobo et al. (2014) Kliem et al. (2009) and Chen et al. (2014) Jimenez et al. (2015) and Graffard and Goux (2006) Fundamental study for core design, Visentini et al. (2014) Hung et al. (2015) Chen and Yuann (2015)

Accident scenarios and phenomena in WCNR

Table 15.1

Continued

Section No

959

Identification

1

2

Type 3

Reactor 4

8

15.4.9

CRE

DBA

PWR

9 10

15.4.12.6 15.4.8

FCB FWLB

DBA

RBMK APR1400

11

15.4.2 and 15.4.12.12

HEBR-1

DBA

CANDU

12

15.4.2 and 15.4.11 15.4.2

HEBR + AMP

BDBA

CANDU

HEBR-2 LBLOCA-1 LBLOCA-2

DBA DBA DBA

RBMK AP1000 BWR

16 17

LBLOCA-3 LBLOCA-4

DBA

EPR PHWR

18

LBLOCA-5

DBA

PWR

13 14 15

19

15.4.12.9

LCC/SW

BDBA

PWR

20 21

15.4.4

LOFA LOFW

AOO DBA

EPR PWR

22

15.4.6

LOOSP

BDBA

VVER, BWR, CANDU

23

15.4.2

MBLOCA

DBA

PWR

Notes and references 5 Part of the RIA class, Todorova and Ivanov (2003) D’Auria et al. (2005) Actuation of passive system, Bae et al. (2014) Positive reactivity insertion and fuel behavior, Rouben (1997) and Horhoianu et al. (1998) Mehedinteanu (2009) D’Auria et al. (2005) Queral et al. (2015) Recirculation line break, Beckmeyer et al. (1979) Arkoma et al. (2015) Positive reactivity insertion, D’Auria et al. (2008a) Double-Ended Guillotine Break (DEGB), OECD/ NEA/CSNI (2009), and D’Auria and Galassi (2001) Connection with PSA, OECD/NEA/CSNI (2011) AREVA (2014) See also LOFA, SBO, and LOOSP, OECD/ NEA/CSNI (2011) Borisov et al. (2013), see Fukushima events in Chapter 16, Tong et al. (2014) and D’Auria et al. (2006) Also identified as IBLOCA, OECD/ NEA/CSNI (2011) Continued

960

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Table 15.1

Continued

Section No

Identification

1

2

24

15.4.12.9

MCP-TRIP

25 26 27

15.4.12.7 15.4.8

MPTR MSLB-1 MSLB-2

28

15.4.12.3

MSLB-PTS

29

15.4.12.5

MSLB + SGTR*

30

15.4.12.4

NC

31

15.4.7

PRISE + AMP

–

15.4.9

RIA

32

15.4.5

Type 3

Reactor 4

AOO

VVER

SBO-1

DBA

APR1400

SBO-2 SBO + AMP

DBA BDBA

PWR VVER

SBLOCA-1 SBLOCA-2

DBA DBA

AP1000 APR1400

37

SBLOCA-3

(B) DBA

PWR-O, EPR

38

SBLOCA-4

DBA

BWR

33 34 35 36

15.4.5 and 15.4.11 15.4.3

39

15.4.12.3

SBLOCA-PTS

40

15.4.3 and 15.4.11 15.4.7

SBLOCA + AMP SGTR

41

Notes and references 5

Groudev and Stefanova (2006), see also LOFA DBA RBMK D’Auria et al. (2005) DBA PWR Jeong et al. (2006) DBA PWR-O In PWR equipped with OTSG, D’Auria et al. (2003a) PWR PTS focus, Jang (2007) BDBA PWR * and multiple SGTR USNRC (1998), Seul et al. (2003), and Jimenez et al. (2013) DBA PWR Flow map including PWR-O, PWR-V, CANDU, etc., and experiments, IAEA (2005a) and Cherubini et al. (2008a) BDBA VVER IAEA (2009b) and D’Auria et al. (2006) See CRE. Qualitative overview of RIA in Section 15.4.9

PWR BDBA

VVER

DBA

PWR

Supported by experiments, Yu et al. (2013) Bittan (2015) D’Auria et al. (2006) Yang et al. (2012a,b) Focusing on loop seal clearing, Kim and Choi (2014) Bandini and De Rosa (2014), Freixa and Manera (2011) Analytis and Coddington (2002) PTS focus, Jang (2007) D’Auria et al. (2006) OECD/NEA/CSNI (1988)

Accident scenarios and phenomena in WCNR

Table 15.1

Continued

Section No

l

Identification

1

2

Type 3

Reactor 4

42

15.4.12.3

SGTR-PTS

43

15.4.12.11

SH-D

DBA

PWR

44

15.4.12.10

TT-1

AOO

BWR

TT-2

AOO

SMR-PWR

45

l

961

PWR

Notes and references 5 PTS focus, Jang (2007) Haste et al. (2010) and Son and Shin (2007) Bousbia-Salah and D’Auria (2002) Haratyk and Gourmel (2015)

Table 15.2 deals with the full list of transient thermal-hydraulic phenomena derived from existing OECD/NEA/CSNI and IAEA documents. Tables 15.3 and 15.4 (the last one supported by Table 15.5) deal with the links phenomena to accident scenarios and phenomena to parameters.

Tables 15.1 and 15.2 are embedded into Section 15.2 which also constitutes the link with Chapter 6 and with the licensing/safety processes for NPP (partly discussed in Chapter 14). Tables 15.3–15.5 constitute the key content of Section 15.3: variables (or time trends) are used to characterize transient scenarios (e.g., listed in Table 15.1) and phenomena (e.g., listed in Table 15.2). The information gathered in Tables 15.1–15.5 is used in Section 15.4 of the present chapter for characterizing the accident scenarios and for connecting those accident scenarios with phenomena and parameters.

15.2

Accident scenarios and phenomena

Events, or accidents, or transients, or, better, postulated initiating events (PIE) are mandatory elements to address the safety requirements for NPP issued by Regulatory Authorities. The acceptance criteria for the design of the ECCS (see e.g., USAEC, 1971, already discussed in Chapter 2) constitute a suitable example: the fulfillment of those criteria must be based upon expected transients for the concerned NPP unit. Thus, PIE are introduced. At their origin PIE have a weak link or no link at all with thermal-hydraulics. However, it comes out that PIE unavoidably cause single- or two-phase coolant evolutions, which determine the NPP performance and allow the definition of safety margins. Then, PIE are characterized by phenomena: this establishes the tight link between safety requirements, PIE, and nuclear thermal-hydraulics. Insights into the concept of PIE and the origins of PIE are discussed in Sections 15.2.1–15.2.5, primarily derived from IAEA (2002a, 2003). Furthermore a list of events suitable for the purpose of the present chapter (i.e., showing the link between PIE and nuclear thermal-hydraulics) is provided in Table 15.1, Section 15.2.3. The list of phenomena and the connection between each phenomenon and at least one NPP type can be derived from Table 15.2 in Section 15.2.5.

962

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Table 15.2 ID S-1-ACC I-1-ASY-L I-2-ASY-D A-1-CHV A-2-CC A-3-CMT A-4-DL A-5-PC A-6-POO I-3-BD I-4-NCBC S-2-BO S-3-CCF1 S-4-CCF2 S-5-CCF3 S-6-CCF4 S-7-CCF5 S-8-CCF6 I-5-BC I-6-CLDO B-1-COH B-2-COP S-9-CO1 S-10-CO2 S-11-CO3 S-12-CO4 I-7-COTH S-13-CRGT A-7-CSC

S-14-ECCB S-15-ED1 S-16-ED2 S-17-ED3 S-18-ED4 S-19-ED5 S-20-ED6 B-3-EV1 B-4-EV2 I-8-FO S-21-GM1 S-22-GM2 S-23-GM3 S-24-GM4 A-8-GDR S-25-HOHT S-26-HT1 S-27-HT2 S-28-HT3 A-9-HTCO S-29-IMPU B-5-IF1 B-6-IF2 I-9-INC S-30-IPU S-31-JPU S-32-LA

A-10-LTS S-33-LVM1 S-34-LVM2 S-35-LVM3 S-36-LVM4 S-37-LVM5 S-38-LVM6 S-39-LSC S-40-LPE S-41-LPF

List of phenomena Phenomena ID

Accumulator behavior Asymmetric loop behavior Asymmetry due to the presence of a dam Behavior of check valves Behavior of containment emergency systems (e.g., PCCS) Behavior of core make-up tanks Behavior of density locks Behavior of emergency heat exchangers including PRHR and IC Behavior of large pools of liquid Blowdown Boiler condenser mode (of NC) Boil-off Boron mixing and transport (also A-12-BO) CCF/CCFL-channel inlet orifice CCF/CCFL-downcomer CCF/CCFL-HL and CL CCF/CCFL-SG tubes CCF/CCFL-surgeline CCF/CCFL-UTP Centrifugal pump Channel and bypass axial flow and void distribution Collapsed level behavior in downcomer Condensation due to heat removal Condensation due to pressurization Condensation in stratified conditions-horizontal pipes Condensation in stratified conditions-PRZ Condensation in stratified conditions-SG-PS Condensation in stratified conditions-SG-SS & BWR-PSP Containment pressure and temperature Core thermal-hydraulics Core wide void and flow distribution CRGT flashing Critical and supercritical flow in discharge pipes Critical flow Critical power ratio De-entrainment Depressurization ECC bypass/downcomer penetration ECC mixing and condensation Entrainment/de-entrainment-core Entrainment/de-entrainment-downcomer Entrainment/de-entrainment-hot leg with ECCI Entrainment/de-entrainment-SG mixing chamber Entrainment/de-entrainment-SG tubes Entrainment/de-entrainment-UP Evaporation due to depressurization (including at geometric discontinuities*) Evaporation due to heat input Flow through openings Global multi-D fluid temperature, void and flow distribution-core Global multi-D fluid temperature, void and flow distribution-downcomer Global multi-D fluid temperature, void and flow distribution-SG SS Global multi-D fluid temperature, void and flow distribution-UP Gravity driven reflood Horizontal heated channel HT [added PH] HT [NCO, FCO, SNB, SANB, CHF/DNB, post-CHF]-core, SG, structures HT [radiation]-core HT [condensation]-SG structures HT condensation in containment structures, with or w/o non-condensable Impeller pump behavior Instability (in boiling channels) Interfacial friction in horizontal flow Interfacial friction in vertical flow Intermittent 2-phase NC Internal pump behavior (specific geometry) [added PH] Jet pump behavior Liquid accumulation in horizontal SG tubes Liquid carry-over

Liquid temperature stratification Liquid-vapor mixing with condensation-core Liquid-vapor mixing with condensation-downcomer* Liquid-vapor mixing with condensation-ECCI in HL and CL* Liquid-vapor mixing with condensation-lower plenum* Liquid-vapor mixing with condensation-SG mixing chamber Liquid-vapor mixing with condensation-UP Loop seal filling and clearance (or clearing) LP entrainment LP flashing

Type SETF ITF ITF

Reactor

Details and Notes Mainly

PWR Shutdown conditions

New reactors

ITF/SETF/Basic PWR-O ITF/SETF/Basic SETF PWR BWR

Also containment

PHW rather than PH

ITF

SETF

PWR

ITF ITF Basic Basic

BWR BWR

See B-4-EV2 Also ITF

N/A See Impeller pump See also phase separation

N/A PWR/BWR

SETF

PWR BWR

ITF ITF ITF SETF

BWR BWR New reactors

ITF

BWR

SETF ITF

PWR

Also BWR wet-well See I-18-PRB and S-42-NCOC See global multi-D See global multi-D Also CANDU and RBMK See TPCF See HT CHF See entrainment See blowdown See liquid vapor mixing

All SETF

Basic Basic ITF

PWR

N/A

* Reversible part Shutdown conditions

PWR/BWR SETF PWR SETF SETF

New reactors CANDU All

New reactors SETF All SETF/ITF Basic N/A Basic ITF SETF SETF ITF

PWR-O ABWR BWR PWR-V

Including HT below Including VVER conditions Also containment External pumps See S-44-PCEI

Also AP-1000

See entrainment and I-24-SBI

New reactors All SETF

PWR

Also containment Also ITF. *Including coldhot liquid mixing (3D effect)

PWR SETF SETF SETF

PWR PWR PWR/BWR

Also ITF See also blowdown

Accident scenarios and phenomena in WCNR

Table 15.2 ID I-10-NC1 I-11-NC2 I-12-NC3 I-13-NC4 I-14-NC5 I-15-NC6 A-11-NC S-42-NCOC S-43-NCG I-16-NTF1 S-44-PCEI S-45-PSB S-46-PS1 S-47-PS2 S-48-PS3 I-17-PFU B-7-PD B-8-PW I-18-PRB I-19-PRZ S-49-QF1 S-50-QF2 I-20-RF I-21-RE I-22-RCM S-51-SEP I-23-SIP S-52-SPR1 S-53-SPR2 S-54-SPR3 I-24-SBI S-55-SDR I-25-SLD S-56-STR A-12-BO I-26-SHH I-27-SULI I-28-SH I-29-NTF2 S-57-HOSG S-58-OTSG S-59-CF1 S-60-CF2 S-61-CF3 A-13-NCG

I-30-VCP B-9-FR

963

Continued Phenomena ID

Type

Mixture level and entrainment-core, downcomer and SG SS NC, 1-phase and 2-phase-PS & SS NC core and downcomer NC core bypass, hot and cold bundles NC core, gap, downcomer, dummy elements NC core, vent valves, downcomer NC with horizontal SG NC RPV and containment and various system configurations Natural convection and H2 distribution Non condensable gas effect including condensation HT in RCS Nuclear fuel behavior Nuclear thermal-hydraulics feedback and spatial effect (see also I-29-NTF2) Nuclear thermal-hydraulics instabilities Parallel channel effects and instabilities PCEI Phase separation at branches (including effect on TPCF) Phase separation/vertical flow with and w/o mixture level-core Phase separation/vertical flow with and w/o mixture level-downcomer Phase separation/vertical flow with and w/o mixture level-pipes & plena Pool formation in UP Pressure drops at geometric discontinuities, including containment Pressure wave propagation Pressure-temperature increase and boiling due to energy and mass input

ITF

PRZ thermal-hydraulics QF propagation/rewet-fuel rods QF propagation/rewet-channel walls, water rods Refill including loop refill in PWR-O Reflood Reflux condenser mode and CCFL Return to nucleate boiling (RNB) Separator behavior (and* flooding, steam penetration, liquid carry-over) SG siphon draining (SG interaction with ESF, including gravity driven) Spray effects-core (including cooling and distribution) Spray effects-OTSG SS Spray effects-PRZ Steam binding (liquid carry-over, etc.) Steam dryer behavior Steam line dynamics Stratification in horizontal flow-pipes (in 1-phase and 2-phase conditions) Stratification of boron Structural heat and heat losses Surgeline hydraulics Superheating in OTSG SS Thermal-hydraulics — nuclear fuel feedback (see also I-16-NTF1) Thermal-hydraulics of horizontal SG, PS and SS Thermal-hydraulics of OTSG, PS and SS TPCF-breaks TPCF-pipes TPCF-valves Tracking of noncondensable gases Valve leak flow (connected with construction, operation, maintenance) Vapor (or steam) carry-under Vapor pull through Void collapse and temperature distribution during pressurization Wall to fluid friction Water accumulation in horizontal SG tubes

Reactor PWR/BWR BWR

ITF

BWR, CANDU*

PWR-V PWR-O PWR-V New reactors SETF SETF PWR SETF/ITF ITF BWR ITF BWR SETF BWR SETF SETF

All

Details and Notes See phase separation SS only for PWR *Also RBMK

Also containment Inside containment Also ITF See I-29-NTF2 Also RBMK, ABWR, etc. See I-16 and S-44 Also ITF (T-branches) Also ITF

ITF Basic Basic

PWR

See also S-8-CCF6

N/A

Also new reactors

ITF

PW/BWR

Containment and shutdown

ITF

PWR PWR/BWR BWR

SETF ITF

PWR/BWR

ITF

PWR

SETF ITF

PWR/BWR PWR BWR SETF PWR-O PWR ITF PWR SETF PWR/BWR ITF BWR SETF PWR New reactors ITF All ITF PWR ITF PWR-O ITF PWR SETF PWR-V SETF PWR-O SETF

PHW rather than PH See reflood and QF *Mainly for BWR Shutdown conditions

Mainly BWR Also ITF See S-2-BO Scaling issue

Also CANDU, PHWR, etc. See spray effect OTSG

All

New reactors ITF ITF SETF ITF BWR Basic N/A ITF PWR-V

See S-43 NCG and containment

See also TPCF Valves See S-51-SEP & I-10-NC1 See S-45-PSB Also basic condensation See liquid accumulation

15.2.1 Establishing PIE The basic objective for nuclear safety, as already discussed in Chapter 2, is the protection of individuals, the society, and the environment from radiations. In order to achieve the objective, the following fundamental safety functions have to be performed (not an exhaustive list): (a) control of reactivity; (b) removal of residual heat from the fuel; and (c) confinement of radioactive materials.

964

Table 15.3

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Cross-link between phenomena and accident scenarios

Table 15.3

Continued

Continued

966

Table 15.3

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Continued

Control of reactivity generally means all the measures taken to avoid inadvertent nuclear criticality, loss of reactivity control, inadvertent power excursions, or reduction in shutdown margin. Removal of heat from the nuclear fuel necessitates that adequate cooling of the fuel be ensured. Confinement of the radioactive materials implies the introduction of suitable barriers and the demonstration that those barriers remain effective under selected circumstances, including the consequences of PIE. The knowledge and the understanding of the framework for the defense in depth (DiD) (e.g., IAEA, 1996, 2005a) including the concepts of prevention and mitigation are essential for a robust and consistent description of the current status of nuclear safety and for industrial applications. This is well beyond the purpose for this section; rather snap-shot information is provided hereafter in relation to the process to establish PIE (interested reader may refer to the listed references). According to INSAG-10 (IAEA, 1996), DiD consists of a hierarchical deployment of different levels of equipment and procedures in order to maintain the effectiveness of physical barriers placed between radioactive material and workers, the public, or the environment, during normal and off-normal operation. To this aim several successive physical barriers (generally, five levels are recognized) for the confinement of radioactive material are in place in any NPP: should one level fail, the subsequent level comes into play. Therefore one may state, IAEA, 2002a: “Once a release of radioactive material is foreseen, either as a routine part of normal operation or as the consequence of an

Accident scenarios and phenomena in WCNR

SBLOCA-3 LBLOCA-1 HEBR-1 CHF FCB SBLOCA-3 SH-D LOFA LOOSP (in CANDU) MSLB-1 NC (PWR-O analysis) ATWS-1a SBLOCA-4 a SBO+AMP SGTR MBLOCA SBLOCA-3

60

S-35-LVM3

SBLOCA-PTS

61 62 63 64 65 66 67 68 69 70 71 72 73 74 75

S-36-LVM4 S-37-LVM5 S-38-LVM6 S-39-LSC S-40-LPE S-41-LPF I-10-NC1 I-11-NC2 I-12-NC3 I-13-NC4 I-14-NC5 I-15-NC6 A-11-NC S-42-NCOC S-43-NCG

LBLOCA-2 SGTRa; LBLOCA-4 SBLOCA-4 SBLOCA-2 LBLOCA-2 LBLOCA-1 NC SBLOCA-4 HEBR+AMP; FCB SBO+AMP a REFLOOD-APR1400 PRISE+AMP SBLOCA-1 CONT-1 a CONT-1

5 6 7 Parameters vs

AP

AP A

13

14

15

AP A P A

A

A

P A

12

Passive System Performance

Coolant Mass in RCS

Coolant Temp (Core. HL, LP, etc.)

8 9 10 11 - Accident Scenario (A) - Phenomena (P) A A P P

P A A

A

RST and Nuclear Fuel Parameter

Core Thermal Power Heat Fl DNBR

SG Level (PS and SS)

4

Flowrate (Core Inlet, Break, etc.)

3

PRZ Level

2

ESF Performance (Excluded Passive Systems)

S-24-GM4 A-8-GDR S-25-HOHT S-26-HT1 S-27-HT2 S-28-HT3 A-9-HTCO S-29-IMPU B-5-IF1 B-6-IF2 I-9-INC S-30-IPU S-31-JPU S-32-LA A-10-LTS S-33-LVM1 S-34-LVM2

1

A A P

Containment Pressure and Temperature

43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59

Section of Chapter vs Phenomenon (and Notes) 15.4.2 15.4.8 15.4.8 15.4.12.2 15.4.12.2 15.4.2 15.4.12.2 15.4.3 15.4.3 15.4.2 15.4.12.4 15.4.12.8 15.4.3 15.4.3 15.4.3 15.4.3 15.4.3; see also 15.4.11 15.4.3 15.4.12.4 15.4.10 15.4.12.9 15.4.12.9; see also 15.4.12.12 15.4.12.3 15.4.12.3 15.4.12.3 15.4.12.10 15.4.2 15.4.2 15.4.3 15.4.2 15.4.2 15.4.2 15.4.2 15.4.3 15.4.3 15.4.2 15.4.2; see any LOCA scenario 15.4.9; any boil-off scenario 15.4.7 15.4.9; see also I-29-NTF2 15.4.12.3 15.4.8; supported by experiment 15.4.2 15.4.2; contributed by CMT 15.4.2; support to 15.4.12.12 15.4.12.1; see any scenario 15.4.12.6; see also S-26-HT1 15.4.11 15.4.12.11 15.4.12.9 15.4.12.4; see any scenario 15.4.12.4; see any scenario 15.4.12.4 15.4.10 15.4.2 15.4.5; see also 15.4.11 15.4.3 (SG SS) 15.4.2 15.4.12.2; see also 15.4.2 and 15.4.3 15.4.12.3; see also 15.4.2 and 15.4.3 15.4.2 15.4.7; see also 15.4.3 15.4.2 15.4.3 15.4.2 15.4.2; all LBLOCA 15.4.12.4 15.4.12.4 15.4.11; see also 15.4.12.4 15.4.12.4 15.4.12.4 15.4.7; see also 15.4.12.4 15.4.3 15.4.12.2 15.4.12.4

Local Fluid Velocity, Flow, Void, Load and DP

Accident Scenario vs Phenomenon LBLOCA-5 LBLOCA-1 MSLB-1 SH-D a SBLOCA-1 CONT-1 LBLOCA-1 a LOFW-PIUS SBLOCA-1 ATWS-1 LBLOCA-1 NC BORON-D SBLOCA-4 a LBLOCA-5 LBLOCA-4 NC SBLOCA+AMP; SBO-2 SBLOCA-1 a SBLOCA-4 ATWS-1; SBLOCA-3 LCC/SW MCP-TRIP SBLOCA-3 SBO-1 SGTR-PTS; SBO-1 TT-2 LBLOCA-2 SBLOCA-4a SBLOCA-1 LBLOCA-1 LBLOCA-1 LBLOCA-1 LBLOCA-4 a LBLOCA-4 SBLOCA-3 SBLOCA-1 HEBR-2 SH-D CONT-1 CRE MSLB-PTS FWLB

Phenomenon

SG Pressure

S-1-ACC I-1-ASY-L I-2-ASY-D A-1-CHV A-2-CC A-3-CMT A-4-DL A-5-PC A-6-POO I-3-BD I-4-NCBC S-2-BO S-3-CCF1 S-4-CCF2 S-5-CCF3 S-6-CCF4 S-7-CCF5 S-8-CCF6 I-5-BC I-6-CLDO B-1-COH B-2-COP S-9-CO1 S-10-CO2 S-11-CO3 S-12-CO4 I-7-COTH S-13-CRGT A-7-CSC S-14-ECCB S-15-ED1 S-16-ED2 S-17-ED3 S-18-ED4 S-19-ED5 S-20-ED6 B-3-EV1 B-4-EV2 I-8-FO S-21-GM1 S-22-GM2 S-23-GM3

No

Core RPV Level

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42

Reactivity and Boron Concentration

Phenomena and variables calculated in accident scenarios

RPV PRZ Pressure

Table 15.4

967

AP

P

A

AP P

P

P P AP

AP

A

AP

AP

A P P

AP

P

A

A P P

A

A

A

P P A A A

PA

A A

AP

AP

AP

A P

P

P A A

P

A

AP

A P

P

P

P P P P P AP

A

A

A

P P

A P

P P P P

P P AP

AP AP

AP AP

AP

A P

A

A P

A

A

AP

P A P AP

P AP

P A A

A

A

A AP P P P P P

A

A A

A

AP

A

A

P

A

P P A

A

A

P

A

AP

P AP

AP P

A

A

A

A

A

AP P P

A

P

A P

AP

AP P A

P P P

A

A

AP P

A

A

P A

A

P P

P

P

P

P

Continued

968

Thermal Hydraulics in Water-Cooled Nuclear Reactors

No 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 107 108 109 110 111 112 113 a

Phenomenon I-16-NTF1 S-44-PCEI S-45-PSB S-46-PS1 S-47-PS2 S-48-PS3 I-17-PFU B-7-PD B-8-PW I-18-PRB I-19-PRZ S-49-QF1 S-50-QF2 I-20-RF I-21-RE I-22-RCM S-51-SEP I-23-SIP S-52-SPR1 S-53-SPR2 S-54-SPR3 I-24-SBI S-55-SDR I-25-SLD S-56-STR A-12-BO I-26-SHH I-27-SULI I-28-SH I-29-NTF2 S-57-HOSG S-58-OTSG S-59-CF1 S-60-CF2 S-61-CF3 A-13-NCG I-30-VCP B-9-FR

P P A

A

A

AP P AP P

Containment Pressure and Temperature

ESF Performance (Excluded Passive Systems)

Passive System Performance

Coolant Mass in RCS

Coolant Temp (Core. HL, LP, etc.)

8 9 10 11 - Accident Scenario (A) - Phenomena (P) P P

Local Fluid Velocity, Flow, Void, Load and DP

5 6 7 Parameters vs

RST and Nuclear Fuel Parameter

4

Core Thermal Power Heat Fl DNBR

3

SG Level (PS and SS)

2

Flowrate (Core Inlet, Break, etc.)

1

PRZ Level

SG Pressure

Section of Chapter vs Phenomenon (and Notes) 15.4.10; supporting 15.4.12.2 15.4.9 15.4.3 15.4.3 15.4.10; see also 15.4.3 15.4.5 15.4.3 15.4.4; see any scenario 15.4.12.7 15.4.12.2 15.4.4; see also 15.4.12.12 15.4.2 15.4.2 15.4.2 15.4.2; supporting 15.4.12.12 15.4.12.4 15.4.12.5 15.4.12.11 15.4.3 15.4.8 15.4.4 15.4.2 15.4.10 15.4.12.10 15.4.12.3 (2-phase) 15.4.3 15.4.5; see any scenario 15.4.6; see SBO-2 & SGTR 15.4.8 15.4.12.12 15.4.7 15.4.8 15.4.2 15.4.7; see also 15.4.3 15.4.5 15.4.3 15.4.12.10 15.4.6; see any scenario

Core RPV Level

Accident Scenario vs Phenomenon ATWS-2; ATWS-3 NC SGTR SBLOCA-3 ATWS-1 SBO-2 SBLOCA-1 CONT-1 MPTR; TT-1 CONT-2 LOFW LBLOCA-1 LBLOCA-2 LBLOCA-2 LBLOCA-3 NC MSLB+SGTR SH-D SBLOCA-4 a MSLB-2 SGTR LBLOCA-1 TT-1 TT-1 NC BORON-D SBO-1 LOOSP a MSLB-2 ATWS-3; HEBR-1 PRISE+AMP MSLB-2 HEBR-2; FCB SBLOCA-3 SBO-1 CONT-1 SBO-1 LOOSP

Reactivity and Boron Concentration

Continued

RPV PRZ Pressure

Table 15.4

12

13

14

15

A P A

A A

A

P PA A

A

A P P

A

AP

AP

A

A

AP P

P P

A AP

A P P A AP P AP

AP

P P P P

P P

P P P

P A

AP P

P AP A A A

AP A PA

AP

AP A

A P

A

PA P P

AP A

A A P

P

A

A

A A A P P A

A

A

P P

P P

See Table 15.5 and related discussion.

accident sequence, this release will be controlled for the normal operation case and limited or delayed, as much as possible, for the accident condition case.” The need to prove the safety of NPP in conditions different from the nominal operation of the system, open the way to the definition of PIE. The consideration of the levels of DiD implies the derivation and the fulfillment of the concepts of prevention and mitigation and, among the other things, impose the way for the design and the optimization of ESF and ECCS. PIEs, either incidents or accidents, may therefore be initiated whenever a failure, malfunction or faulty operation of a system or component creates harm or endangers one or more of the established safety functions. Thus, the term PIE refers to an unintended event, including operating errors or equipment failures, which, directly or indirectly, endanger fundamental safety functions. Such an event necessitates protective actions (automatic, manual, on-site, and/or off-site) to prevent or to mitigate the undesired consequences to plant equipment, plant personnel, the public, and the environment. For the purposes of accident analysis, it is reasonable to group all initiating events into categories. There are different sets of criteria for grouping, thus leading to different event

List of “equivalent ADDED accident SCENARIOS” for phenomena characterization

N

Scenario from Table 15.1

Added scenario

Added referencesx

Behavior of check valves Behavior of density locks CCF/CCFL downcomer Channel and bypass axial flow and void distribution CRGT flashing

SBLOCA-1 – LBLOCA-5 SBLOCA-4

Water-hammer LOFW-PIUS ECC-bypass BWR-NC

SBLOCA-4

–

LBLOCA-4

LSTF-SBLOCA

ATWS-1 SBO + AMP

ABWR-pump HOSG-cond

LBLOCA-4

SG-inlet plenum

Hammersley et al. (2000) Tasaka et al. (1992) Yun et al. (2004) Kok et al. (1997) and Yang et al. (2012a, 2012b) Sehgal and Bechta (2016)* (see also Aumiller et al., 2000; Kolev, 2006; Ylonen, 2008)** Nakamura et al. (2009) (see also Khadamakar et al., 2011) Huang et al. (2007) Hyvarinen (1996) and Schaffrath et al. (2001) Khadamakar et al. (2011) (see also Nakamura et al., 2009)

–

REFLOOD-APR1400

CONT-1

LSTF-SBLOCA-NC

MSLB-2

–

MSLB-2

–

Phenomenon ID

A B C D

A-1-CHV A-4-DL S-4-CCF2 I-5-BC

E

S-13-CRGT

F

S-18-ED4

G H

S-30-IPU S-32-LA

I

S-37-LVM5

J

I-14-NC5

K

S-43-NCG

L

S-53-SPR2

Entrainment/de-entrainmentSG mixing chamber Internal pump behavior Liquid accumulation in horizontal SG tubes Liquid-vapor mixing with condensation—SG mixing chamber NC core, vent valves, downcomer Noncondensable gas effect including condensation HT in RCS Spray effects-OTSG SS

M

I-28-SH

Superheating in OTSG SS

Accident scenarios and phenomena in WCNR

Table 15.5

Cho et al. (2009) and Damerell and Simons (1993a,b)+ Takeda et al. (2013) (see also Park et al., 2003) Sankovich and McDonald (1971)++ (see also Guimaraes, 1992) Guimaraes (1992)++ (see also Sankovich and McDonald, 1971) 969

*, Thermal-hydraulic details of CRGT; **, fundamental blowdown; +, various phenomena analyzed; ++, design principles of OTSG.

970

Thermal Hydraulics in Water-Cooled Nuclear Reactors

lists. The grouping of transients aims at constituting consistent sets of events which form the DBA envelope. Grouping possibilities (IAEA, 2002a) include (a) (b) (c) (d)

principal effect on potential degradation of fundamental safety functions; principal cause of the initiating event; frequency and potential consequences of the event; and relation of the event to the original NPP design (for existing plants).

Each category of events is typically subdivided into several more specific events. In some cases a deeper event subdivision is adopted: events which are expected to occur during the plant lifetime are called anticipated operational occurrences (AOO) or anticipated transients. AOO which are associated with the failure of the scram system are called anticipated transients without scram (ATWS). Grouping based on the degradation of safety functions implies listing of safety functions and the search for mechanisms which harm or damage the integrity of the functions. Grouping by principal cause of the initiating events considered in the reactor design brings, for instance, to the categories exemplified in Table 15.1. Grouping by frequency and potential consequences implies a preliminary PSA study, where probability of the event and consequences are estimated. A relationship between probability, consequence, and acceptance criteria is established for each event: the combination of high probability and high consequence for an assigned event is not allowed. Design modifications are needed should such a situation occur. Grouping in relation to the original plant design reveals necessary in order to upgrade the safety of an existing NPP according to new regulations, new computational methods, or needs for design modernization (e.g., power uprating, replacing of key components, etc.). The achievement of a meaningful list of PIE implies the consideration of different (all) grouping possibilities, including the last one when applicable. A large number of individual accident scenarios may result such that a detailed computational analysis may not reveal practicable. Limiting scenarios bounding and enveloping phenomena ad ranges of variations of parameters which characterize those phenomena are typically preselected before full application of calculation procedures (e.g., BEPU, see Chapter 14). In the process of identifying limiting scenarios, the following substeps are recommended (IAEA, 2002a): l

l

l

l

l

l

accident analyses done for similar designs; engineering judgment and expert reviews; “bottom-up” methods in reliability analysis such as failure modes and effects analyses; real operating experience to determine the reliability of equipment; “near misses” or precursor events; and actual events occurred in NPP similar to the concerned one.

15.2.2 Acceptance criteria Acceptance criteria at the same time constitute the motivation for the analysis of PIE and drive the analysis by imposing precision targets and in some cases methods for the analysis. Acceptance criteria are full responsibility of Regulatory Authority and not the subject for the present book (or chapter). However, because of their importance, a few notes derived from IAEA (2002a) are reported as follows.

Accident scenarios and phenomena in WCNR

971

Acceptance criteria set: (a) the numerical limits on the values of predicted parameters, (b) the conditions for plant states during and after an accident, (c) the performance requirements on systems, and (d) the requirements on the need for, and the ability to credit, actions by the operator. Acceptance criteria, as already discussed, are applied to licensing calculations whatever its type (i.e., conservative and best estimate). In the case of BEPU approach (Chapter 14), specific additional criteria may be needed, e.g., criteria for accepting the validation of a computational tool and of the analysts, criteria for deciding the applicability of uncertainty method when wide safety margins are predicted from the “nominal” BE calculation. These criteria may be developed by the analysts on the behalf of the designer or owner for the NPP and approved by the regulatory body. The analyst may set analysis targets to limit economic loss from AOO. An example would be the prevention of fuel dry-out for a loss of flow. As a key feature, more stringent criteria apply for events with a higher probability of occurrence: for instance if DBA and AOO are distinguished based on their frequency, a “no consequential containment damage” criterion is appropriate for DBA, whereas a “no cladding damage” criterion would be appropriate for AOO; similarly, a “no boiling crisis” criterion is appropriate for AOO whereas “cladding temperature less than 1204°C” criterion (in addition to others) is used for LOCA DBA. In relation to basic acceptance criteria fixed by analysts (on the behalf of vendor/ designer/owner), which are more restrictive than the legal acceptance criteria fixed by Regulatory Authorities, the following can be mentioned: (a) The dose to individuals and the public, different for AOO and DBA. These may be given together with the specifications for the duration of the calculated exposure and for the atmospheric conditions to be assumed. (b) An event must not generate a more serious plant condition without an additional independent failure (having low or acceptable probability). Thus an AOO must not evolve into a DBA and a DBA must not evolve into a BDBA or a severe accident. (c) Systems necessary to mitigate the consequences of an accident must not be made ineffective because of conditions caused by the accident, e.g.: (c1) The containment must not be damaged in an LOCA to the extent that it cannot perform its function because of the dynamic effects of whipping of primary coolant pipes; jet forces from the break; pressure generated internally by the break or by combustion of hydrogen; pressure within internal compartments; and high temperatures due to the break or due to combustion of hydrogen. (c2) ECCS pipes must not be damaged by the dynamic forces in an LOCA to the extent that the system becomes ineffective. (c3) If the functioning of any shutdown (non-ECC) system is necessary in an LOCA, it must not be damaged by the dynamic effects of the pipe break. (d) Systems designed for accident mitigation must not be the origin for the plant components to loads or conditions that would exceed the design or failure limits for the accident condition, e.g., thermal and mechanical loads, water hammer, and pressure wave induced loads upon RPV internals. (e) The pressure in the coolant systems must not cause a pressure boundary failure in addition to the accident (e.g., stuck open SRV). l

l

l

l

l

972

Thermal Hydraulics in Water-Cooled Nuclear Reactors

(f ) In the case of AOO the probability of failure of the fuel cladding resulting from a heat transfer crisis must be insignificant. (g) For DBA the fuel damage must be limited in order to ensure coolable core geometry; dispersal of fuel and fission products must be prevented in case of reactivity initiated accidents (RIA). (h) If operator intervention is necessary in an event, it must be demonstrated that the operator has sufficient time, adequate EOP and corresponding training, and reliable information available to initiate and complete the intended action. (i) Sufficient time and adequate means must be available to implement AMP in BDBA conditions, i.e., attempting to prevent the loss of geometric integrity for the core (similar conditions apply for the management of sever accident, not considered here). (j) Accident analysis needs to be continued to the point in time that the plant can be shown to have reached a safe and stable shutdown state, so that: the core remains subcritical; the core remains in a coolable geometry and there is no further fuel failure (i.e., in the long term, before possible fuel removal from the RPV); thermal power is being removed by the appropriate heat removal systems till possible fuel removal from the RPV; and releases of fission products from the containment have ceased, or an upper bound of further releases can be estimated. l

l

l

l

Furthermore, proper attention shall be given to ECCS acceptance criteria other than related to PCT. This may require additional specifications (i.e., acceptance criteria not necessarily part of regulatory documents), like, in case of LOCA:






keeping structural integrity of fuel following unavoidable ballooning of clad and coolable geometry for fuel channels; keeping structural integrity for the RPV internals (e.g., tolerable stresses in case of pressure wave propagation from the break, as already mentioned); assuming loss of shutdown capability by control rods, thus ensuring suitable boron concentration for the long term; in relation to tolerable thickness for crud and oxide; and in relation to maximum burn-up and percentage of mixed uranium-plutonium (MOX) fuel.

The role of nuclear thermal-hydraulics is evident to show fulfillment of the listed acceptance criteria. Some of those criteria are assessed in safety analysis; others may be the subject of specific design calculations. Examples of criteria (e.g., 64 generalized design criteria) can be found in USNRC (1995). Four categories of events together with selected acceptance criteria are defined on the basis of event frequency and potential radiological consequences (ANSI, 1995). Acceptance criteria applicable for WWER reactors are summarized for two categories of event (IAEA, 1997): AOO and “postulated occurrences.”

15.2.3 The accident scenarios: AOO and DBA A set of PIE representative for the envelope of AOO and DBA in NPP is considered in this section. As already mentioned: l

l

the idea is to close a virtual circle including accident scenarios, phenomena, and thermal-hydraulic variables; LWR transients (i.e., primarily PWR and BWR) with a few examples of other reactor types which adopt water as coolant (i.e., CANDU, VVER, RBMK, and PHWR) are selected.

Accident scenarios and phenomena in WCNR

973

Table 15.1 (5 columns and 45 rows) has been created with the objective to gather a comprehensive set of accident scenarios (i.e., according to IAEA, 2002a), which is suitable for characterizing phenomena (i.e., based on the list derived in Chapter 6, see also below). Namely, the column 1 of the table identifies the sections of the present chapter which include the categories of accident scenarios; in columns 2 and 5 individual NPP accident scenarios are reported resulting from calculations described in the literature; columns 3 and 4 deal with elements to characterize the selected individual NPP accident scenarios. An attempt has been made, in the selection of documents/references associated with accident scenarios (column 5 of the table), to encompass the widest possible range of variation for the nature of the document and for the target of the NPP calculation. Therefore, documents have been selected in the last column ranging from journal papers, to internationally agreed reports (e.g., IAEA and OECD), to industry analysis reports, to technology/research project reports, to reports supporting regulations, without distinguishing between simplified and sophisticate approaches. Furthermore no emphasis is given to the quality of calculations; otherwise recent documents are preferred to “old” documents, but examples of old documents are included. The common contents for all documents are constituted by the availability of time trends of variables used to characterize accident scenarios: those variables are adopted here to fix the link between accident scenarios and phenomena. Table 15.1 also provides an idea of the variety of situations where nuclear thermal-hydraulics is applied to the accident analysis including approaches for the application.

15.2.4 Other events: external, “low power,” not core related The current section is introduced having in mind the target to characterize the boundaries (i.e., the narrow field) for the performed activity within the sector of nuclear reactor safety. Events important for the evaluation of the overall risk of an individual NPP include the following: l

l

l

l

l

Originated by natural causes “external” to the NPP like earthquake and flooding, or impact of a meteorite also hitting the surroundings of the NPP. Originated by industrial activities “external” to the NPP like accidental explosion of a truck and arrival of dangerous gas cloud (explosive or poisoned), or impact of an aircraft and including erroneous management of military weaponry. Originated by “internal” accidental events like fire of a diesel generator gasoline tank or missiles generated by the failure of rotating machinery. Not originated at full power: A shutdown transient is actually part of Table 15.1 (i.e., row 43); however, a systematic evaluation of possible events and characterization of phenomena in not full power conditions was not performed. Originated by human sabotage. A large variety of situations may be invented by the human mind to create an accident in NPP, ranging from using chemical explosive, to the injection of de-borated water into the RCS, to coercing the actions of operators.

Furthermore, “Radioactive releases” include events which do not represent the consequences of another event given above; i.e., the release is a direct result of the failure

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

of the component which contains radioactive material. This is the case of the fall of an irradiated fuel assembly during its moving inside or outside the containment, as well as any cooling failure occurring into the spent fuel pool. In all those cases nuclear thermal-hydraulics may support the needed analysis for NPP safety demonstration; however, related accident scenarios and phenomena are not considered within the present framework (i.e., not part of Table 15.1). Noticeably and out of the present framework, the Fukushima Daiichi accident (some details given in Chapter 16) is originated by an external PIE (see e.g., D’Auria et al., 2012). However, thermal-hydraulic phenomena took place: those phenomena occurred in the condition before “loss of geometric integrity for the core” are discussed in Chapter 16.

15.2.5 The transient thermal-hydraulics phenomena Phenomena have been established within the framework of activities performed within international institutions, see also below. Phenomena of interest for the present chapter deal with the condition DBA, before the occurrence of the loss of geometric integrity for the core. The DBA concept has already been discussed in the book: namely, the diagram in Fig. 2.9 of Chapter 2 may be of help. The concerned condition covers some BDBA situations, e.g., in terms of probability of occurrence of a selected PIE, and in cases outside the DBA boundary, when AMPs are utilized to prevent core degradation (e.g., fast depressurization of secondary side of SG). The nuclear thermal-hydraulic phenomena are collected and characterized in Chapter 6. The list of phenomena considered hereafter is similar but does not necessarily coincide with the list in Chapter 6. Four categories of phenomena are distinguished in Table 15.2: basic (B), separate effect (S), integral effect (I), and addressing the design of “new reactors” (A). Both Reactor Coolant System (RCS) and containment phenomena are considered within the four categories. Originating documents for deriving Table 15.2 are: l

l

l

l

l

l

the CSNI SETF CCVM (OECD/NEA/CSNI, 1993); the CSNI ITF CCVM (OECD/NEA/CSNI, 1987, 1996a); consideration is also given to OECD/NEA/CSNI (1989b); the CSNI VVER CCVM (OECD/NEA/CSNI, 2001); various CSNI reports for containment (OECD/NEA/CSNI, 1986, 1989a, 1999, 2014); phenomena from those reports are considered also combining as far as possible with phenomena from other listed sources; the OECD advanced reactor phenomena classification (OECD/NEA/CSNI, 1996b); the IAEA list of phenomena for advanced reactors (IAEA, 2009a); consideration is also given to IAEA (2001) (Tecdoc 1203), IAEA (2002b) (Tecdoc 1281), IAEA (2005b) (Tecdoc 1474), and IAEA (2012) (Tecdoc, 1677).

The phenomena are listed in alphabetic order in Table 15.2. In order to keep compact the same table a higher use is made of acronyms than in the original documents. Three phenomena have been added, which are not explicitly reported in the list of originating documents: these are reported into clear green rows in Table 15.2.

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In some cases the distinction between ITF and SETF phenomena is only formal (i.e., there should be no consequences in the application of the related information). Table 15.2 includes 113 “independent-phenomena” (i.e., labeled rows in the first column of the table): -

-

9 basic phenomena [B-1 to B-9] originated from the OECD SETF CCVM; 61 SETF phenomena [S-1 to S-61]: 58 among those are originated from the OECD SETF CCVM plus 3 ones added [S-25, S-30, and S-42, clear green rows in Table 15.2] within the present context and dealing with horizontal heated channels, internal pumps, and convection flows inside containment, respectively; 30 ITF phenomena [I-1 to I-31] originated from the OECD ITF CCVM; and 13 “advanced reactor” phenomena [A-1 to A-13] derived from above-cited OECD and IAEA documents.

The considered phenomena, namely the “advanced reactor” phenomena, plus the S-42 phenomenon (already mentioned) are assumed to characterize the reactor containment scenarios, too; these include full pressure and pressure suppression containment and the bubble condenser adopted in some NPP equipped with VVER-440; however, no phenomenon is related to the ice-containments, which are excluded from the present framework. Making reference to basic phenomena, the possible steam generation at abrupt discontinuities (i.e., caused by the reversible part of the total pressure drop) is included in the phenomenon B-3-EV1. Downstream a geometric discontinuity (e.g., in the presence of a sharp edge possibly combined with high Re flow), the local pressure may fall below the saturation pressure corresponding to the fluid temperature; vaporization may occur and void may appear which become subcooled void because of the sudden pressure recovery. Subcooled voids may affect the total pressure drop and, if present, the TPCF in the downstream pipe (D’Auria et al., 2003b).

15.3

The cross-link between accident scenarios and phenomena

Within the top-down approach in nuclear system thermal-hydraulics, four hierarchical levels are considered in relation to the topic Accident Analysis: [Level 1] accident scenarios; [Level 2] phenomenological windows; [Level 3] phenomena, and [Level 4] time trends of variables or parameters. It is sometimes useful to introduce the fifth topic, called Relevant Thermal-hydraulic Aspect (RTA) (see Billa et al., 1991). The RTA topic shall appear between the phenomena and the time trends as outlined as follows. -

Accident scenarios are (those) listed in Table 15.1. Phenomenological windows are introduced, as an example whenever appropriate, in order to characterize time spans inside accident scenarios. Phenomena are introduced in Chapter 6 and are listed in Table 15.2: phenomena are originated from internationally agreed and recognized definitions; otherwise RTA is introduced between phenomena and time trends: for instance, the phenomenon I-19-PRZ (Pressurizer thermal-hydraulics) can be characterized by the RTA “PRZ emptying” and “PRZ filling.” RTA is not used in the present chapter.

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-

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Variables, or time trends (of variables), or parameters are introduced in order to characterize accident scenarios and phenomena (with the help of phenomenological windows) and are discussed later.

The targets here are to outline the thermal-hydraulics features of accident scenarios in water cooled nuclear reactors and to complement the description of phenomena provided in Chapter 6. As already mentioned, the list of phenomena is derived from internationally established documents and the overall number (113) should be considered the minimum number of phenomena expected within the DBA envelope making reference to the concerned list of water cooled reactors.

15.3.1 The procedure The cross-connection between phenomena and accident scenarios is defined in Table 15.3. One-hundred-thirteen (113) phenomena are listed in the first column of Table 15.3 and forty-five (45) accidents scenarios are given in the top row of the table. Namely, two accident scenarios from Table 15.1 are reported in each cell of the top row in the following way: accident scenarios 1 (bottom) and 2 (top) are reported in the first cell; accident scenarios 3 (bottom) and 4 (top) are reported in the second cell, and so on up to the 23rd cell which contains only the scenario number 45 from Table 15.1. Furthermore, making reference to the first column, each phenomenon has two rows, A and B, to connect with the bottom and the top scenario (listed in the first row of Table 15.3), respectively. Then, Table 15.3 includes the following symbols: -

-

-

-

“ ” means “phenomenon part of the concerned accident scenario and characterized by at least one figure from corresponding calculation results part of the paper(s) referenced in Table 15.1 and discussed in the section of the present chapter also reported in Table 15.1.” “ ” means “phenomenon part of the concerned accident scenario and characterized to some extent by corresponding calculation results part of the paper(s) referenced in Table 15.1 and discussed in the section of the present chapter also reported in Table 15.1.” “ ” means “phenomenon part of the concerned accident scenario and corresponding calculation results (i.e., typically not considered in available time trends of paper(s) referenced in Table 15.1).” “ ” (red star) and “ ” (clear blue star) mean “phenomenon associated with accident scenario in Table 15.4, third and fourth step for building Table 15.4, respectively; no star implies entering into the fifth step for building Table 15.4.” l

The role of Table 15.3 within the present context is to establish one-by-one mutual correspondence between phenomena and accident scenarios and to provide a visual mapping of the correspondence. The target for Table 15.4 is to characterize the phenomena listed in Table 15.2 and the accident scenarios listed in Table 15.1, throughout the use of parameters. The activity shall consider the cross-link between phenomena and accident scenarios established in Table 15.3. In general terms one may state that: l

Accident scenarios in NPP can be characterized in terms of: - nuclear reactor design and boundary and initial conditions;

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adopted computational tools, including SYS TH code, nodalization, use options, target of the analysis; - imposed and resulting time sequence of events; - phenomenological windows and phenomena; and - time trends of variables or parameters. Phenomena expected in case of accidents in NPP can be characterized in terms of: - Phenomena expected in case of accidents in NPP can be characterized in terms of: - physical models and equations; - ranges of variations of influential parameters; and - time trends of variables or parameters. -

l

Time trends of variables (or parameters) appear in both lists. So, the idea for Table 15.4 is to consider the comprehensive list of phenomena in Table 15.2 and the minimum reasonable number of transient scenarios listed in Table 15.1, in addition to the information from Table 15.3 as already mentioned. In order to achieve the target, phenomena are listed first; then each accident scenario is associated with one or more phenomena. Furthermore, in Table 15.4: -

Time trends of variables (or parameters) representative of accident scenarios are reported in the 15 columns on the right side. “A” in the variables column/cell indicates that the variable is selected to represent the accident scenario in the figures part of Sections 15.4.2–15.4.12. “P” in the variables column/cell indicates that the variable is selected to represent the phenomenon in the figures part of Sections 15.4.2–15.4.12. Cells in clear yellow associate the accident scenario and the section where the accident scenario is outlined (consistently with Table 15.1).

The first step for building-up Table 15.4 consists in associating each accident scenarios listed in Table 15.1 with one phenomenon (i.e., creating the clear yellow cells: so, 45 clear-yellow-cells sets are part of Table 15.4; the related phenomenon and accident scenario are characterized by the “ ” in Table 15.3). The second step [for building-up Table 15.4] consists in associating the remaining 68 phenomena (i.e., 113-45), with one of the sections describing accident scenarios in Table 15.1 (i.e., filling the fourth column starting from left). Then, one of the accident scenarios described in the section (i.e., from Table 15.1) is associated with the phenomenon (“ ” in Table 15.3). The third step [for building up Table 15.4] consists in characterizing the clear yellow-labeled phenomena. In this step the cross-link is established between those phenomena (i.e., 45 phenomena, second column from left) and the parameters calculated from accident scenarios (last 15 columns). The fourth step [for building up Table 15.4] consists in ensuring that each phenomenon (i.e., 113 phenomena listed in the second column) has at least one characterizing “P.” In the case of clear yellow-marked phenomena, the step implies checking that at least one “P” appears in the last 15 columns: if this condition is not true an accident scenario, selected among those listed in Table 15.1, is added in the third column and a corresponding “P” is placed in 1 (or more) of the 15 columns. In the case of nonmarked

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

phenomena one scenario is selected for the third column and the corresponding “P” is added in 1 (or more) of the last 15 columns. In both cases use is made of the information in Table 15.3. The fifth eventual step [for building up Table 15.4] is needed when no parameter is found (or reported) from the database of accident scenarios (i.e., references cited in Table 15.1). In this case a list of “equivalent ADDED accident SCENARIOS” shall be created. The first and second steps have been carried out considering “key” expected phenomena part of the accident scenario: this is the reverse procedure compared with what done to derive the phenomena list at the basis of the cited originating documents. The third and fourth steps imply the prior identification of parameters (variable time trends), which are “suitable” to characterize each phenomenon. This is done in Sections 15.4.2–15.4.12 where each phenomenon is listed and suitable parameters are listed in square brackets. The role of Table 15.4 (and of the supporting Table 15.5) is to establish the bases for the description of accident scenarios considering phenomena and available parameters in Sections 15.4.2–15.4.12. The overall process streamlined by Tables 15.1–15.5, may appear intricate and arbitrary. This is also connected with the existence of multiple solutions occurring when piecing together phenomena and accident scenarios; for instance as a limit case, the basic phenomenon B-7 (Pressure drop at geometric discontinuities including containment) can be associated with every scenario. The process is controlled by the variables available from the calculations documented in the selected references (last column of Tables 15.1 and 15.5). However, as a result (i.e., the expected outcome from the process), an overall picture of transient thermal-hydraulics applied to nuclear reactor safety is derived. Furthermore, the reader may identify the section where any phenomenon is visualized with the help of variables characterizing the accident scenarios; the reader may also recognize the complexity of nuclear system thermal-hydraulics. In the present process for building Table 15.4, the fifth step has been necessary. Thirteen (13) phenomena (i.e., taken from Table 15.2) needed an “equivalent ADDED accident SCENARIO.” These are given in Table 15.5 together with the related (new) reference documents. The listed phenomena, with two noticeable exceptions, are actually part of the modeling resources adopted to calculate the accident scenarios in Table 15.1: i.e., proper parameters needed to depict those phenomena (i.e., A-4-DL and I-14-NC5 in Table 15.5) are not part of reference documents provided in Table 15.1. Furthermore, the “equivalent ADDED accident SCENARIOS” (column 5 in Table 15.5) is not necessarily an NPP calculation. Rather, the related support references include parameter trends suitable for the characterization of phenomena. In Table 15.1, reference documents (last column) deal with calculation of accident scenarios in nuclear reactors; otherwise in Table 15.5 reference documents (last column deal with characterization of concerned phenomenon).

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The bold characters accident scenarios listed in the fourth and the fifth column of Table 15.5 are added in Table 15.4; “P” is added in Table 15.4, where appropriate, considering the availability of parameters from documents in the last column of Table 15.5. The tight connection among phenomena, accident scenarios, and time trends of variables (or parameters) is the first (obvious) result from the process documented in Table 15.4 (supported by added information in Table 15.5).

15.4

The characterization of phenomena and accident scenarios

15.4.1 Background The results from the cross-link procedure for relevant phenomena and accident scenarios in nuclear thermal-hydraulics are outlined in this chapter. Two sub-steps are distinguished for describing the outcomes of the activity: (A) The accident scenarios identified by the sections of the chapters in the first column of Table 15.1 are outlined by the use of time trends (typically consisting of the variables indicated in columns 15 to 1 going from right to left in Table 15.4). (B) The phenomena associated with each section (fourth column of Table 15.4 starting from left) are directly connected with variables describing the accident scenarios.

The activity at the step (A) does not substitute the thermal-hydraulic analysis needed to check the consistency of any set of calculation results and shall not be considered neither an exhaustive description nor an acceptable description of the concerned accident scenario. The activity at the step (B) is based upon variables available from the (arbitrarily selected) references listed in Tables 15.1 and 15.5. Each of Sections 15.4.2–15.4.12.12, except Sections 15.4.12.1–15.4.12.4, 15.4.12.8, and 15.4.12.12 (dealing with single aspects of accident scenarios), is divided into three parts, the first two parts accomplishing the step (A), the third one accomplishing the step (B): l

l

l

(Part 1) Qualitative accident scenario (Part 2) Quantitative accident scenario: variable trends and TSE (Part 3) Phenomena connection with accident scenario

In part 1, thermal-hydraulic characteristics of the accident scenario are listed, having in mind the various classes of water cooled reactors. In a few cases, references in addition to those reported in Tables 15.1 and 15.5 are cited. In part 2, a reasonable minimum number of diagrams, typically less than three, and a table with the sequence of imposed and resulting events are used to depict the accident scenario, where “imposed (I) ¼ part of boundary conditions” and “resulting (R) ¼ obtained from thermal-hydraulic calculation.” The calculated variable trends are derived from references in Tables 15.1 and 15.5. The reported time for events is valid for a single accident scenario and is expected to be representative for the category of accidents considered in the section (i.e., typically the result of a BE calculation where uncertainty is not evaluated).

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

Related to part 3, it shall be recalled that phenomena may be associated with different accident scenarios with the exception of some of the integral phenomena (“I”-labeled in Table 15.2), which are specific for selected accident scenarios: each phenomenon can occur or can be relevant in one or more (or all) accident scenarios and in one or more reactor types, as well. So, phenomena are preferably connected to the accident scenario, which is the topic of the section. Insights into phenomena based on information in addition to what available from references cited in Tables 15.1 and 15.5 are provided in some cases and properly referenced. Finally, not all relevant phenomena, which are expected in the concerned scenario, are necessarily part of the section where the scenario is described. Summing up and related to figures in part 3 of each section: l

l

l

l

Black characters are used for phenomena listed in the section (i.e., each of the 113 phenomena in Table 15.2 is considered one time in the figures). In some cases (e.g., Section 15.4.5) clear green characters are used for identifying phenomena already described in different sections (i.e., which are black in other sections). White characters into clear blue boxes are used for phenomena, which are not part of the section and are part of sections where those phenomena could not be visualized (see e.g., Fig. 15.13). Red characters are used for information other than phenomena (e.g., PHW).

The structure of Sections 15.4.12.1–15.4.12.4, 15.4.12.8, and 15.4.12.12, related to single aspects or single components of accident scenarios, includes (a) topics and facts for the concerned aspect and (b) phenomena connection with accident scenario (same as in other sections).

15.4.2 The LBLOCA and the IBLOCA The LBLOCA in PWR with UTSG is discussed first. The IBLOCA (or MBLOCA) and the scenarios in other reactor types are discussed in the second part of the section.

15.4.2.1 Qualitative accident scenario: LBLOCA The key features of an LBLOCA and key connected facts, CL break in PWR UTSG between MCP and RPV, are: l

l

l

Occurrence of a large break having area till the guillotine break (or “2  100% A”) of the largest pipe in the system, i.e., either the pipe connected with the RPV or the header if the RPV is not part of the system: the break is typically located in the most challenging position as far as core integrity is concerned. TPCF is a key phenomenon (e.g., R
eocreux, 1974; D’Auria and Vigni, 1980; Bartosiewicz et al., 2010). Mechanical loads of RPV internal components and external supports and structures caused by pressure wave propagation and jet impingement in the initial tenths of seconds after the break, and strongly affected by the break opening time (e.g., Vigni and D’Auria, 1979). Void generation (also) in the core region caused by the arrival of the depressurization wave which cause lack of moderation and “decay-power” production with the noticeable exception of those RCS where a typically slight positive void reactivity coefficient causes a power peak (PWR with high-boron concentration in the core at BOL or BOC conditions).

Accident scenarios and phenomena in WCNR l

l

l

l

l

l

l

l

981

Fast depressurization of RCS, the magnitude is in the order of MPa/s, occurs in the early period, a few seconds, after the break occurrence. Accident subdivided into three PHW, or phases, called Blowdown, Refill, and Reflood (BD, RF, and RE, respectively, within the present context; these other than PHW also constitute thermal-hydraulic phenomena). Flashing and somewhat homogeneous conditions occur in the RCS for a couple dozen seconds since the transient starts. Excursion occurs in the value of fuel pin surface temperature (RST). This cannot be avoided and must be controlled to prevent the irrecoverable loss of core integrity (i.e., brittle rupture of clad and release of volatile and nonvolatile fission products into the coolant). Otherwise, clad ballooning and rupture with release of part of volatile fission products cannot be avoided in existing reactors. Stagnation point occurs. The concept of stagnation point is simple: in a closed pressurized loop, the occurrence of a “full size double-ended break” in one point causes two flow streams directed toward the two ends; a region creates unavoidably where the two streams take origin. This is the stagnation region or the stagnation point. Flow velocities are close to zero in the stagnation point and HTC is very low. Furthermore, simple pressure balance analysis shows that favorite region for stagnation point is the loop region with the highest value for pressure drops. Hence the core region in a PWR constitutes a candidate region for the location of the stagnation point. Finally, in a complex loop subject to two-phase flow conditions stagnation point may continuously shift during the transient (i.e., a moving stagnation point condition occurs). This is due to changes during the transient in the repartitions of pressure drops along the paths of the two streams (each path going from the stagnation point to the break), which are different. Additional key LBLOCA characterizing phenomena or events are early core rewet, ECC bypass, steam binding, and occurrences of BD PCT and RE PCT. Role of selected SSC is that PRZ depressurizes later than RPV and piping owing to TPCF occurrence in surgeline; PRZ stored water may contribute to core cooling in the early period of the accident, i.e., by contributing to the early core rewet in the upper region of core; SG behaves as a heat source early during the transient and till full recovery and have a minor role till the end of reflood because of steam present in the PS part of SG tubes and consequent (very) low HTC (i.e., typically less than 100 W/m2 K); MCP work in flow reversal mode (namely in BL) and two-phase conditions soon after the break and have a role in determining the location of the stagnation point; containment pressurizes and liquid level forms in the sump, needed for long-term cooling. Design of selected ESF (including ECCS) and actuation modalities are largely based upon the concerned accident scenario (i.e., noticeably the accumulators). ACC and LPIS are key ECCS to protect the RCS keeping the concerned accident parameter values within DBA conditions.

Classifications of LOCA imply approximations and any parameter selected for the classification may reveal questionable. Hereafter, the following rough “blow-down” (or depressurization)-based classification is proposed making reference to the dimensional parameter AR/V (where AR is the break area [m2] and V the volume of the PS excluding the PRZ [m3]) evaluated in m-1: AR/V > 2.0 e-4 (i.e., up to around 30.0 e-4) 0.2 e-4  AR/V  2.0 e-4 AR/V < 0.2 e-4 (i.e., down to around 0.01 e-4)

LBLOCA IBLOCA SBLOCA

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The origin of the LOCA issue is the high pressure of RCS needed to ensure a suitable thermal efficiency for the electricity production by fission. The geometric configuration of nuclear reactor systems, the nominal-operational value of the pressure, and the linear power of fuel rods play the key role for the evolution of the transient. The LOCA classification of existing commercial reactors based upon geometric configuration brings to the following classes: (A) RPV equipped nuclear systems to be distinguished in: A1: without SG, or the BWR class; further subdivision can be:
A1-1 with external pumps;
A1-2 with external pumps and jet pumps; and
A1-3 with internal pumps (ABWR). A2: with SG, or the PWR class; further subdivision can be:
A2-1 equipped with UTSG;
A2-2 equipped with OTSG;
A2-3 equipped with HOSG (or VVER); and
A2-4 PHWR with moderator cooling loop. (B) Pressurized channel-type nuclear systems to be distinguished in: B1: RBMK type, solid graphite moderated and B2: CANDU type, (low pressure) heavy water moderated. l

l

l

l

For almost all categories, the size and the nominal power of the RCS may bring further differences in relation to: the numbers of loops, of MCP and of pressure channels, and the SG size. The nominal operational value of pressure brings to two LOCA-classes of reactors characterized by initial pressure values around 7 and 15 MPa, respectively. A1 and B1 reactors from the previous classification (geometric configuration based) belong to the former class and remaining ones belong to the latter class. All concerned reactor cores are equipped with fuel rods. The maximum value for design linear power, also controlled by LOCA evolutions, is around 45 kW/m and equal in all classes. Nine main classes of water cooled nuclear reactors are distinguished from the previously mentioned process, as far as expected LOCA scenarios are concerned. The design of ESF and related EOP, including passive systems, introduces further differences inside each class. Then, one may expect that a few dozen transient scenarios (i.e., two to four in each class to demonstrate the worst one) need to be analyzed to cover the LBLOCA performance for all existing NPP. A similar classification procedure can be repeated (this is not done hereafter) in relation to each of the selected accident scenarios (Sections 15.4.3–15.4.12.12) causing an overall number of transient scenarios needed to characterize all existing water cooled reactors in the order of 1000.

15.4.2.2 Quantitative accident scenario: variable trends and TSE - LBLOCA The LBLOCA scenario for a PWR equipped with UTSG can be derived from Table 15.6 and Figs. 15.1–15.3, taken from AP1000 calculation (Queral et al., 2015), row 14 in Table 15.1. Differences between a standard PWR and AP1000

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LBLOCA in PWR with UTSG: (selected) imposed and calculated time sequence of events

Table 15.6 N

Event

I/R

Time (s)

1 2 3 4 5

Break occurrence Break opening time Scram Turbine trip MCP-trip

I I I I I

0.0 103 0.1 0.1 0.1

6

R

1

R

3

R R

5 7

R

10

R

15

Low influence upon the scenario

12

Pressure wave propagation Flow reversal at core inlet CHF occurrence End of subcooled BD PCT occurrence during BD HPIS actuation/ CMT recirculation PRZ emptying

R

20

13

ACC actuation

I/R

20

14 15

Break uncovery Containment pressure peak

R R

25 30

Pressure becomes nearly the same in PRZ and RPV RCS pressure below an imposed set-point TPCF with mostly steam at the break Containment ESF and energy loss to environment contribute to lower the pressure

16

LPIS actuation/ CMT draining Clad ballooning and rupture Minimum mass inventory in RCS PCT occurrence during RE Containment and RCS pressure equalize Reflood completed

I/R

40–250

R

40

In a few regions of the core (typically)

R

50–70

End of BD and start of RF

R

80

R

100

R

100–200

LPIS sump recirculation*/ IRWST draining Reactor in stable conditions End of the event

I/R

1000

R

1000

R

1800

7 8 9 10 11

17 18 19 20

21 22

23 24

Notes

Typically imposed in calculation Also resulting from calculation MCP mode of operation uninfluential upon the scenario Causing loads on RPV internals and containment Possibly, stagnation occurrence in the core region Pressure below saturation pressure at CL conditions

Different time in different locations of the core *RHR mode

When the condition above is reached

984

Thermal Hydraulics in Water-Cooled Nuclear Reactors 18 PHW-BD-SBC

1000

PHW-RE

PHW-BD-SAT PHW-RF

900

16 RCS pressure PCT lower section

14

Hot rod

800 700 600

S-40-LPE B-3-EV1

10

500 8

I-20-RF S-49-QF1 I-7-COTH

400 6

ACC Setpoint

4

Temperature (°C)

Pressure (MPa)

12

300 200

S-41-LPF

2

100

0 0

40

80

120 Time (s)

160

200

0 240

Fig. 15.1 LBLOCA in PWR (AP1000) and phenomena characterization: RCS pressure and RST (two positions). 30

S-15-ECCB

1

S-16-ED2

0.9

Break flow tot Level core

25

S-14-ECCB

Level DC

0.8

0.6 15

0.5 S-59-CF1 S-36-LVM4 S-20-ED6 S-38-LVM6 S-24-GM4

10

Level (–)

Mass flow (kg/s) × 103

0.7 20

0.4 0.3

S-33-LVM1

0.2 5 0.1

I-20-RF

0 0

50

100

150

200

250

300

0 350

Time (s)

Fig. 15.2 LBLOCA in PWR (AP1000) and phenomena characterization: break flowrate and DC/core level.

Accident scenarios and phenomena in WCNR

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30 25

450 400

A-3-CMT CMT-A

15

CMT-B

120 10 5

100

S-1-ACC

350

0

ACC mass flow (kg/s)

0

70

140

210

280

350

Time (s)

80

300 250

ACC CMT IRWST

A-8-GDR

A-3-CMT

200

60

150

40

100 20

CMT & IRWST mass flow (kg/s)

Mass flow (kg/s)

S-38-LVM6

20

50 0 0

130

260

390

520

650 780 Time (s)

910

1040

1170

0 1300

Fig. 15.3 LBLOCA in PWR (AP1000) and phenomena characterization: ACC, CMT, and IRWST flowrate.

are emphasized in the TSE Table 15.6 and in the discussion later. The following can be stressed: l

l

l

l

l

l

l

l

l

The three PHW (Fig. 15.1), BD, RF, and RE, are typical of LBLOCA in any PWR. Containment pressurization is not part of the documented calculation results. The emptying of the PRZ, depending upon the location of the stagnation point, may contribute to the early cooling of the core; the same applies to LP and UH flashing. The loads upon RPV internals and containment structures must be calculated to ensure core integrity after the initial phases of the transient. HPIS in standard PWR and PRHR in AP1000 have a modest impact upon the early phases of the LBLOCA (i.e., up to 300 s). The same is true for ESF in the SS of SG (e.g., AFW). A correspondence can be identified between ACC in standard PWR and in AP1000. In this connection, CMT draining and, later-on, IRWST draining may be seen as corresponding to LPIS actuation (tank feeding) and LPIS sump recirculation. The sump recirculation must be demonstrated taking into account of challenges to the pump (LPIS and/or RHR) cavitation put by liquid temperature in the sump and by debris eventually passing the filters installed at the pump suction location. The opening of vent valves in PWR equipped with OTSG mitigates the effect of steam binding (see also Fig. 15.5). The opening of ADS in AP1000 also mitigates the effects of steam binding.

986 l

Thermal Hydraulics in Water-Cooled Nuclear Reactors

In the case of AP1000, pressure oscillations (e.g., caused by condensation) may occur in the circulation between CMT and RPV and following interactions between RCS and containment.

The quantitative description of the transient scenario is made with reference to a DEGB LBLOCA in CL for a 1000 MWe typical PWR equipped with UTSG, considering six (groups of ) quantities. (I) RPV and PRZ pressure. Five periods can be distinguished in the RPV pressure: (I-a) Pressure wave propagation lasting around 1 s after the start of break opening. This depends upon the concerned location, i.e., close to the break or far away. Pressure undershoots and recovery due to flashing-boiling may occur. (I-b) Subcooled blowdown. The period having duration of a few seconds ends when the local pressure achieves, one after the other, the saturation pressure corresponding to HL and CL (or UP and LP) coolant temperature at nominal conditions. Depressurization rate is of the order of MPa/s. (I-c) Saturated blowdown, with void fraction close to zero at the break. Coolant boiling hitches the pressure decrease in the RPV and depressurization rate is much slower than in the previous period (i.e., in the order of (1/10) MPa/s). Time duration of the period is in the order of 10 s. (I-d) Saturated blowdown, with void fraction close to unity at the break. Steam flow at the break accelerates again the depressurization rate up to values close to 1 MPa/s. However, the decrease of the absolute pressure in the RPV and the arrival at the break of the liquid injected by ECCS slow down the depressurization rate to values of (1/10) MPa/s. (I-e) Pressure equalization in RPV and containment. The conditions for TPCF occurring in previous periods disappear (roughly when RPV pressure achieves a value twice the value in the containment and “Bernoulli-flow” establishes at the break). The transition TPCF ! “Bernoulli-flow” may occur during a time period of a few seconds. The “Bernoulli-flow” condition at the break remains valid for the entire reactor recovery period (several dozen minutes or even hours), with RPV pressure typically above containment pressure due to continuous energy production in the core. Local condensation phenomena and pressure oscillations in the containment may change this situation during short-time periods. ACC and LPIS actuations typically occur during the period (I-d). The PRZ pressure remains above the RPV pressure since the beginning of the transient and till the emptying of the PRZ. The occurrence of TPCF conditions in the surgeline (surgeline diameter in the range 1/5 to 1/10 of CL diameter) is at the origin of the difference between RPV and PRZ pressure. Pressure equalization in RPV and PRZ typically occurs during the period (I-d). The SG pressure remains close to nominal-initial conditions till the start of the period (I-e). (II) RST in hot rod at PCT location. The RST time performance may be (very) different in different regions of the core. Different time performances also occur in the same rod at different elevations. Other than the geometrical coordinates in the core, RST is affected by burn-up (and, in some situations by the amount of boron in the core, i.e., when large amounts may cause reactivity excursions following coolant flashing in the core in the early period of the transient). The local linear power (i.e., q0 in the units of kW/m) largely affects RST. The following time periods can be distinguished in the RST time trend. (II-a) Possible initial fall starting from the initial value: This can occur owing to time changes in the two quantities which control the RST (i.e., fission power and coolant HTC). The decrease in fission power may be faster than the decrease in HTC.

Accident scenarios and phenomena in WCNR

987

CHF occurrence and start of RST excursion till possible BD PCT. CHF cannot be avoided during LBLOCA. Film boiling conditions establish causing RST increase till a possible turnaround point, BD PCT, determined by RCS intrinsic cooling. Overall time duration for the first two periods is less than 10 s and the BD PCT is in the order of 900°C. Occurrence of stagnation point in the RCS largely affects RST value during periods (II-a) and (II-b). Clad ballooning with consequent ductile break may occur in this period as well as in the forthcoming periods. (II-c) Possible (partial or complete) early core rewet. RST decrease after BD PCT may be caused by LP flashing and/or coolant flow entering the core from PRZ and UH. This phenomenon is (again) largely affected by the location of the stagnation point. So, the number of loops and the mutual position of the loop with break and the loop with PRZ affect the phenomenon. The phenomenon is more pronounced in the case of short core (e.g., like in the nuclear reactor LOFT utilized for LBLOCA experiments in the 1970s and 1980s; Reeder, 1978). The duration of the early core rewet phenomenon is in the order of 10 s and causes lowering of RST for a few tens kelvin till the full quench (i.e., RST back at values close to coolant saturation temperature). (II-d) RST increase till RE PCT. Following the occurrence at (II-b) (i.e., with or without the occurrence of early rewet), RST increases till the main turnaround point called RE PCT or simply PCT. The slope of the RST versus time curve remains positive in this period and decreases close to the occurrence of the PCT. Various phenomena (see also below) contribute to the RST value. A key role is taken by the ACC actuation. The period is controlled by the refill conditions and duration at the core PCT location can be of the order of dozen seconds. It must be demonstrated that PCT value remains below 2200°F and other criteria for ECCS design are fulfilled (namely for H2 production which implies a time duration for film boiling conditions). (II-e) “Precursory cooling” till conditions for rewet. Following (RE) PCT occurrence, the effect of ECCS injection causes improved cooling conditions for the clad still not sufficient for the recovery (rewet). This is called precursory cooling and involves droplet transport into the core. QF is progressing typically in a bottom-up direction. CCFL may occur at the UTP (or UCSP) of the core. The precursory cooling period duration is affected by several parameters including location of the observed clad region and distances from QF and spacer grids. Typical values are in the range from a few seconds to a few minutes with RST decreases in the range from a few tens kelvin to a few hundred kelvin. Steam binding and level in UP are expected in this period. (II-f ) Rewet and Return to Nucleate Boiling. Once the local conditions for rewet are locally established, the QF may cross the observed position and RST suddenly drops to values close to the saturation temperature at the RCS (and containment) pressure. The process occurs in a few seconds and RST changes can be up to a few hundred kelvin. (II-g) Long-term cooling. After the occurrence of the rewet, satisfactory cooling conditions are expected to establish in the core and the RST remains connected with coolant temperature. All of this occurs during the so-called long-term cooling period. (III) Coolant flow at core inlet and outlet. Core flow is essential for removing power during nominal operation. MCP provides the power to establish the pressure distribution in the RCS and design flow-rate across the core. Following LBLOCA occurrence, the pressure distribution in RCS depends upon the critical pressure (and the fluid velocities) at the break location (i.e., TPCF conditions). The MCP in the broken loop and in the intact loops (II-b)

988

Thermal Hydraulics in Water-Cooled Nuclear Reactors

typically starts their coast-down due to (presumable) loss of electrical power at within tenths of seconds after the break occurrence. They have some role in establishing the location of the stagnation point in the primary loop. However, cavitation starts almost immediately after the break occurrence and even MCP in the intact loops have little role in determining core flow: the effect of MCP inertia typically expires within 10 s. Three main periods can be distinguished for core inlet flowrate: (III-a) Decay from nominal value and occurrence of flow reversal. In a few seconds (limit has been set to 10 s), flow rate reverses its direction at core inlet and becomes fully driven by break pressure. This condition ends when RF starts. Vaporization including flashing in LP may cause positive peaks during this period. (III-b) RF and RE till equalization of pressure between RCS and containment. Positive value for core flow-rate occurs, needed for the advancement of the QF, which is driven by the ECCS injection flows. (III-c) NC occurrence between core and DC. The DC level is established by the intervention of ECCS (it does not overpass the location of the CL pipe in the broken loop, because of the large break). This determines the NC conditions sufficient for the end of RE and for removing core decay power in the long term. Core outlet flowrate, owing to the possible occurrence of the stagnation point in the core, is disconnected from core inlet flow-rate till the end of the period (III-b). During the period (III-a) core outlet flow may remain positive (i.e., same direction as during nominal conditions) also following vaporization of (part of ) core coolant inventory. During the period (III-b) and part of the period (III-c), the net core flow at top of the core may be close to zero. However, upward steam flow and downward liquid flow may occur. The steam flow may cause CCFL at the UTP leading to the formation of liquid level in the UP. During the period (III-c) a stable steam, or two-phase mixture flow, may establish and the core outlet flow is again connected with the core inlet flow. (IV) Mass inventory in PS and SS. The discussion mentioned previously allows a simple description of the time trend of mass inventory in the RCS PS. Mass inventory decreases following the break occurrence till a value in the range 10%–30% of the initial value. Its increases start with the RF period at the moment when break flow becomes lower than ECCS injected flow. At the end of the RE period the RPV and some regions of the loop (e.g., the loop seal) are full of liquid. The performance of mass inventory in RCS PS is reflected by the core and/or the RPV downcomer level. Mass inventory of SG SS remains nearly constant during the LOCA event (i.e., till the end of the period (II-f )). Later on the mass inventory in the SG is controlled by proper EOP aimed at the recovery and long-term cooling of the RCS. (V) Containment pressure. The pressure in full pressure containment system increases soon after the break occurrence due to the energy flow from the RCS. It attains a peak value in the range 0.4–0.6 MPa at the time 10–30 s into the transient. Then it starts to decrease, owing to the following reasons:
increase in the power removal by condensation on the walls and structures;
heat and mass losses to the environment;
actuation of specific ESF in the containment (noticeably, spray systems); and
actuation of ECC in RCS which also results in subcooled (related to the containment pressure) liquid to the break suitable for condensing steam in the containment. During the early containment pressure decrease period (i.e., at a time 1–2 min since the break occurrence), containment and RCS pressures equalize as discussed in item (I-e). (VI) Sump containment level. A devoted “sump compartment” is designed and constructed in the containment to collect the liquid coolant in the containment including the condensate.

Accident scenarios and phenomena in WCNR

989

At the bottom of the sump there is the suction of LPIS (or RHR) pumps for the ensuring the core long-term cooling. Condensate level starts to form immediately after the break and a level of a couple of meter (or more) is expected to exist when RHR or LPIS pumps start suctioning liquid from the sump at a time which is 150 –300 since the break occurrence (see also Table 15.6, row 22). A nearly stable level in the sump is expected to remain for several hours since the event occurrence allowing core cooling and the full recovery of the NPP. As already mentioned thermal-hydraulic challenges associated with the liquid suction from the sump by LPIS or RHR pumps are the presence of debris and the temperature of the liquid which may cause cavitation.

The depicted scenario for RPV pressure and RST, variables (I) and (II), is consistent with the curves in Fig. 15.1. Namely, RST at two locations in the core are reported. The core inlet flow, variable (III), is not given in the considered curves below. However, an idea of core outlet flow can be drawn from Fig. 15.4 where fluid velocities in one HL are reported. The trend of mass inventory in RCS PS, variable (IV), can be derived from the level curves given in Fig. 15.2. Consistent LBLOCA time trends for pressure and level in containment, variables (V) and (VI), can be found in Section 15.4.12.2. In addition typical time trends for liquid delivery by ECCS can be found in Fig. 15.3. Flowrate and modality of actuation of ACC are qualitatively similar in standard PWR and in AP1000. CMT flow during the draining period and IRWST flow in AP1000 replace (roughly) LPIS flow in short- and long-term cooling mode.

10

1.20 S-1-ACC

8

1.00 0.80

Velocity (m/s)

4

S-17-ED3

2

0.60

0 0.40

−2 −4

Steam velocity

−6

Liquid velocity

−8

Void fraction

Void fraction (–)

6

0.20 0.00 −0.20

−10 10

15

20 Time (s)

25

30

Fig. 15.4 LBLOCA in PWR and phenomena characterization: fluid velocities and void fraction in HL.

990

Thermal Hydraulics in Water-Cooled Nuclear Reactors

15.4.2.3 Phenomena connection with accident scenario: LBLOCA The phenomena listed in the second and the third columns of Table 15.7 are connected with LBLOCA from the cross-link process in Tables 15.3–15.5. Those phenomena are associated with variables in the fourth column of the table.

Phenomena visualized by variables representative of the accident scenario LBLOCA

Table 15.7

Phenomenon ID for: LBLOCA No.

Acronym

Description

Fig. no.

Notes

1 2

S-1-ACC A-3-CMT

15.3 15.3

3 4

I-3-BD I-7-COTH

Accumulator behavior Behavior of core make-up tanks Blowdown Core thermal-hydraulics

5

S-14-ECCB

15.2

6

S-15-ED1

7

S-16-ED2

8

S-17-ED3

9

S-20-ED6

10

B-3-EV1

11

S-24-GM4

12 13

A-8-GDR S-25-HOHT

14 15

S-31-JPU S-33-LVM1

16

S-36-LVM4

17

S-38-LVM6

ECC bypass/Downcomer penetration Entrainment/deentrainment—core Entrainment/deentrainment—downcomer Entrainment/ de-entrainment—hot leg with ECCIa Entrainment/ de-entrainment—UP Evaporation due to depressurization (including at geometric discontinuities) Global multi-D fluid temperature, void, and flow distribution—UP Gravity-driven reflood Horizontal heated channel HTa Jet pump behaviora Liquid-vapor mixing with condensation—core Liquid-vapor mixing with condensation—lower Plenum Liquid-vapor mixing with condensation—UP

See also Fig. 15.4 Almost entire transient Also PHW—BD Variety of local phenomena Also S-4-CCF2, row C in Table 15.5

15.1 15.1

15.2 15.2

During DC level rise

15.4

15.2

High core level

15.1

Not including “at geometric discontinuities”

15.2

15.3 15.22 15.13 15.2

During QF advancement

15.2

15.2

High core level, level rise speed

Accident scenarios and phenomena in WCNR

Table 15.7

991

Continued

Phenomenon ID for: LBLOCA No.

Acronym

Description

Fig. no.

Notes

18

S-40-LPE

15.1

19

S-41-LPF

Lower plenum entrainment Lower plenum flashing

20

S-49-QF1

15.1

21

S-50-QF2

22

I-20-RF

23 24 25

I-21-RE I-24-SBI S-59-CF1

QF propagation/rewet— fuel rods QF propagation/rewet— channel walls, water rodsa Refill including loop refill in PWR-O Reflood Steam binding TPCF—Breaks

Toward end of BD, start of RF During BD (start of BD-SAT) See also RE

15.1

15.13 15.2

Also PHW—RF

15.1 15.5 15.2

Also PHW—RE

a

Characterized by accident scenarios other than the LBLOCA in PWR.

4

1200 CT-1 CT-2 QFE-1 QFE-2

Clad temperature (K)

1000

3.5 3

900

2.5 Vent valves open

800

2 700 1.5

600

I-24-SBI

1

500

Quench front elevation (m)

1100

0.5

400 300 0

100

200

300 Time (s)

400

500

0 600

Fig. 15.5 REFLOOD-APR1400 and phenomena characterization: quench front and RST.

Phenomena are associated with the time trends of the PWR LBLOCA, namely AP1000 calculation, e.g., Figs. 15.1–15.3, except those at rows, 8, 13, 14, 20, and 24 in Table 15.7. Namely, additional diagrams, Figs. 15.4 and 15.5, have been introduced (i.e., based on the cross-link in Tables 15.3 and 15.4 and the reference documents in Tables 15.1 and 15.5):

992 l

l

l

l

Thermal Hydraulics in Water-Cooled Nuclear Reactors

“Entrainment/De-entrainment—Hot Leg with ECCI” is characterized through the LBLOCA in PWR calculation at row 18 in Table 15.1, D’Auria and Galassi (2001), Fig. 15.4. “Horizontal heated channel HT” is characterized through the LOOSP in CANDU calculation at row 22 in Table 15.1, Tong et al. (2014), Fig. 15.22. “Jet pump behavior” and “QF propagation/rewet—channel walls, water rods” are characterized through the SBLOCA in BWR calculation at row 38 in Table 15.1, Analytis and Coddington (2002), Fig. 15.13. “Steam binding” is characterized through the added scenario REFLOOD-APR1400 data at row J in Table 15.5, Damerell and Simons (1993a), Fig. 15.5.

All the phenomena in Table 15.7 are visually characterized based on NPP calculations (Table 15.1) or specific data (typically SETF data in Table 15.5) related to water cooled reactors. The following additional notes apply. (1) Accumulator behavior (visualized, Fig. 15.3). The ACC initial (design) pressure is determined considering the ECC bypass phenomenon. A too high initial pressure causes more coolant to the break. A too low initial pressure causes unacceptable RST excursion. Compromise is needed. The ACC liquid and gas volume must be consistent with the RCS PS volume and with the core power. (2) Behavior of Core Make-up Tanks (visualized, Fig. 15.3). CMT are part of the passive systems of AP1000. The combination ACC, CMT, ADS, PRHR, and IRWST in AP1000 (roughly) substitutes ACC, LPIS, and HPIS in standard PWR. Recirculation phase of CMT is an NC phenomenon (gravity-driven) initiated by different liquid density in suction and delivery lines. The CMT draining is also gravity driven and is made stable by the formation of a thin (thickness of the order of mm) saturated liquid layer which prevents a direct contact (thus avoiding DCC) between steam coming from the core and liquid in the CMT. The break of the layer is expected to cause oscillations which may challenge the cooling process of the core. (3) Blowdown (visualized, Fig. 15.1). This constitutes a PHW and a phenomenon at the same time and is described by pressure trend in the present section. (4) Core thermal-hydraulics (partially visualized, Fig. 15.1 and BWR related). This is described by PWR RST trend in the present section. The three-dimensional core TH performance (i.e., void, temperature, fluid velocities, and—burn-up affected—local linear power) must be noted. In relation to time trends no differences occur between BWR and PWR cores. (5) ECC bypass/Downcomer penetration (indirectly visualized, Fig. 15.2). Large values of break flowrate (and fluid velocities) coupled with DC emptying caused by (fast) depressurization are at the origin of the phenomenon in the early period of the LBLOCA scenario. This shall be associated with CCFL in DC (phenomenon S-4-CCF2, row C in Table 15.5). (6) Entrainment/De-entrainment—Core (indirectly visualized, Fig. 15.2). The transport of liquid droplets by steam and two-phase mixture (e.g., entrainment) occurs in the core at high steam velocities. De-entrainment in the core consists in the deposition of liquid on the solid surfaces. Spacer grids have an influence upon both entrainment and de-entrainment. The phenomenon is more pronounced during the RF PHW, which corresponds to the (indirect) characterization in Fig. 15.2. (7) Entrainment/De-entrainment—Downcomer (indirectly visualized, Fig. 15.2). What reported for core at the item mentioned earlier is applicable for downcomer. (8) Entrainment/De-entrainment—Hot leg with ECCI (visualized, Fig. 15.4). The occurrence of this phenomenon is expected in PWR with HL injection. The calculation of different

Accident scenarios and phenomena in WCNR

(9)

(10)

(11)

(12)

(13)

(14)

(15)

(16) (17) (18)

(19) (20)

993

countercurrent velocities in HL for liquid and steam implies the availability of a model for entrainment/De-entrainment. Entrainment/De-entrainment—UP (indirectly visualized, Fig. 15.2). What reported for core at the item (6) is applicable for UP. In this case, UTP plays a role for entrainment and de-entrainment and CRGT play a role primarily for de-entrainment. Evaporation due to depressurization (visualized, Fig. 15.1). The RCS pressure during BD is the result of various phenomena. The evaporation due to depressurization provides a key contribution to the instantaneous pressure value. The evaporation at geometric discontinuities (part of the present phenomenon) is further discussed in Section 15.4.4. Global multi-D fluid temperature, void, and flow distribution—UP (indirectly visualized, Fig. 15.2). Any TH quantity calculated in the core at any time during the transient (noticeably during the calculation period considered in the reported figures) is the results of multi-D fluid temperature, void, and flow distribution. Core level in Fig. 15.2 for the short period identified in the figure satisfies this condition. What is written in relation to phenomenon (4) above also applies here. Gravity-driven reflood (visualized, in Fig. 15.3, considering the connection with the time trends given in Fig. 15.1). Reflood caused by the CMT injection is “gravity driven.” RST time trends constitute a direct indication for this phenomenon. Horizontal heated channel HT (visualized, Fig. 15.22). A CANDU reactor calculation is at the origin of Fig. 15.22; LOOSP (not LBLOCA) constitutes the target for the calculation. The relationship between void fraction and RST provides an indication of HT in a horizontal channel. Flow stratification puts a challenge for the modeling of the phenomenon: in the same channel cross-section, fuel rods in nucleate boiling and in film boiling, e.g., with RST differing for hundreds kelvin, coexist. Jet pump behavior (indirectly visualized, Fig. 15.13). An SBLOCA BWR calculation is at the origin of Fig. 15.13. The JP position relative to core in the RPV, thus the geometric design of the JP, is established based on TH analyses, namely LOCA (this is not evident from Fig. 15.11, where DC level and not level inside JP are reported). The upper elevation of the JP is sufficiently close to TAF (TOC reported in the figure): the head generated by the subcooled fluid inside JP during LBLOCA compensates the saturated mixture level on the core side and keeps “covered” the core. The JP bottom elevation is fixed in order to allow radially uniform fluid velocity in the BOC region during nominal operation. Liquid-Vapor mixing with condensation—Core (indirectly visualized, Fig. 15.2). Direct contact condensation conditions occur during the RF PHW in LP, core region, and UP. The time trend of core (and DC) level provides an indication of the time when the process occurs. Liquid-Vapor mixing with condensation—Lower Plenum (indirectly visualized, Fig. 15.2). What reported for core at the item listed earlier is applicable for LP. Liquid-Vapor mixing with condensation—UP (indirectly visualized, Fig. 15.2). What reported for core (item 15) is applicable for UP. Lower Plenum entrainment (indirectly visualized, Fig. 15.1). The transport of liquid droplets by steam and two-phase mixture (e.g., entrainment) occurs in the LP at high steam velocities. High steam velocities occur during the flashing. Entrainment into the core region is beneficial for core cooling during BD. Lower Plenum flashing (indirectly visualized, Fig. 15.1). The flashing process is a consequence of the fast depressurization. QF propagation/rewet—fuel rods (visualized, Fig. 15.1). RST time trends provide a direct visualization of the reflood process, including the advancement of the QF. Additional information in Fig. 15.5 and in the description of RST trend in this section.

994

Thermal Hydraulics in Water-Cooled Nuclear Reactors

(21) QF propagation/rewet—Channel walls, Water rods (indirectly visualized, Fig. 15.13). The phenomenon refers to the different quench time for fuel rods, fuel channel box (or pressure tube), and water rods (dummy tubes inside FA in BWR, also used for the insertion of cluster control rods in PWR). Channel walls are considered in the calculation which originates Fig. 15.13 (details of the different behavior of the channel wall, or the pressure tube wall, and the fuel clad during the heat-up process in the core can be observed in Fig. 15.22 related to CANDU). (22) Refill including loop refill in PWR-O (visualized, Fig. 15.2). The level in the core region characterizes the RF phenomenon. (23) Reflood (visualized, Fig. 15.1). RE is the phenomenon and the PHW which shall be related to the overall system performance during LOCA. Otherwise, QF propagation and rewet relate to the core and the local clad TH conditions. However, RE phenomenon is visualized by RST time trends and what reported at item (20) applies. (24) Steam binding (visualized, Fig. 15.5). Steam binding in the late period of reflood is originated by liquid vaporized from the contact with superheated structures like the RPV internals and the SG tubes. The vaporization causes pressure increase and liquid-level depression in the core, this delaying the QF advancement. The QF time trend provides a direct visualization for the phenomenon. Data derived from the 2D/3D experimental program are shown in Fig. 15.5: the actuation of vent valves between core and DC mitigates the steam binding consequences, all other conditions being the same for the two reflood experiments. The differences between the corresponding curves provide a quantitative evaluation of the phenomenon. (25) TPCF—Breaks (visualized in Fig. 15.2). The performance of TPCF at the break can be derived from the description of the pressure trend in the present section.

15.4.2.4 LBLOCA in reactors other than PWR with UTSG LBLOCA in reactors other than PWR equipped with UTSG, here including the standard PWR and the AP1000, may evolve in different ways from a quantitative view-point (i.e., compared with what described earlier). However, key phenomena qualitatively evolve according to the provided discussion. Notes related to LBLOCA in those reactors in addition to, or as a replacement for what stated earlier, are: l

l

l

l

l

l

PWR with OTSG: Steam binding is avoided or controlled by the opening of vent valves between core and downcomer inside RPV. PHWR, CANDU, and RBMK (in the last two cases following break of header upstream the core): The occurrence of fission power peak within a few seconds after the break is expected with a magnitude controlled by the break opening time. In the case of PHWR, the delayed flashing of initially subcooled moderator is at the origin of the fission power excursion. CANDU and RBMK: The occurrence of TPCF in channel feeders (case of break of header in the channel inlet region) is expected. BWR and ABWR: The pressurization is limited inside containments by steam condensation in the PSP (wet-well region). VVER-440 (part of the PWR-V category in Table 15.2): Some of the reactors are equipped with bubble condenser containment type (in one NPP ice condensers are installed) to cope with LBLOCA. VVER-440: A venting-to-the-atmosphere system is installed in the containment to prevent pressure rise above around 0.25 MPa.

Accident scenarios and phenomena in WCNR

995

15.4.2.5 IBLOCA or MBLOCA in PWR IBLOCA (also called MBLOCA) are typically originated by break areas smaller than “1  100 A” (according to the classification at beginning of the section) (e.g., see OECD/NEA/CSNI, 2011). Key aspects during the transient are qualitatively similar to LBLOCA and include the following: (a) transient subdivision into three PHW; (b) occurrence of PCT; (c) need for ACC to cool the core; and (d) negligible role for SG. The IBLOCA analysis is needed for the safety evaluation of reactors to find the worst accident scenario, e.g., to find the highest PCT as a function of the break area. When the break area becomes smaller than a certain value (see also earlier), the SG role becomes important in keeping cooled the core and SBLOCA phenomenology takes place as discussed in Section 15.4.3.

15.4.3 The SBLOCA class The SBLOCA scenario in PWR with UTSG is discussed first. Differences in the scenarios in other reactor types are outlined in the second part of the section.

15.4.3.1 Qualitative accident scenario: SBLOCA Let us start the SBLOCA scenario description by emphasizing one difference related to LBLOCA. One may approximately state that the duration of a typical LBLOCA (till the RCS and containment pressure equalization, or till the full quench of the core) is in the order of 100 s; then, the duration of an SBLOCA (till the RCS and containment pressure equalization) is in the order of 1000 s. However, SBLOCA characterized by several thousand second duration, are part of the DBA envelope and are analyzed (see e.g., Congiu et al., 1996; D’Auria et al., 1996b). SBLOCA scenarios depend upon (a) break area; (b) break location; and (c) ESF actuations as fixed by EOP and operator actions. This causes a large variety of potential scenarios resulting in different safety margins and safety implications. Issues like pressurized thermal shock (PTS) occurrence, recriticality or boron dilution occurrence, MCP operation modality, occurrence of PORV or SRV stuck open and partial availability of AFW, or their combination determine the evolution of an SBLOCA. The first outcome is that the attempt to define a “typical” SBLOCA may reveal meaningless and even misleading. Nevertheless, a typical SBLOCA needs to be identified having in mind the purposes of the present document (i.e., to characterize the scenario and to connect with phenomena). An SBLOCA originated by an around 3% break in CL is considered for the qualitative characterization of the scenario. Key features for the scenario can be listed as follows: l

MCP is assumed to stop early into the transient. The stop may be caused by the scram occurrence, which may cause a perturbation to the electrical grid. Alternatively, MCP may be stopped following EOP. MCP in operation bring to (a) wider safety margin to the occurrence

996

l

l

l

l

l

Thermal Hydraulics in Water-Cooled Nuclear Reactors

of DNB in the core; (b) a tight thermal coupling between PS and SS; and (c) higher mass loss at the break. The key drawback for the MCP-on (other than the difficulty to ensure a sufficient reliability for the electrical grid) shall be associated with the time when the stop occurs (either due to cavitation or following a nonthermal-hydraulic event): the stop may cause void collapse and DNB in the core with SG conditions not necessarily suitable for establishing NC. NC establishes between core and SG when MCP coast-down is nearly completed. Flow regimes in RCS during NC are discussed in Section 15.4.12.4. Three subsequent RPV level (core side) depressions are envisaged and constitute three potential DNB local occurrences in the core (i.e., not necessarily wide spread in the core): - Loop Seal (LS) controlled: This occurs when RCS pressure is around the saturation value corresponding to the fluid temperature in CL at nominal conditions. The origin of this is the mutual position between the LS bottom and the TAF: LS bottom is around 1.5 m below the BAF owing to construction constraints for the mechanical design of the loops including the MCP geometrical configuration. Conditions may establish in the RCS where LS is full of liquid and two-phase mixture or steam are present in other parts of the loop. The liquid in the loop seal creates a plug which puts an obstacle to NC causing a depression in the core level. The situation may stay for a few dozen seconds causing an RST excursion typically less than 200 K. Further level depression in the core causes pressure rise in the HL region which “clears” the LS (LS clearing occurrence) restoring NC. - ACC controlled: This occurs primarily when HPIS is not available (this may signify a BDBA condition) or not sufficient to compensate the mass loss from the break. The RCS pressure is close to the ACC actuation pressure. As already mentioned, the ACC design pressure is determined primarily based on LBLOCA conditions. However, mass depletion in PS may occur during SBLOCA such that ACC intervention is needed to prevent RST excursion. A proper design of ACC pressure avoids or limits the RST excursion. - LPIS controlled: Once ACC are empty and mass loss continues at the break, a new core-level depression event is expected when RCS pressure achieves the LPIS actuation value. In this case the eventual RST excursion rate is higher than in the previous cases, mostly because of the low-pressure condition. Again, a proper design for LPIS intervention conditions (primarily head for the LPIS pumps) avoids or limits the RST excursions (see e.g., Belsito and D’Auria, 1997). SG has the key role for keeping cooled the core. SG SS constitutes the heat sink for the NC loop, where core constitutes the heat source. SG level and pressure are typically controlled (feed and bleed) to attain a depressurization corresponding to around 50 K/h. PS pressure is connected with SG pressure. Core bypass affects SBLOCA scenarios (see e.g., D’Auria and Galassi, 1990a). In the RPV, bypass flow-paths which connected the following regions can be identified (where the direction is identified in nominal operating conditions; this direction may change when MCP stop):
LP to UP, e.g., water rods guiding the cluster control rods.
DC to UP, caused by unsealed connection of HL nozzles and barrel.
DC and UH and UH to UP (including via CRGT) or UP to UH. This is needed to create a circulation in the UH region and affects the initial coolant temperature inside UH. PRZ behavior has a role primarily in case of SBLOCA originated by a break in the PRZ (e.g., PORV stuck open, see Chapter 16 and the description of the Three Mile Island event).

Accident scenarios and phenomena in WCNR l

l

l

997

Break flow may be controlled by stratified conditions upstream the break. Thus the mutual position of the possible break nozzle and the CL axis largely affect the break flow during the transient. Phenomena like vapor pull through and liquid carry-over may occur with break nozzle located at the bottom or at the top of the CL. Mixing at ECC ports. Under the conditions of low flows, mixing between cold liquid injected by ECC and the coolant at the location of the ECC port into the RCS have a role in the motion of the mixed fluid and may affect the global system performance, e.g., by controlling NC flow across the core or the homogeneity of core cooling. Nitrogen and boron effects. N2 may enter the RCS following ACC liquid delivery. N2 may concentrate in the U-tubes of SG decreasing the HT capabilities and affecting NC flow. Changes in boron concentration may occur, namely boron dilution as discussed in Section 15.4.12.8.

A variety of TH topics, PHW or phenomena (not necessarily part of the list in Table 15.2), are associated with SBLOCA. An attempt is made to summarize a following comprehensive list of those topics, including a short outline or reference to more detailed discussions: l

l

l

l

l

l

l

l

l

l

l

l

l

MCP-on or MCP-off: The preferred industry solution nowadays appears the condition MCP-off following the recognition of an SBLOCA by I & C and/or by operators. MCP-restart: Some EOP or AMP includes the possibility of MCP restart to improve the instantaneous conditions of core cooling. NC and related phases: Discussed in Section 15.4.12.4. NC in the presence of many parallel and interacting loops in AP1000. The following NC loops may simultaneously establish in case of AP1000: core-SG, core-CMT, and core-PRHR and in late phase of the transient core-IRWST. Interactions among the loops may generate inducing complex phenomena difficult to analyze. PTS occurrence: Discussed in Section 15.4.12.3. Boron dilution occurrence: Discussed in Section 15.4.12.8. RPV bypass flow-paths: As a difference from LBLOCA bypass flow paths inside the RPV may largely affect the SBLOCA scenario, see the qualitative description earlier. Specific AM strategies and AMP are related to SBLOCA: Discussed in Section 15.4.11. PRZ filling and simultaneous core uncovery: Phenomenon S-7-CCF5 in Table 15.2 (or row 17 in Table 15.4), discussed in this section, see also the TMI-2 scenario in Chapter 16. Stratification in two-phase—horizontal pipes: Mentioned in the present section and in Section 15.4.12.4. Stratification in single-phase, including co-current and counter-current flows and formation of saturated layer at the possible interface with steam: Phenomenon S-35-LVM3 in Table 15.2 (or row 60 in Table 15.4), discussed in Section 15.4.12.3 and mentioned in Section 15.4.2 in relation to CMT draining. Vapor pull through and liquid carryover at break: Outlined in this section and part of phenomenon S-45-PSB in Table 15.2 (or row 78 in Table 15.4), discussed in this section. Heat losses, structural heat release (see also I-26-SHH in Table 15.2), flashing of dead ends connected with the RCS and (small) leakages through imperfectly closed valves. These shall be carefully accounted for in calculating SBLOCA scenarios and introduce distortions in the experimental simulation by ITF (see e.g., D’Auria et al., 1985).

998

Thermal Hydraulics in Water-Cooled Nuclear Reactors

15.4.3.2 Quantitative accident scenario: Variable trends and TSE - SBLOCA As a difference from the DEGB LBLOCA a spectrum of break sizes characterizes the SBLOCA. This reflects in a variety of NPP analyses and in a larger number of typical scenarios. So, a few SBLOCA scenarios are considered for the quantitative analysis below. In all cases the ratio AR/V is within the boundaries for SBLOCA classification proposed in Section 15.4.2. Five main phenomenological windows may be distinguished in typical SBLOCA in standard UTSG equipped PWR (not reported in the following figures because of specific features of the concerned NPP calculations). As a further difference with LBLOCA, PHW in SBLOCA shall not be considered sequential and time intersections may occur. (I) Subcooled BD: from transient start till emptying of PRZ. (II) NC: from MCP stop till reactor recovery (end of transient). NC is characterized by different “modes” including one-phase, two-phase, two-phase with siphon condensation and Reflux-Condenser. (III) RCS PS boil-off: from the end of subcooled BD (or from the end of ACC injection) till the time when LPIS flow becomes larger than break flow (loop refill occurs). (IV) SG control: starting from the time when SG pressure and level are controlled by EOP (or operators) and till reactor recovery (end of transient). (V) Coupling PS and SS: when PS pressure is driven by SG pressure (NC allows full transfer of core power to SG heat sink).

The SBLOCA scenario for a PWR equipped with UTSG can be derived from Table 15.8 (related to an SBLOCA having approximately 3% break area in standard

SBLOCA in PWR equipped with UTSG: (selected) imposed and calculated time sequence of main events

Table 15.8 N

Event

I/R

Time (s)

Notes

1

Break occurrence

I

0

3

Scram

I

10

Break opening time is of low importance in SBLOCA Depends upon various conditions and upon I & C

4 5

Turbine trip MCP-trip

I I

10 20

6

R

80

7 8

NC establishes, core-SG (one-phase) PRZ emptying End of subcooled BD

R R

90 100

9

NC two-phase

R

100

Connected with scram and depending upon EOP Starting from the end of MCP coast-down In small break area SBLOCA, RCS pressure may increase

Accident scenarios and phenomena in WCNR

Table 15.8

999

Continued

N

Event

I/R

Time (s)

Notes

10

First depression of RPV level, core side Effective SG SS pressure and level control HPIS starts

R

120

I

120

Possible first CHF occurrence, LS controlled Feed and Bleed by SRV and AFW

I

130

R

150

R

200

R

300

R

300

I/R

300

R

400

R

400

R R

450 450

I/R

500

Only AP1000: ADS 1–3 23 Containment pressure

I R

600 –

24 Nuclear fuel behavior Only AP1000: ADS 4 25 LPIS sump recirculation* 26 Reactor in stable conditions 27 End of the event

R I I/R

– 1000 1200

R

1500

R

1800

11

12 13 14 15 16 17 18 19 20 21 22

Pressure (nearly) equalization PS-SS NC two-phase/siphon condensation Second depression of RPV level Possible break uncovery ACC actuation End of ACC liquid delivery NC two-phase/reflux condensation Minimum MI in PS Third depression of RPV level, core side LPIS starts

PRHR and CMT recirculation start earlier in AP1000 PS pressure not always higher than SG pressure Pressure and flow oscillations possible Possible second CHF occurrence, ACC controlled Largely depending upon specific local configuration RCS pressure < imposed set-point (higher in AP1000)

Conditions for boron dilution may occur End of PHW III Possible third CHF occurrence, LPIS controlled CMT draining in AP1000. Conditions for PTS Negligible containment role upon RCS performance No fuel damage expected To allow IRWST actuation *RHR mode. IRWST draining in AP1000

When the earlier condition is reached

PWR, including notes related to AP1000), Figs. 15.6–15.8 taken from AP1000 calculation (Yang et al., 2012a) (row 35 in Table 15.1, 1000 break or, approximately, 6% break), Figs. 15.9 and 15.10 taken from APR1400 calculation (Kim and Choi, 2014) (row 36 of Table 15.1, 600 break, or approximately, 3% break), Fig. 15.11 taken from Three Mile Island Unit 2 accident calculation (Bandini and De Rosa, 2014) (row

1000

Thermal Hydraulics in Water-Cooled Nuclear Reactors 18

250

16

RCS pressure ACC flow

200

CMT-1 flow

12 150 10 8

A-10-LTS

A-6-POO

100

6 4

Mass flow (kg/s)

Pressure (MPa)

14

50

2 0 0

250

500

750

1000

1250

1500

1750

0 2000

Time (s)

Fig. 15.6 SBLOCA in PWR (AP1000) and phenomena characterization: RCS pressure and flows from ACC and CMT.

10

45 Core/UP level Break flow

35

Water level (m)

S-46-PS1

8

40

I-17-PFU

30 25

7 20 TAF

6

15

Break flow (kg/s) × 102

9

10 5 5 4 0

250

500

750

1000 Time (s)

1250

1500

1750

0 2000

Fig. 15.7 SBLOCA in PWR (AP1000) and phenomena characterization: break flow and RPV level (core side).

37 in Table 15.1, PORV stuck open), Fig. 15.12, related to LSTF experiments (Nakamura et al., 2009) (rows F and I in Table 15.5), and Figs. 15.13 and 15.14 taken from BWR calculation (Analytis and Coddington, 2002) (at row 38 in Table 15.1). Making reference to the reported time trends and in addition to what stated for the qualitative analysis of the SBLOCA, the following can be added:

Accident scenarios and phenomena in WCNR

1001 180

250 Core power

160

PRHRS heat removal rate ADS 1/2/3 flow (avg)

140

ADS 4/A & 4/B flow (avg)

120

A-11-NC A-13-NCG

150

100

A-6-POO

80 100 60 A-5-PC

50

Mass flow (kg/s)

Power (MW)

200

40

S-45-PSB A-7-CSC

20 0 0

250

500

750

1000

1250

1500

1750

0 2000

Time (s)

Fig. 15.8 SBLOCA in PWR (AP1000) and phenomena characterization: core and PRHR power and ADS flows. 650

9

APR1400 core lvl

8

APR1400 core temp

ATLAS TEST core lvl

600

ATLAS TEST clad temp

7 6

550

5 4

500

S-39-LSC

3

Clad temperature (K)

Collapsed water level (m)

10

450

2 1 S-39-LSC

0 0

200

400

600

800

1000

400 1200

Time (s)

Fig. 15.9 SBLOCA in PWR (APR1400) and phenomena characterization: RST and RPV level (core side) in reduced height facility experiment (ATLAS) and in reactor. l

l

l

l

l

ACC flowrate partially (only) compensates break flow. PRHR removes (in AP1000) a low part of the core power. So SG cooling remains important in AP1000. RPV level (core side) is not expected to fall below TAF in the considered AP1000 scenario. LS clearing is calculated at a time in the prototype (APR1400) different from time measured in the test facility. LS filling is at the origin of the level depression which cause the occurrence of the first CHF condition in the ATLAS experiment.

1002

Thermal Hydraulics in Water-Cooled Nuclear Reactors 12

1.2

APR1400 DC LVF

Collapsed water level (m)

10

1

8

0.8

6

0.6 TAF S-4-CCF2

4

0.4 S-5-CCF3

2

0.2

0

0

400

800

1200

1600

2000

2400

Liquid void fraction of COL-1A (–)

APR1400 DC Lvl

0 2800

Time (s)

Fig. 15.10 SBLOCA in PWR (APR1400) and phenomena characterization: RPV level (DC side) and liquid fraction in CL. 12

3000 PRZ level PCT

Level (m)

10

2500

8

2000

6

1500

4

1000

S-8-CCF6

2

Temperature (K)

S-7-CCF5

500

0 0

12

24

36

48

60 72 Time (min)

84

96

108

120

0 132

Fig. 15.11 SBLOCA in PWR-O (TMI-2) and phenomena characterization: PRZ level and RST. l

ADS actuation (in AP1000), ADS 1–3 and later on ADS 4, is needed for the IRWST draining and the long-term cooling.

15.4.3.3 Phenomena connection with accident scenario: SBLOCA The phenomena listed in the second and the third columns of Table 15.9 are connected with SBLOCA from the cross-link process in Tables 15.3–15.5. Those phenomena are associated with variables given in Figs. 15.6–15.12 in the fourth column of the table.

Accident scenarios and phenomena in WCNR

1003

1.2 HL exp

Normalized liquid level (m)

1

SG U-tube exp S-18-ED4

0.8 S-37-LVM5

0.6

0.4

0.2

0 0

250

500

750

1000 Time (s)

1250

1500

1750

2000

Fig. 15.12 SBLOCA in LSTF (ITF) and phenomena characterization: level in SG mixing chamber and U-tubes. 12

750

700 10

CLL 10%

S-52-SPR1

8

600

TOC

550 6

JP suction elevation

500 4

450

I-11-NC2, Section 15.4.12.4 S-31-JPU, Section 15.4.2

Peak clad temperature (K)

Downcomer level (m)

650

PCT 10%

S-50-QF2, Section 15.4.2

400 BOC

2

S-3-CCF1

350

DC bottom = JP delivery elevation

0

300

0

50

100

150

200

250

300

350

400

450

500

Time (s)

Fig. 15.13 SBLOCA in BWR and phenomena characterization: downcomer level and RST.

The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Behavior of emergency heat exchangers including PRHR and IC (visualized, Fig. 15.8). PRHR removes a part of core power.

8 Pressure (10%)

S-46-PS1

Pressure (1%)

3.2 2.8

Level (5%) I-5-BC, Section 15.4.12.4

Level (1%)

2.4

5

2

4

1.6 3 1.2 2 0.8 1

0.4

0 0

50

100

150

200

250

300

350

400

450

0 500

Time (s)

Fig. 15.14 SBLOCA in BWR, 1%, 5%, and 10%, and phenomena characterization: RPV pressure and collapsed level in the high-power channel.

Phenomena visualized by variables representative of the accident scenario SBLOCA

Table 15.9

Phenomenon ID for: SBLOCA No.

Acronym

Description

Fig. no.

1

A-5-PC

15.8

2

A-6-POO

3

S-3-CCF1

4

S-4-CCF2

5

S-5-CCF3

6

S-6-CCF4

7

S-7-CCF5

8

S-8-CCF6

Behavior of emergency heat exchangers including PRHR, IC Behavior of large pools of liquid CCF/CCFL— channel inlet orificea CCF/CCFL— downcomer CCF/CCFL—HL and CL CCF/CCFL—SG tubes CCF/CCFL— Surgeline CCF/CCFL—UTP, causing pool formation in UP

9

A-7-CSC

Critical and supercritical flow in discharge pipes

15.8

Notes

15.8

See also Fig. 15.6

15.13

Occurring in BWR

15.10

Also S-14-ECCB and row C in Table 15.5

15.10 15.51 15.11 15.16 and 15.11

See “NC” Section 15.4.12.4 TMI-2 scenario LOFW with PS MI decrease (SBLOCA equivalent). Also I-17-PFU below Downstream ADS-3 valves

High power CHAN CLL (m)

Level (10%)

S-13-CRGT, Section 15.4.8

6 Dome pressure (MPa)

3.6

Pressure (5%)

7

Accident scenarios and phenomena in WCNR

Table 15.9

1005

Continued

Phenomenon ID for: SBLOCA No.

Acronym

Description

Fig. no.

Notes

10

S-18-ED4

15.12

ITF data, row F in Table 15.5

11

S-19-ED5

15.51

See “NC” Section 15.4.12.4

12

A-10-LTS

15.6

13

S-37-LVM5

Unavoidable in CMT behavior ITF data, row I in Table 15.5

14

S-39-LSC

15

A-11-NC

16

S-45-PSB

17

S-46-PS1

18 19

I-17-PFU S-52-SPR1

20

A-12-BO

21

A-13-NCG

Entrainment/ De-entrainment—SG mixing chamber Entrainment/ De-entrainment—SG tubes Liquid temperature stratification Liquid-Vapor mixing with condensation— SG mixing chamber Loop seal filling and clearance NC RPV and containment and various system configurations Phase separation at branches (including effect on TPCF) Phase separation/ vertical flow with and without mixture level—core Pool formation in UP Spray effects—core (including cooling and distribution)a Stratification of boron Tracking of noncondensable gases

15.12

15.9 15.8

15.8

ADS and break critical flows

15.14

Boiling process, also Fig. 15.7

15.7+ 15.13

+Not caused by CCFL Occurring in BWR

15.60

See “boron” Section 15.4.12.8 +In containment

15.8+

a

Characterized by accident scenarios other than the SBLOCA in standard PWR.

(2) Behavior of large pools of liquid (indirectly visualized, Fig. 15.8). PRHR is immersed in a pool. Natural convection flows inside the pool, liquid temperature stratification, and boil-off in the pool (late stage of an accident) are part of the calculation. (3) CCF/CCFL—Channel inlet orifice (indirectly visualized, Fig. 15.13). This is related to a BWR and is expected in the SBLOCA4 scenario (row 38 in Table 15.1), during the DC liquid-level depression period.

1006

Thermal Hydraulics in Water-Cooled Nuclear Reactors

(4) CCF/CCFL—Downcomer (indirectly visualized, Fig. 15.10). Level depression in DC during depressurization is typically associated with boiling: The upward steam flow causes the conditions for CCF occurrence. (5) CCF/CCFL—HL and CL (visualized, Fig. 15.10, for CL). Low liquid fraction (or high void fraction) causes the conditions for CCF occurrence. (6) CCF/CCFL—SG tubes (visualized in Section 15.4.12.4, related to NC). NC constitutes a key phenomenon and a PHW during SBLOCA. CCFL-induced filling and draining of U-tubes is part of siphon condensation process described in Section 15.4.12.4, see also phenomenon (11) later. (7) CCF/CCFL—Surgeline (visualized, Fig. 15.11). PRZ filling and RCS mass inventory decrease including core-level depression constitute typical conditions which establish when PORV cycles or remains stuck open (SBLOCA in PRZ in the last case). This is the well-known event occurring in the TMI-2, as also discussed in Chapter 16. (8) CCF/CCFL—UTP (associated with S-38-LVM6, visualized in Section 15.4.4, see also Fig. 15.11 related to TMI-2). CCF and CCFL at UTP typically induce liquid-level formation in UP and consequent liquid-vapor mixing with condensation (this phenomenon is called S-38-LVM6 in Table 15.2 and visualized in Fig. 15.2). The phenomenon S-8-CCF6 has been associated with a loss of feedwater (LOFW) scenario where/when MI depletion occurs in PS (i.e., an RCS configuration characteristic for SBLOCA). (9) Critical and supercritical flow in discharge pipes (indirectly visualized, Fig. 15.8). The phenomenon occurs in a typically 10-m long pipe downstream a relief valve. Such a pipe is installed downstream the ADS-1 to 3 valves in AP1000 RCS in order to allow steam condensation in the IRWST (a similar pipe is not present in ADS-4 in order to minimize pressure differences between PS and containment to allow IRWST discharge). A critical section established at the valve location and supercritical steam flow, Mach > 1, is expected in the downstream pipeline, together with formation of pressure shock fronts. (10) Entrainment/De-entrainment—SG Mixing Chamber (indirectly visualized, Fig. 15.12). Experiments devoted to the study of the performance of the SG mixing chamber (or inlet and outlet plenum) have been performed in experimental facilities including LSTF as reported in row F of Table 15.5, see also following phenomenon (13). (11) Entrainment/De-entrainment—SG Tubes (visualized in Section 15.4.12.4, related to NC). NC constitutes a key phase and a PHW during SBLOCA. Entrainment and de-entrainment inside U-tubes are part of siphon condensation process described in Section 15.4.12.4, see also earlier phenomenon (6). (12) Liquid temperature stratification (indirectly visualized in Fig. 15.6). The CMT recirculation and draining occurrences imply the calculation and the occurrence of liquid stratification inside the tank. (13) Liquid-Vapor mixing with condensation—SG Mixing Chamber (indirectly visualized, Fig. 15.12). Experiments devoted to the study of the performance of the SG mixing chamber (or inlet and outlet plenum) have been performed in experimental facilities including LSTF as reported in row I of Table 15.5, see also earlier phenomenon (10). (14) Loop seal filling and clearance (visualized, Fig. 15.9). The role of LS can be found in qualitative and quantitative descriptions of SBLOCA above. The core-level depression associated with LS formation causes a dry-out occurrence (namely in the upper region of the core, also shown in the figure), which is quenched by the LS clearing. (15) NC RPV and containment systems and various system configurations (visualized, Fig. 15.8). The IRWST draining is part of a global NC condition, which establishes between core and containment system of AP1000. (16) Phase separation at branches, including effect on TPCF (visualized, Fig. 15.8). Phase separation at branches is expected in different location during SBLOCA in PWR, e.g., at PRZ

Accident scenarios and phenomena in WCNR

(17)

(18)

(19)

(20)

(21)

1007

surgeline connection with HL. The TPCF in ADS-1 to 3 is affected by phase separation at the branch between suction pipeline (upstream the discharge valve) and the RCS, or the HL pipe. Phase separation/vertical flow with and without mixture level—Core (indirectly visualized, Fig. 15.7). Boiling in the core, including subcooled boiling, is expected following the SBLOCA induced depressurization even in situations where the RPV level in the core region remains above TAF. Pool formation in UP (visualized, Fig. 15.7). The “UP pool formation” phenomenon may be originated by CCFL at UTP and implies nonuniformity in the distribution of void fraction along the RPV axis. A deep nonuniformity in the void fraction does not appear from the diagram in Fig. 15.7. However, the formation of a time-varying level in UP constitutes the result of the calculation. Spray effects—Core including cooling and distribution (visualized, Fig. 15.13). This is related to a BWR and is expected in the SBLOCA4 scenario (row 38 in Table 15.1), following the actuation of HPCS and, later on, of LPCS. Stratification of boron (visualized, Fig. 15.60). Boron mixing and transport including stratification in vertical components (like DC, liquid pools, etc.) constitute expected SBLOCA phenomena. Boron-related phenomena are described in Section 15.4.12.8. Tracking of noncondensable gases (indirectly visualized, Fig. 15.8). The establishment of global NC conditions, which establishes between core and containment system of AP1000, implies the effect of noncondensable gases.

15.4.3.4 SBLOCA in reactors other than PWR with UTSG SBLOCA in reactors other than PWR equipped with UTSG, here including the standard PWR and the AP1000, may evolve in different ways from a quantitative view-point (i.e., compared with what described earlier). However, key phenomena qualitatively evolve according to the provided discussion (D’Auria et al., 1996a). Notes related to SBLOCA in those reactors in addition to, or as a replacement for what stated earlier, are: l

l

l

l

l

The NC in reactors like BWR, PWR-O, PWR-V (VVER), CANDU, PHWR, and RBMK shows different characteristics from PWR-UTSG. This is discussed in Section 15.4.12.2. Single-phase or two-phase stratification in two-phase horizontal pipes, in the presence of nuclear fuel, is typical of CANDU reactors, see also discussion in Section 15.4.2. The presence of fuel boxes in BWR, PHWR, and VVER-440 affects the SBLOCA scenario also by causing a specific NC flow path, i.e., between core active region and bypass region. Cooling of the core by moderator during SBLOCA occurs in CANDU and PHWR. The actuation of ADS is foreseen during BWR SBLOCA. The ADS actuation causes the SBLOCA to evolve into an IBLOCA with break in the steam region. The ADS actuation is needed to allow the intervention of the low-pressure ECCS and the system recovery.

15.4.4 The LOFW The LOFW scenario in PWR with UTSG is discussed considering the three parts proposed in Section 15.4.1. The following preliminary notes apply:

1008 l

l

l

l

Thermal Hydraulics in Water-Cooled Nuclear Reactors

The LOFW accident scenario is EOP and operator action dependent, like all or most of long-lasting accidents (e.g., time duration >about 300 ). So, different LOFW evolutions shall be expected in individual NPP units. No loss of integrity for RCS occurs. Low (or no) role for the containment because of item listed earlier. Part of group of four scenarios where similarities in system performance are expected (at least in relation to some periods and some phenomena), e.g., LOFA (or MCP-trip), LOFW, SBO, and LOOSP (respectively, Sections 15.4.12.9, present section, 15.4.5, and 15.4.6).

Differences in the scenarios in other reactor types are outlined in the second part of the section.

15.4.4.1 Qualitative accident scenario: LOFW Feedwater allows power removal from SG SS during the nominal operation. The loss of FW is immediately detectable because of the decrease in the level in the affected SG. In a few seconds scram and isolation of the turbine occur. Then, the recovery procedures for the system are actuated. Key features for the scenario can be listed as follows: l

l

l

l

Core cooling by natural circulation, assuming MCP-trip occurs early into the transient. During the initial dozen seconds PS pressure may increase till the opening of the PORV and SS pressure increase may also cause opening of SRV. MI decrease in PS may occur due PORV opening. Typically this does not cause DNB situations in the core. MI recovery may occur due to the HPIS intervention. RCS recovery is based upon feed and bleed of SG.

15.4.4.2 Quantitative accident scenario: variable trends and TSE - LOFW Four main phenomenological windows may be distinguished during typical LOFW in standard UTSG equipped PWR: (I) Loss of heat sink and PS pressurization lasting a few dozen seconds. (II) Affected SG boil-off, till AFW restoration: this may take a few minutes. (III) SG feed and bleed and PS recovery till the actuation of RHR (PS pressure around 2 MPa): this may take a couple hours. (IV) Long-term cooling: from the previous event till the time when stable cooling conditions are reached at nearly atmospheric pressure: this may take a few hours.

The LOFW scenario in case of unavailability of AFW for a PWR equipped with UTSG can be derived from the TSE in Table 15.10 and Figs. 15.15 and to 15.16 taken from OECD/NEA/CSNI (2011) (row 21 in Table 15.1). The recovery of RCS is achieved by feed and bleeds in PS; namely, combined operations of HPIS and PRZ discharge valves allow the recovery. Making reference to the reported time trends and in addition to what stated for the qualitative analysis of the LOFW, the following can be added: l

Following an initial peak during PHW-1, PS pressure stabilizes at a value close to saturation temperature in UP during the PHW-2.

Accident scenarios and phenomena in WCNR

1009

Table 15.10 LOFW in PWR equipped with UTSG: (selected) imposed and calculated events N

Event

I/R

Time (s)

Notes

1

Loss of main and auxiliary feed water

I

0

2

Scram

I

12

3

Condenser dump valves begin to cycle PORV and PRZ safety valves open MCP-trip

I/R

12

I/R

12

100% Initial power. The transient started by ramping down main feed water to zero over 5 s Following low SG level: 8 m in any SG To bring the RCS to “no load average coolant temperature” PRZ pressure exceeds 17 MPa

I

370

Steam dump valves cycling stops HPIS actuation PRZ PORVs manually locked open Voiding of UP begins NC stop in PS RPV level below TAF PRZ full of water

R

500

MCP operation largely affect the scenario SG SS almost empty

I I

610 670

Primary system feed enabled Two PORV available

R R R

770 800 800

R

1100

R

2400

R

6600

R

8000

4 5 6 7 8

9 10 11 12 13 14 15

l

l

l

l

PORV flow equals HPIS flow ACC actuation Transient calculation terminated

PORV exit void fraction equal to zero

Initial ACC pressure equal to 4.64 MPa

The operator control of the PORV (at around 670 s into the transient as from Table 15.10) causes the PS pressure to fall at the SS pressure. The situation of pressure equalization is achieved without causing core uncovery (Fig. 15.16). Following the pressure equalization, the RCS configuration can be synthesized as follows (Figs. 15.15 and 15.16): - PORV bleed flow balanced by HPIS feed flow; - liquid level at top of PRZ (consistently with the condition PORV open); - UP level formation and core level below TAF (row 11 in Table 15.10); and - RST not experiencing dry-out (however, margin to DNB not known). Further lowering of the RCS pressure causes ACC intervention during PHW-4 and, later on, the RHR-LPIS actuation and the recovery of the unit.

1010

Thermal Hydraulics in Water-Cooled Nuclear Reactors

20

50

PHW-1 PHW-2

PHW-3

PHW-4

18

40 S-54-SPR3

14

RCS pressure SG SS pressure PORV HPIS PRZ level

12 10 8

35 30 25 20

I-19-PRZ

6

15

4

10

Level (m) Mass flowrate (kg/s)

Pressure (MPa)

16

45

5

2 B-7-PD

0 0

1000

2000

3000

4000

5000

6000

7000

0 8000

Time (s)

Fig. 15.15 LOFW in PWR and phenomena characterization: PS and SS pressure, PRZ level, and HPIS and PORV flowrates.

590

1

580 PCT

S-38-LVM6

0.8

UP liquid fraction

560

0.6

S-8-CCF6, Section 15.4.3

550 0.4

540

Liquid fraction (–)

Temperature (K)

570

530 0.2 520 510 0

1000

2000

3000

4000

5000

6000

7000

0 8000

Time (s)

Fig. 15.16 LOFW in PWR and phenomena characterization: liquid fraction in UP and RST.

15.4.4.3 Phenomena connection with accident scenario: LOFW The phenomena listed in the second and the third columns of Table 15.11 are connected with LOFW from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables in Figs. 15.15 and 15.16, as reported in the fourth column of the table.

Accident scenarios and phenomena in WCNR

1011

Table 15.11 Phenomena visualized by variables representative of the accident scenario LOFW Phenomenon ID for: LOFW No.

Acronym

Description

Figa. no.

1

B-7-PD

15.15

2 3

I-19-PRZ S-54-SPR3

Pressure drops at geometric discontinuities including containment PRZ thermal-hydraulics Spray effect-PRZ

Notes

15.15 15.15

a

The use of Fig. 15.16 is finalized to the visualization of the phenomenon S-8-CCF6 from Table 15.9.

The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Pressure drops at geometric discontinuities including containment (visualized, Fig. 15.15). The transient evaluation of PDGD is needed to calculate fluid velocity in any location of a thermal-hydraulic system (e.g., RCS and containment) and occurs in any accident scenario. As already mentioned, the phenomenon has been (arbitrarily) assigned to LOFW and is visualized through the HPIS flowrate. Furthermore, two parts for the total pressure drop at any geometric discontinuity shall be distinguished known as reversible and irreversible contribution. Only the latter term contributes to PDGD and is considered in calculations by SYS TH codes through the “K-loss” coefficient. The reversible part may be at the origin of local evaporation if the saturation pressure associated with upward liquid temperature is higher than the pressure low-value which establishes within a short distance (typically mm) at the discontinuity. In the case of the inlet of a cylindrical nozzle/pipe where TPCF occurs at the outlet, vapor generated in at the entrance may largely affect TPCF. This local phenomenon is called cavitation at sharp edge (D’Auria et al., 2003a). Thus, the evaporation occurrence at geometric discontinuity is also relevant to phenomenon B-3-EV1 in Section 15.4.2. (2) PRZ thermal-hydraulics (visualized, Fig. 15.15). PRZ performance is relevant during all scenarios in PWR. The situation of PRZ full and PS emptying depicted in Figs. 15.15 and 15.16 is possible in the case of PORV continuous operation and occurrence of CCFL at the surgeline connection with HL. (3) Spray effects—PRZ (indirectly visualized, Fig. 15.15). During LOFW with MCP in operation, the PRZ pressure can be controlled by spray line to prevent the PORV opening or to mitigate the mass loss from the PORV.

Furthermore, phenomenon S-8-CCF6 from Table 15.9 is visualized in Fig. 15.16.

15.4.5 The SBO The SBO accident scenario is part of group of four scenarios where similarities in system performance are expected (at least in relation to some periods and some phenomena), e.g., LOFA (or MCP-trip), LOFW, SBO, and LOOSP (respectively, Sections 15.4.12.9, 15.4.4, present section, and 15.4.6).

1012

Thermal Hydraulics in Water-Cooled Nuclear Reactors

SBO evolves into LOOSP when the on-site electrical power generation (i.e., from Diesel Generators) is lost. However, the distinction in terms between SBO and LOOSP is not always recognized within the international community.

15.4.5.1 Qualitative accident scenario: SBO Qualitative accident scenario for SBO reflects the scenario described for LOFW. Furthermore, following scram (i.e., assuming the availability of the scram function): Basically, in case of SBO: LOFW and MCP-trip occur. PRZ PORV actuation and SRV actuation in SG are not avoidable. Diesel Generators (DG) enter in operation at the earliest time, typically a few seconds after the event. AFW is the key system to recover the plant. NC is the only mechanism to remove decay power from the core. PHW is basically those identified in case of LOFW (Section 15.4.4).

-

15.4.5.2 Quantitative accident scenario: variable trends and TSE - SBO PHW in case of SBO is similar to those identified in the case of LOFW (as already mentioned) and is indicated in Figs. 15.17 and 15.18. Typical SBO scenarios for a PWR equipped with UTSG can be derived from the TSE in Table 15.12 and Figs. 15.17 and 15.18 taken from Bittan (2015) and D’Auria et al. (2006), rows, 33 and 34 in Table 15.1, respectively.

500

17

Pressure (MPa)

16.5

PHW-2

I-26-SHH

16

PHW-3 RCS pressure Loop flow SU-LI flow Core level

S-61-CF3

S-6-CCF4

300 200

I-9-INC

15.5

400

100 15

0 S-48-PS3 I-27-SULI S-7-CCF5

14.5

I-10-NC1, Section 15.4.12.4

14 0

1000

2000

3000 4000 Time (s)

5000

6000

Mass flowrate (kg/s) Rated level (%)

PHW-1

−100 −200 7000

Fig. 15.17 SBO in PWR with UTSG and phenomena characterization: RCS-PS pressure, flowrate in one loop, and across PRZ Surgeline and core level.

Accident scenarios and phenomena in WCNR

20

PHW-1 PHW-2

1013

350

PHW-3

18

330 RCS pressure SG SS pressure PCT (2/3 core elevation) RCS residual mass integral

S-37-LVM5

14 I-15-NC6

12 10

310 290 270 250

S-32-LA

8

230

6

210

4

I-13-NC4, Section 15.4.12.4

2

Temperature (°C) Mass (ton)

Pressure (MPa)

16

190 170

0 0

5000

10000

15000

20000 25000 Time (s)

30000

35000

150 40000

Fig. 15.18 SBO in PWR with HOSG and phenomena characterization: RCS-PS and SS pressure, RST, and RCS-PS MI.

Table 15.12 SBO in PWR equipped with UTSG: (selected) imposed and calculated events N

Event

I/R

Time (s)

Notes

1 2

I I

0 0

DG are called in operation

3

Station blackout Scram, MCP-trip, LOFW, Isolation of turbine NC starts

R

40

4 5 6

AFW starts PORV first opening SG boil-off starts

I R R

40 40 100

7 8 9 10

SRV first opening NC stops Core uncovers ECCS actuation

R R R I

200 – – –

11

ACC actuation

R

2000

12 13

ERVCS actuation RCS recovery

R R

– 7000

At the end of MCP coast-down Following DG power Feed and Bleed by AFW and SRV, respectively Not expected Not expected Passive safety systems (e.g., ACC) are operational Following PS pressure <4 MPa SAMP, not expected

1014

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Following SBO occurrence scram occurs (almost instantaneously). Loss of flow in PS occurs together with partial loss of the heat sink (no FW available). This causes a boil-off condition in both PS and SS. Openings and cycling of PORV and SRV in PRZ and SG are expected. Specific EOP can be designed to mitigate the consequences and to recover the RCS.

15.4.5.3 Phenomena connection with accident scenario: SBO The phenomena listed in the second and the third columns of Table 15.13 are connected with SBO from the cross-link process in Tables 15.3–15.5. Those phenomena are associated with variables in Figs. 15.17 and 15.18, as reported in the fourth column of the table. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Liquid accumulation in horizontal SG tubes (indirectly visualized, Fig. 15.18). Horizontal tubes in VVER-440 and VVER-1000 are connected with inlet and outlet headers at different elevations. Condensation is at the origin of liquid formation in the horizontal tubes and may affect NC. Additional information, not in the form of time histories, can be derived from fundamental research documented in the papers at row H of Table 15.5, Hyvarinen (1996), and Schaffrath et al. (2001). (2) Phase separation/vertical flow with and without mixture level—Pipes and Plena (visualized, Fig. 15.17). The mixture core level allows the (direct) visualization for the phenomenon. Phase separation without mixture level shall be taken as a synonymous of Collapsed Liquid Level (CLL). CLL established in a saturated liquid occurs in the absence (or low) of thermal power exchange and depressurization rate. Otherwise mixture level establishes which is effective for ensuring nucleate boiling conditions. Mixture level formation in the secondary side of HOSG by Melikhov et al. (2011).

Table 15.13 Phenomena visualized by variables representative of the accident scenario SBO Phenomenon ID for: SBO No.

Acronym

Description

Fig. no.

Notes

1

S-32-LA

15.18

See row H in Table 15.5

2

S-48-PS3

15.17

See Melikhov et al. (2011) for mixture level in the SS of HOSG

3

I-26-SHH

15.17

Relevant in any transient. See also Fig. 15.17

4

S-61-CF3

Liquid accumulation in horizontal SG tubes Phase separation/vertical flow with and without mixture level—Pipes and Plena Structural heat and heat losses TPCF—valves

15.17

Accident scenarios and phenomena in WCNR

1015

(3) Structural heat and heat losses (indirectly visualized, Fig. 15.17). The release of structural heat and the value of heat losses constitute a challenge in the experimental simulation of RCS. In scaled ITF, the values of thermal power from passive structures and heat losses to environment per unit core power are (much) greater than in the NPP. Furthermore, the values of both quantities can be affected by errors. Rather than a phenomenon, structural heat and heat losses shall be seen as a boundary condition affecting various phenomena and creating distortions in the simulation of accident scenarios in NPP by ITF. Structural heat and heat losses may play a relevant role in all transients. (4) TPCF—Valves (indirectly visualized, Fig. 15.17). Opening of PORV and SRV of SG causes TPCF. The geometric complexity of flow paths in valves, including the possible occurrence of multiple critical sections and/or changes in critical section location during a transient, makes necessary the experimental evaluation of TPCF in valves. The TPCF during PORV cycling causes the pressure fluctuations shown in Fig. 15.17.

In Figs. 15.17 and 15.18, phenomena which are not part of Table 15.13 are indicated in clear green (see also discussion in Section 15.4.1). Furthermore, phenomena I-10-NC1 and I-13-NC4 from Section 15.4.12.4, Table 15.27 are visualized in Figs. 15.17 and 15.18 (white characters in light blue label), respectively.

15.4.6 The LOOSP The LOOSP accident scenario is part of group of four scenarios where similarities in system performance are expected (at least in relation to some periods and some phenomena), e.g., LOFA (or MCP-trip), LOFW, SBO, and LOOSP (respectively, Sections 15.4.12.9, 15.4.4, 15.4.5, and present section). LOOSP is the follow-up of SBO when the on-site electrical power generation (i.e., from Diesel Generators) is lost. However, the distinction in terms between SBO and LOOSP is not always recognized within the international community. Reactor protection system is designed to avoid the occurrence of LOOSP (i.e., stopping the accident severity to the level of SBO). Therefore LOOSP is a BDBA. The interest for the accident scenario in the present chapter is till the time when core damage occurs.

15.4.6.1 Qualitative accident scenario: LOOSP The qualitative accident scenario in the case of LOOSP is the same as SBO roughly for about 1 h after the event occurrence: -

Decay power produced by core is removed by NC to the SG. PRZ pressure stabilizes close to the opening/closure pressure of PORV: this causes coolant MI loss. SG pressure stabilizes close to the opening/closure pressure of SRV: SG emptying occurs in a time between 0.5 and 1 h. PORV cycling causes the well-known condition of PRZ full of coolant and core-level depressed: CCFL at surgeline connection with HL in combination with depressurization at PRZ top caused by valve opening are at the origin of the phenomenon. It may be noted

1016

Thermal Hydraulics in Water-Cooled Nuclear Reactors

here that the presence of a large volume PRZ in the RCS is not beneficial for core cooling during the LOOSP scenario.

In the absence of the actuation of any AM procedure (e.g., timely PS and/or SS depressurization) MI lost causes dry-out in core region and severe core damage at time which is around 2 h after the occurrence of the event. The actuation of AMP (see later, also discussed in Section 15.4.11) may largely shift the time of core damage.

15.4.6.2 Quantitative accident scenario: variable trends and TSE - LOOSP PHW in case of LOOSP is similar to those identified in the case of LOFW (as already mentioned) and is indicated in Figs. 15.19–15.22. Typical LOOSP scenarios can be derived from the TSE in Table 15.14, PWR with UTSG, and Fig. 15.19, VVER-1000, taken from D’Auria et al. (2006) (row 22 in Table 15.1), Fig. 15.20, APR1400 taken from Yu et al. (2013) (row 32 in Table 15.1), and Figs. 15.21 and 15.22, CANDU, taken from Tong et al. (2014) (row 22 in Table 15.1). Possible objectives for (quantitative) LOOSP analyses are: A. to estimate times of first PORV and SRV opening, loss of SG heat sink, stop of NC, core uncovery, core temperature above (licensing) acceptable value, and core loss of geometric integrity; B. CET values and timing for possible start of AMP; and C. to estimate the applicability and validity of AMP.

The estimation of validity of SAMP for mitigation of severe accident consequences may constitute an objective for application of severe accident computational tools. 600

14 PHW-1

500

PHW-2

PHW-3

PHW-4

SG SS pressure

I-20-RF I-21-RE

10

PCT

400

PRZ level 8

300

S-26-HT1 I-15-NC6 S-26-HT1

6

Level (m)

Pressure (kgf/cm2) Temperature (⬚C)

12

RCS pressure

I-27-SULI

S-10-CO2, Section 15.4.12.3 200 4 S-57-HOSG

100

2 S-1-ACC

0 0

5000

10,000

15,000

20,000

25,000

30,000

0 35,000

Time (s)

Fig. 15.19 LOOSP in PWR with HOSG and phenomena characterization: PS and SS pressure, RST (PCT elevation), and PRZ level.

Accident scenarios and phenomena in WCNR

1017

20

Timing affected by I-26-SHH

PHW-1

0

S-61-CF3

16 S-26-CHT1

Pressure (MPa) Level (m) Mass (kg/1.E4)

14

−50

PRZ pressure SG SS level

12

−100

MSSV mass integral SG heat removal rate

10

−150

8

−200

6

B-4-EV2

Power (MW)

18

50 PHW-2

−250

4 −300

2 0 0

1000

2000

3000

4000

−350 5000

Time (s)

Fig. 15.20 LOOSP in PWR with UTSG and phenomena characterization: PRZ pressure, SG SS level and thermal power exchanged across SG, and mass lost from MSSV. 12

PHW-2

PHW-1

140 PHW-3

Pressure (MPa)

100

I-10-NC1

8

ROH pressure SG SS pressure PHTS mass inventory SG SS mass inventory PRZ LRV mass flowrate

6 4 2

80 60 40 20

S-61-CF3

0 0

2000

4000

6000

Mass (kg/1 .E3) Mass flowrate (kg/s)

120

10

8000

10,000

0 12,000

Time (s)

Fig. 15.21 LOOSP in CANDU and phenomena characterization: PS and SS pressure and MI and flowrate across the PRZ relief valve.

The calculation in Fig. 15.19 shows the effectiveness of selected AMP to keep the core coolable for more than 10 h. The calculations in Figs. 15.20–15.22 show timing (predicted or foreseeable) for core damage in the range between 2 and 4 h applicable for current PWR and CANDU, in the last case considering the intrinsic availability of the heat sink constituted by the moderator.

1018

Thermal Hydraulics in Water-Cooled Nuclear Reactors

1200

1.2 Fuel clad temperature

1 Temperature (K)

Calandria tube temperature

0.8

12th bundle void fraction

600

0.6 S-25-HOHT, Section 15.4.2

0.4

Void fraction (–)

Pressure tube temperature

900

300 0.2 0 0

3000

6000

9000

0 12,000

Time (s)

Fig. 15.22 LOOSP in CANDU and phenomena characterization: core void fraction and RST.

Table 15.14 LOOSP in PWR with UTSG: (selected) imposed and calculated events N

Event

I/R

Time (s)

1 2

I I

0 0

3 4

Station blackout—LOOSP Scram, MCP-trip, LOFW, Isolation of turbine NC starts SRV first opening

R R

0 10

5 6 7 8 9 10

PORV first opening SG boil-off completed NC stops Core uncovers SIS actuation SIT actuation

R R R R I R

100 2000 5000 5000 – –

11

ERVCS

R

–

12

Core damage occurring

R

5000

Notes

Failure of DG on demand and consequently of AFW SRV cyclically open (SG SS pressure >8.0 MPa)

Caused by PS MI decrease Mixture level blow TAF AMP may be actuated RCS pressure always higher than SIT pressure (4.51 MPa) This is an SAMP (preparation for this may start)

15.4.6.3 Phenomena connection with accident scenario: LOOSP The phenomena listed in the second and the third columns of Table 15.15 are connected with LOOSP from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables in Figs. 15.19 and 15.21, as reported in the fourth column of Table 15.15. In Fig. 15.19 one phenomenon from

Accident scenarios and phenomena in WCNR

1019

Phenomena visualized by variables representative of the accident scenario LOOSP

Table 15.15

Phenomenon ID for: LOOSP No.

Acronym

Description

Fig. no.

1 2

I-27-SULI B-9-FR

Surgeline hydraulics Wall to fluid friction

15.19 15.21

Notes Can be visualized from any transient

Section 15.4.12.3 is also visualized (clear blue label). Figs. 15.19–15.21 include identification of phenomena visualized in different accident scenarios (clear green) and Fig. 15.22 allows visualization of a phenomenon in Section 15.4.2 (clear blue label). The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Surgeline hydraulics (indirectly visualized, Fig. 15.19). Emptying and filling of PRZ, as well as mass and energy exchange between PRZ and HL are affected by surgeline hydraulics. This phenomenon also determines the conditions for CCFL (in the surgeline) and the consequent filling of PRZ (see also Doi et al., 2012). (2) Wall to fluid friction (indirectly visualized, Fig. 15.19). The transient evaluation of wall to fluid friction is needed to calculate fluid velocities at any location of a thermal-hydraulic system (e.g., RCS and containment) in single- and two-phase conditions. The phenomenon occurs in any accident scenario. It has been (arbitrarily) assigned to LOOSP and is visualized through the core

15.4.7 The SGTR The SGTR is originated by the leakage from one or more tubes or the (guillotine) rupture of those tubes in SG. In the case of VVER, the acronym PRISE includes leakages and ruptures (“Primary to Secondary”), which can also occur through the headers which support the tubes in HOSG. A Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR) as well as a PRISE in VVER can lead to radioactivity release to the environment bypassing the containment. The radioactivity may be transported via the secondary system, exiting through the SRV of the affected Steam Generator. That is the main reason why SGTR historically have been treated in a special way within Deterministic Safety Analysis (DSA), focusing on the radioactive release more than in the possible core damage, as it is done in the other LOCA. This also constitutes the motivation for the present section. Section 15.4.12.5 in this chapter deals with correlations between MSLB and SGTR and includes mentioning of the multiple SG tube rupture. Therefore focus for the present section is the single-tube rupture.

1020

Thermal Hydraulics in Water-Cooled Nuclear Reactors

15.4.7.1 Qualitative accident scenario: SGTR In a single-tube SGTR event, the pressurizer pressure and level start to decrease simultaneously with the decrease of the Main Feedwater flow of the faulted SG to compensate the extra mass flow from the ruptured tube. In this phase, the operator can guess which SG is the faulted one with the help of radiation monitors in the steam lines. The operators can start to manage the transient before the scram decreasing the turbine load and therefore the reactor power. They can also increase the charge flow and stop the letdown trying to compensate the break flow of the ruptured tube. Once the reactor is shutdown (manually o automatically), the operators have to identify (or confirm the identification) the faulted SG and isolate it stopping the AFW and closing the Main Steam Isolation Valve and the turbo-driven pump steam feed. Once the faulted SG is isolated, the main goal is to stop the faulted tube leak via primary and secondary pressure equalization. Therefore, in a first step, the PS is cooled down via SS depressurization through intact SG SRV or steam dump to the condenser. Then the PS is depressurized to the faulted SG pressure via PRZ spray or PORV. The following topics are relevant for the SGTR, accident scenario: A. Radioactivity release. The radioactivity release to the environment with containment bypass is of main concern for the accident scenario as already mentioned. B. Accident parameters. Boundary conditions for the accident like elevation the rupture and time of detection of the event may affect the overall scenario and the radioactivity release. The elevation of the break affects the release of radioactive materials into the liquid or steam phase of the SG SS. In the former case lower release to the environment and delay in detection of the accident may be expected. Moreover, the level in the affected SG SS should be kept above the elevation of the break, although identification of break elevation is typically not possible. The time of detection of the event depends upon settings of operation set-points for system or parameters like PRZ and SG level and CVCS, also affected by unavoidable leakages between PS and SS during nominal operation. SGTR detection time including time needed to identify the faulted SG may range between a few minutes and a couple of dozen minutes. C. Boron. Flow reversal at the rupture (i.e., from SS to PS) may cause on the one hand stop of radioactivity release into the SG and on the other hand inlet of deborated liquid into the primary coolant. So the EOP design aim at keeping close the PS and SG pressure with suitable margin to avoid inlet of SS water into the PS. It may be noted that “boron risk” for core integrity depends upon the time into the fuel cycle period: higher at BOC and lower at EOC (consequences of boron dilution also discussed in Section 15.4.12.8). D. Pressurized Thermal Shock. Fast PS depressurization and overfeed by HPIS (highly borated water) may cause an overall decrease of PS to SS mass transfer and radioactivity lost from PS, but create the potential for PTS. EOP design take into account of such risk (PTS discussed in Section 15.4.12.3). E. Integrity of SS challenged. Overfeeding of SG SS, i.e., the broken SG because of the break flow and because of the need to minimize radioactivity release to environment (as discussed in item B.) and the intact SG in order to ensure (or to make easier) the control of PS pressure, may be part of the EOP design (see also “quantitative” description of the accident scenario later). This causes the condition of “solid” SS with risk for water hammer and/or condensation induced water hammer (IAEA, 2009b).

Accident scenarios and phenomena in WCNR

1021

15.4.7.2 Quantitative accident scenario: variable trends and TSE—SGTR The TSE table related to the Doel-2 SGTR accident is reported in Table 15.16 as summarized by Reocreux (2008). Typical SGTR scenarios can be derived from Table 15.16 and Figs. 15.23 and 15.24, Doel-2, PWR with UTSG taken from OECD/NEA/CSNI (1988) (row 41 in Table 15.1) and Fig. 15.25, VVER-440, taken from IAEA (2009b) (row 31 in Table 15.1). The time histories in Figs. 15.23 and 15.24, either calculation results or data measures during the Doel-2 event, give an idea of the management of the event. Three PHW are distinguished identified in Table 15.16 as diagnosis phase, mitigation phase part 1 (getting the control of the RCS) and mitigation phase part 2 (toward the RCS recovery). The PRISE in VVER-440, Fig. 15.25, has been calculated under the condition of isolation for the feedwater flow to the affected SG, and of trip of the main coolant pump associated with the defective SG. In addition, cooldown through BRU-K (condenser dump valve) on intact loops, primary system depressurization, and termination of the HPSI system intervention were imposed. The time histories in Fig. 15.25 show (a) the asymmetric SG performance and (b) the occurrence of the over-feed condition for all SG.

Table 15.16 SGTR in PWR equipped with UTSG: events from Doel-2 accident (1979) Initiating event—about 1 s -

Rupture: 7 cm longitudinal crack in U bend

-

Break flow around 15 kg/s

Diagnosis phase—may take 15 min -

Decrease of PRZ level.

-

Decrease of PS pressure.

-

Increase of level in affected SG.

-

Activity in affected SG.

Mitigation phase (getting the control of RCS)—approximately 1 h: -

in the loop with broken SG for short time

Faulted SG: (a) Isolation (except discharge turbo pump); (b) discharge valve set point at the maximum. Intact SG: (a) Steam discharge to achieve: (a1) the highest depressurization rate; (a2) PS temperature decrease; (b) opening steam discharge (8 min opening) to turbine FW pump to keep high the level

-

MCP: tripped to minimize power input to PS

-

ESF actuation: (a) Low PRZ level causes safety injection signal, then: (b) 4 HPIS pumps started, and (c) PS pressure stabilized

Mitigation phase (toward the RCS recovery)—approximately 3 h -

Equalization of PS and affected SG pressure to reduce leak rate.

-

MCP: restarted: (a) to spray in PRZ, (b) to decrease PS pressure; (c) to increase HPIS flow.

-

Safety injection canceling phase: (a) HPIS tripped after check of suitable sub-cooling in PS; (b) operation of charging and let-down system; (c) continuous PS pressure decrease under high sub-cooling and high PRZ level conditions.

-

PS pressure at the value of Head of RHR pumps: long term cooling-recovery starts.

Thermal Hydraulics in Water-Cooled Nuclear Reactors

18

10

PHW2

PHW1

16 PRZ pressure

Pressure (MPa)

14

PRZ level

12

8 6

10

4

8

2

6

0

4 −2

2 0 0

500

1000

1500

2000

Collapsed water level (m)

1022

−4 2500

Time (s)

Fig. 15.23 SGTR in PWR with UTSG and phenomena characterization: PRZ pressure and level.

Break mass flow (kg/s)

14

15

12 14

10 8

13

I-8-FO S-60-CF2

6

12

4

Flow rate

2

Collapsed water level (m)

16

16

11

Water level

0 0

500

1000

1500

2000

10 2500

Time (s)

Fig. 15.24 SGTR in PWR with UTSG and phenomena characterization: break flow and affected SG level.

15.4.7.3 Phenomena connection with accident scenario: SGTR The phenomena listed in the second and the third columns of Table 15.17 are connected with SGTR from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables in Figs. 15.24 and 15.25, as reported in the fourth column of Table 15.17. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible):

Accident scenarios and phenomena in WCNR

1023

3.3 3.1 I-15-NC6

2.9

SG 1 SG 2 SG 3 SG 4 SG 5 SG 6

Water level (m)

S-57-HOSG

2.7 2.5 2.3 2.1 1.9 1.7 1.5 −600

600

1800

3000

4200

Time (s)

Fig. 15.25 PRISE in VVER-440 (PWR with HOSG) and phenomena characterization: level in SG SS.

Table 15.17 Phenomena visualized by variables representative of the accident scenario SGTR Phenomenon ID for: SGTR No.

Acronym

Description

Fig. no.

Notes

1

I-8-FO

Flow through openings

15.24

2

I-15-NC6

NC with horizontal SG

15.25

3

S-57-HOSG

15.25

4

S-60-CF2

Thermal-hydraulics of horizontal SG, PS, and SS TPCF—pipes

See also Section 15.4.12.11 See also Fig. 15.18 and Section 15.4.12.4 See also Fig. 15.18

15.24

(1) Flow through openings (visualized, Fig. 15.24). The phenomenon deals with RCS shutdown conditions when the PS is open to the containment (Section 15.4.12.11). Modeling features needed to predict shutdown conditions flow at openings may be the same (or similar) needed to predict flows at SGTR break when TPCF conditions are not established and flow reversal (i.e., from SS to PS) occurs. (2) NC with horizontal SG (visualized, Fig. 15.25). Natural circulation during SGTR is expected in the loop with the affected SG. In intact loops the MCP may be in operation to allow RCS pressure control by spray in PRZ. (3) Thermal-hydraulics of horizontal SG, PS, and SS (visualized, Fig. 15.25). Heat transfer across SG is essential to recover RCS following SGTR. One may also state that SGTR is a special SBLOCA, and SBLOCA is characterized by the importance in the role of SG. Asymmetric SG performance is expected during SGTR and can be noted from Fig. 15.25.

1024

Thermal Hydraulics in Water-Cooled Nuclear Reactors

(4) TPCF—Pipes (visualized, Fig. 15.24). TPCF at the rupture location occurs during the “undetected period” (i.e., nominal RCS conditions with anomalies in PRZ and SG levels) and, once the SGTR scenario occurrence is recognized, in the period when SS pressure remains below the critical pressure that establishes at the rupture location. The geometric shape of the rupture may have a role in TPCF values (see F
eburie et al., 1993).

15.4.8 The MSLB and the FWLB In spite of specific control program, two guillotine pipe breaks of feed water system piping have occurred in 1990 and 1993 at Unit 1 and Unit 2 of Loviisa NPP in Finland. These two pipe breaks and inspections performed have revealed that wall thinning can be very local in nature, and wall thinning of similar components in parallel lines can be completely different (Korhonen and Hietanen, 1994). After the first guillotine pipe break, the extent of control program was increased to comprise about 600 items to be inspected in each unit every year. This did not avoid the second guillotine break in 1993. An MSLB constituted the topic of the first benchmark organized by OECD/NEA/ NSC to demonstrate the capabilities of coupled three-dimensional neutron kinetics thermal-hydraulics codes. The MSLB was assumed in the Three Mile Island unit 1 and implied the calculation of recriticality or return-to-power (e.g., D’Auria et al., 2003a). The earlier two statements give an idea of the importance of the MSLB issue in PWR. In the case of BWR, (a) the MSLB may cause the largest (pressure and temperature) loads for containment, (b) the opening of ADS, e.g., following an SBLOCA is equivalent to an MSLB. Furthermore, qualitative aspects may not be distinguishable between MSLB and FWLB. Namely, full pipe cross-section MSLB creates a faster depressurization transient than a full pipe cross-section FWLB. MSLB complementary topics also relevant in other accident scenarios are discussed in this section: (a) Transient three-dimensional neutron flux calculated for nonhomogeneous core conditions during MSLB is visualized. (b) Snap-shot information about thermal-hydraulics of the OTSG is given, connected with the OECD/NEA/NSC TMI-1 benchmark and with the phenomena listed in Table 15.2 (also part of Table 15.19).

15.4.8.1 Qualitative accident scenario: MSLB and FWLB Depressurization of the SG SS constitutes the first effect of MSLB (and FWLB): from the blowdown analysis view point the SS of SG has similar features as the RPV of a BWR (e.g., including initial pressure and average void fraction in the system). Specific topics of interest for MSLB (and/or FWLB) are: A. MSLB and RIA. Fast depressurization implies fast temperature decrease in SS. This causes a step coolant temperature decrease in PS, which propagates into the reactor core inducing a positive reactivity perturbation (further discussed in this section).

Accident scenarios and phenomena in WCNR

1025

B. MSLB and SGTR. Pressure wave propagation from the break location and fluid-dynamic loads associated with high fluid velocity may challenge the integrity of SG tubes, i.e., creating the conditions for SGTR discussed in Section 15.4.12.5. C. MSLB and PTS. Continuous operation of MCP after the event in combination with possible actuation of HPIS may create the condition for large fluid temperature decreases and negative temperature gradients inside DC of RPV, i.e., the potential for PTS, discussed in Section 15.4.12.3. D. In some NPP all steam lines and all FW lines cross the containment in the same zone: isolation valves are installed in each line. The break of one pipe in that region may affect the integrity of other pipes: a hypothetic situation of multiple SL and FW line break is created, which may also challenge the maximum pressure allowable for the containment. E. The MSLB originated by a break downstream the location of isolation valve constitutes the reference design accident for the turbine building. Pressurization of the building is terminated by closure of MSIV: design of accident detection system and of called ESF shall be consistent with the maximum allowable pressure in the building.

15.4.8.2 Quantitative accident scenario: variable trends and TSE—MSLB and FWLB Typical MSLB and FWLB scenarios can be derived from Table 15.18 (MSLB in PWR), and Fig. 15.26, MSLB in PWR with UTSG, taken from Jeong et al. (2006) (row 26 in Table 15.1), Fig. 15.27, FWLB in APR1400 experimental simulator, taken from Bae et al. (2014) (row 10 in Table 15.1), Fig. 15.28, MSLB in PWR with UTSG and OTSG, taken from Jeong et al. (2006) and D’Auria et al. (2003a) (rows 26 and 27 in Table 15.1, respectively), and Fig. 15.29 taken from Sankovich and McDonald (1971) and Guimaraes (1992) (rows L and M in Table 15.5, respectively).

Table 15.18 MSLB in PWR with UTSG: (selected) imposed and calculated events N

Event

I/R

Time (s)

Notes

1

Break occurrence 100% SL section Scram Turbine isolation MDNBR MCP operation Return-to-power conditions SG blowdown completed RCS recovery

I

0

R I R I R

10 10 10 – 60

R

70

I

7000

Abrupt decrease in turbine inlet steam flow Originated by high neutron flux Consequence of scram Occurring before scram Depending upon EOP See for example, OECD/NEA/NSC benchmark, Ivanov et al. (1999) Pressure in SG attains containment pressure Approximately

2 3 4 5 6 7 8

Thermal Hydraulics in Water-Cooled Nuclear Reactors 610

16

590

Cold Leg A1 (broken side)

580

Cold Leg B1 (intact side)

13

I-1-ASY-L

Hot Leg A (broken side)

570

12

560

11

5 Minimun DNBR (−)

Pressure

14

6

600

I-1-ASY-L

Temperature (K)

Pressure (MPa)

15

Reactor trip signal

1026

550

4 3 2 1

540

10 0

5

10

15 Time (s)

20

25

30

0 0

780

4000

5

10 15 Time (s)

20

Density decreases

720

2400 Density

1600

Power

700

800

680

Total core power (MWt)

3200

100% Power

740

Reactor trip signal

Coolant density (kg/m3)

I-1-ASY-L

760

0 0

5

10

15 Time (s)

20

25

30

Fig. 15.26 MSLB in PWR with UTSG and phenomena characterization. Top left: PRZ pressure, coolant temperature in HL and CL. Bottom: power and average coolant density in the core. Top right: DNBR. 0.5

Normalized pressure (−)

0.4

I-1-ASY-L S-23-GM3

SG-1

0.3

SG-2 0.2

0.1

0 0

1000

2000

3000

4000

5000

6000

Time (s)

Fig. 15.27 MSLB in PWR with UTSG (simulation of ITF experiment) and phenomena characterization: asymmetry in SG pressure and long-term system performance.

Typical RCS performance during MSLB can be derived from Fig. 15.26. The break occurrence causes a cold water plug arriving into the core in a few seconds, which causes a smooth fission power excursion and scram. Fluid temperature in HL is affected about 10 s after the event. DNBR achieves a minimum just before scram, that is, owing to the slight power increase.

Accident scenarios and phenomena in WCNR

1027

t=0s

t=0s

t = 60 s

t = 90 s

Fig. 15.28 MSLB in PWR with UTSG (left) and OTSG (right) and phenomena characterization: three-dimensional performance of neutron flux and (detail in top left) local void production in the hot FA.

600 Temperature (K)

Primary

1.0E−02 S-53-SPR2

PSD steam pressure (Pa)

1.0E−03

I-28-SH

Secondary

580

560

S-58-OTSG

1.0E−04

540 0

4 8 12 Distance from bottom (m)

1.0E−05

16

1.0E−06 1.0E−07 1.0E−08 1.0E−09 1.0E−10 0

0.03

0.06

0.09

0.12

0.15

0.18

0.21

0.24

0.27

0.3

Frequency (Hz)

Fig. 15.29 Thermal-hydraulics of OTSG (associated, in the present document to MSLB accident scenario): PSD of the OTSG outlet pressure and (detail) temperature profile along the axis for PS and SS fluid.

The expected asymmetric performance of SG following FWLB is depicted in Fig. 15.27 (related to the ATLAS facility simulating APR1400). The same diagram gives an idea of the time needed for the recovery of the RCS.

1028

Thermal Hydraulics in Water-Cooled Nuclear Reactors

The three-dimensional core power (or thermal neutron flux, or fission power) at different times during the transient can be seen from the sketches in Fig. 15.28 (left related to UTSG PWR, same calculation at the origin of Fig. 15.26; right related to OTSG PWR where a stuck withdrawn CR is assumed in the same region of the core where [cold] flow the broken SG is coming). Occurrence of subcooled voids in the hot FA of the core can also be seen in Fig. 15.28, left. A simplified overview of the OTSG thermal-hydraulic performance, which is relevant to any transient scenario and to phenomena listed in Table 15.2 can be derived from Fig. 15.29. On the basis of a suitable thermal-hydraulic model, the study of the pressure spectrum derived from NPP data (this shows a maximum at the frequency of 0.21 Hz) brought to the proposal of design modification for minimizing the oscillations. The oscillations were originated by density wave propagating through the boiling length.

15.4.8.3 Phenomena connection with accident scenario: MSLB and FWLB The phenomena listed in the second and the third columns of Table 15.19 are connected with MSLB and FWLB from the cross-link process in Tables 15.3–15.5. Those phenomena are associated with variables in Figs. 15.26–15.29 and with Figs. 15.68 and 15.14 taken from other sections of this chapter, as reported in the fourth column of Table 15.19. Table 15.19 Phenomena visualized by variables representative of the accident scenario MSLB (and FWLB) Phenomenon ID for: MSLB and FWLB No.

Acronym

Description

Fig. no.

1

I-1-ASY-L

2

I-2-ASY-D

15.26/ 15.27 15.68

3

S-13-CRGT

Asymmetric loop behavior Asymmetry due to the presence of a dam CRGT flashing

4

S-23-GM3

5

S-53-SPR2

6

I-28-SH

7

S-58-OTSG

Global multi-D fluid temperature, void, and flow distribution—SG SS Spray effects—OTSGSS Superheating in OTSG SS Thermal-hydraulics of OTSG, PS and SS

15.14

15.27

15.29 15.29 15.29

Notes

Part of the shutdown transient Part of SBLOCA + ADS scenario in BWR Occurring in various accident scenarios

Calculation results in Fig. 15.28 take into account of these phenomena

Accident scenarios and phenomena in WCNR

1029

The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Asymmetric loop behavior (visualized, Figs. 15.26 and 15.27). The basic nature of MSLB and FWLB accident scenarios induces asymmetric loop (RCS) behavior. This also reflects in radial and azimuthal nonuniformities in thermal-hydraulics and neutron flux in the core (Fig. 15.28). (2) Asymmetry due to the presence of a dam (indirectly visualized, Fig. 15.68). Dams are located in primary loop to keep a certain collapsed liquid level during RCS maintenance and refueling. Dams may introduce asymmetries in the loop performance and constitute part of the modeling for reactor shutdown analyses (Section 15.4.12.11). (3) CRGT flashing (indirectly visualized, Fig. 15.14). Flashing of CRGT located in the lower plenum of BWR is delayed related to flashing of surrounding liquid in the LP. The reason for this is the higher subcooling inside CRGT (i.e., related to LP fluid) caused by high-pressure liquid continuous injected by an RPV external source for ensuring the high-pressure sealing and to prevent leakages of RPV fluid. Flashing is expected during depressurization caused by ADS opening in Fig. 15.14 (i.e., an MSLB type of situation). Geometric details about CRGT can be found in the paper by Sehgal and Bechta (2016); blowdown studies dealing with the influence of subcooling can be found in the fundamental studies by Aumiller et al. (2000), Kolev (2006), and Ylonen (2008) (cited at row E in Table 15.5). (4) Global multi-D fluid temperature, void, and flow distribution—SG SS (indirectly visualized, Fig. 15.27). An FWLB transient causes multi-D transient inside the SG. The modeling of the phenomenon affects the calculation of pressure shown in Fig. 15.27. Primary system parameters (Figs. 15.26 and 15.28) are also affected by multi-D modeling of SG SS through the HT across tubes. (5) Spray effects—OTSG-SS (indirectly visualized, Figs. 15.28 and 15.29). The thermalhydraulics of OTSG includes spray effects (in SG vertical tubes) and superheating for steam at the outlet of SG. Namely, feedwater in the OTSG is introduced through 32 spray nozzles connected to 14-inch semicircular headers (Sankovich and McDonald, 1971). Spray effect is part of the modeling capabilities to calculate OTSG performance at nominal conditions and during any accident scenario. The same considerations apply to the OTSG phenomena S-53-SPR2, I-28-SH, and S-58-OTSG in Table 15.19. (6) Superheating in OTSG SS (visualized, Fig. 15.28). See discussion at item (5). (7) Thermal-hydraulics of OTSG, PS and SS (visualized, Figs. 15.28 and 15.29). See discussion at item (5).

15.4.9 The RIA—reactivity excursion class of accident scenarios A variety of RIA may occur in each NPP and are of interest to nuclear thermal-hydraulics. A comprehensive listing of RIA situations considered for each NPP as well as related accident scenarios is well beyond the purpose of the section. Chernobyl event discussed in Chapter 16 constitutes an RIA case. RIA and related scenarios are also part of safety demonstration for NPP. This constitutes the motivation for the section.

1030

Thermal Hydraulics in Water-Cooled Nuclear Reactors

15.4.9.1 Qualitative accident scenario: RIA-CRE Selected accident scenarios associated with RIA and relevant issues and topics (not an exhaustive list, including repetitions from other sections) are (see also D’Auria et al., 2004): A. Background issue 1: RIA analysis (as a difference from majority of accidents) needs the consideration of instantaneous core status. Conditions like HZP, HFP must be distinguished. For instance in LOCA analysis starting from HFP causes a challenge upon ESF more severe than starting from HZP. This is not necessarily true in the case of RIA where a number of initial core status need proper consideration (e.g., including cold zero power and intermediate status with the presence of neutron sources triggering the fission reaction). B. Background issue 2: Beginning of Cycle (BOC), End of Cycle (EOC) conditions, as well as intermediate conditions (MOC) and BOL, need proper consideration in RIA analyses (see also the earlier item). C. Background issue 3: Material composition of nuclear fuel including presence of plutonium (MOX fuel) is important for RIA analyses. Namely, transient production of poison materials like xenon and samarium needs proper consideration. D. Background issue 4: RIA analysis needs three-dimensional transient modeling including suitable thermal-hydraulics (proper consideration of the phenomenon “Global multi-D fluid temperature, void, and flow distribution—Core”) and neutron physics models considering local material properties, noticeably nuclear cross-sections. The same analytical capabilities are needed for ATWS analyses discussed in Section 15.4.10 (rows 1–3 in Table 15.1). E. Background issue 5: RIA analyses are needed for a number of component designs. For instance, (i) to confirm the acceptability of the design (worth) of a single CR, (ii) to confirm the design of a group of CR which move together; and (iii) to fix the maximum tolerable boron concentration at BOC. F. Background issue 6: During an RIA transient nuclear fuel structure and geometric configuration may change (e.g., fuel fragmentation, ballooning, etc.). This requires specific consideration. G. Example 1 for accident scenario, instability in BWR: oscillations in neutron flux coupled with oscillations in thermal-hydraulic parameters like local velocity, pressure, temperature, and void fraction may occur (and did occur in NPP) in BWR core (D’Auria et al., 1997b, see also Chapter 16 and the discussion about LaSalle-2 and Oskarshamn-2 events). H. Example 2 for accident scenario, MSIV closure in BWR. Void collapses in the core happen following pressure wave generated at the closing of the valves. The accident scenario is qualitatively similar to the turbine trip event (discussed in Section 15.4.12.10, see also row 44 in Table 15.1), this one mitigated by the quick opening of the condenser dump valves. I. Example 3 for accident scenario, boron dilution in PWR. A boron diluted plug may form in certain SBLOCA transients, namely during the NC mode called reflux condenser (Section 15.4.12.4). Reactivity excursion may occur when the boron diluted liquid is transported into the core although mixing with borated fluid in the RPV may occur, as discussed in Section 15.4.12.8 (row 4 in Table 15.1). J. Example 4 for accident scenario, MSLB in PWR. Cold liquid plug may be generated following the fast depressurization of one SG, inducing positive reactivity excursion in the core region, as discussed in Section 15.4.8 (rows 10, 26, and 27 in Table 15.1). K. Example 5 for accident scenario, FP excursion following LOCA in PHWR, CANDU, and RBMK. Following an LOCA, an FP excursion may take place when (slight) positive void

Accident scenarios and phenomena in WCNR

1031

Normalized power (−)

1.5 1.25 1 0.75 0.5

HZP (avg) HFP (avg)

0.25 0 0

0.5

1

1.5 2 2.5 Core height (m)

3

3.5

Fig. 15.30 RIA background analysis for a core of a PWR. Left: differences in radial/azimuthal core power calculated by two methods. Right: axial peaking factors at HZP and HFP. reactivity is tolerated in the core design. This is discussed in relation to CANDU by Rouben (1997) (row 11 in Table 15.1), in relation to PHWR by D’Auria et al. (2008a) (row 17 in Table 15.1), for details see Pecchia et al. (2011), and in relation to RBMK, by D’Auria et al. (2005) (row 13 in Table 15.1), see also D’Auria et al. (2008d). L. Example 6 for accident scenario, CRE in any reactor design. Control Rod Ejection is a key event to demonstrate the safety of NPP and is discussed later.

Pioneering research to set-up the bases for transient three-dimensional coupled thermal-hydraulic and neutron kinetics calculations has been carried out at Pennsylvania State University (e.g., Ivanov and Baratta, 1999). An idea of the performed activity can be derived from Fig. 15.30, including the following diagrams: (a) power differences calculated in each FA by two methods; (b) axial power distribution at HZP; and (c) axial power distribution for the same core at HFP. The availability of qualified distributions for concerned quantities is a prerequisite to perform transient RIA analyses.

15.4.9.2 Quantitative accident scenario: variable trends – CRE CRE scenario is triggered by a failure in control rod drive mechanisms or in connected hardware. The ejection of one rod occurs in time period less than 1 s and the event is terminated by scram a few seconds after the triggering (TSE table not reported). Typical CRE scenario can be derived from Figs. 15.30 and 15.31, CRE in PWR with UTSG (HZP and HFP conditions), taken from Todorova and Ivanov (2003) (row 8 in Table 15.1). Issues to be addressed from the (quantitative) analysis of the CRE are: l

l

Generation of local voids in transient subcooled boiling HT mode (e.g., D’Auria et al., 2004). Example of subcooled void fraction time history can be found in Fig. 15.28. Generation of pressure pulse in the RPV. This can be “absorbed” in PRZ, provided a suitable flow area is available in the surgeline (D’Auria and Galassi, 2004).

1032

Thermal Hydraulics in Water-Cooled Nuclear Reactors

4000 3500

570 Total power (MW)

3000 2500

565

2000 560

1500 1000

555

Power (avg) - HFP transient

500

Core average fuel temperature (K)

575 B-4-EV2 S-21-GM1

Temperature (avg) - HZP transient

550

0 0

1

2

3

4

5

6

Time (s)

Fig. 15.31 CRE in PWR with UTSG and phenomena characterization: core power and fuel average temperature.

l

l

l

Simultaneous (or delayed) occurrence of CRE and SBLOCA with break located in the upper head of the RPV, consequence of rod ejection. Break opening makes milder the pressure pulse in RPV. Possible occurrence of transient CHF and consequent RST excursion and possible change in fuel thermal properties (also mentioned in “background issue 6,” earlier) (D’Auria et al., 2004). Local feedback (around the ejected CR) of transient values for coolant temperature, density, and void and nuclear fuel parameter values like average temperature also affected by gap and pellet conductance and by possible changes in geometric configuration.

15.4.9.3 Phenomena connection with accident scenario: CRE The phenomena listed in the second and the third columns of Table 15.20 are connected with MSLB and FWLB from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables in Figs. 15.26–15.29 and with Figs. 15.68 and 15.14 taken from other sections of this chapter, as reported in the fourth column of Table 15.19. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Evaporation due to heat input (indirectly visualized, Fig. 15.31; visualized, Fig. 15.28). The basic phenomenon occurs in all scenarios where void is produced in the core region. Subcooled voids are generated following CRE and constitute an example for connecting accident scenario and phenomenon.

Accident scenarios and phenomena in WCNR

1033

Table 15.20 Phenomena visualized by variables representative of the accident scenario CRE Phenomenon ID for: CRE No.

Acronym

Description

Fig. no.

Notes

1

B-4-EV2

15.31

2

S-21-GM1

See also detail in Fig. 15.28 See also Fig. 15.30

3

S-44-PCEI

Evaporation due to heat input Global multi-D fluid temperature, void, and flow distribution—core Parallel channel effects and instabilities PCEI

15.31

16.3A and B*

*Chapter 16, Section 16.4.3

(2) Global multi-D fluid, temperature, void, and flow distribution—Core (indirectly visualized, Fig. 15.31). The consideration of three-dimensional core performance is essential in the case of CRE, as discussed earlier. The time histories shown in Fig. 15.31 constitute the result of a calculation where global multi-D fluid, temperature, void, and flow distribution are taken into account. (3) Parallel channel effects and instabilities PCEI (visualized, Fig. 16.3A and B in Section 16.4.3 of Chapter 16). The instability in BWR is an important topic in nuclear thermal-hydraulics (i.e., deserving a specific section like Boron, PTS, etc.). This is discussed in Chapter 16, including NPP data and proper literature references.

15.4.10 The ATWS ATWS is an NPP system condition following an assigned PIE where scram is not fully actuated following the occurrence of related signals, that is, failure of scram on demand (a more comprehensive definition of ATWS can be found elsewhere). ATWS may be considered as an attribute of selected accident scenarios. For instance, the following transients can be defined: MCP-trip/ATWS, CRE/ATWS, SGTR/ATWS, MSLB/ATWS, and SBLOCA/ATWS. The common practice in NRS is to associate (relatively) “high” probability transient scenarios with the ATWS condition. Therefore, one should not find realistic analyses for LBLOCA/ATWS or LOOSP/ATWS. The key feature of any ATWS is the interaction between thermal-hydraulics and neutron physics parameters: coupled three-dimensional modeling and calculations are needed.

15.4.10.1 Qualitative accident scenario: ATWS Basic ATWS topics are: A. Core operation in nominal conditions is typically associated with the best configuration for total reactivity (here intended as the capacity which ensures the highest thermal neutron flux). This is true when the same fluid has both the functions of coolant and moderator; i.e., the statement does not apply in the case of CANDU, PHWR, and RBMK. The same

1034

B.

C.

D. E.

Thermal Hydraulics in Water-Cooled Nuclear Reactors

statement applies only partially to BWR and may not apply even in PWR when the boron concentration is very high. The connection between initial statement and ATWS is as follows: the ATWS implies a starting PIE, thus the perturbed system may not be in the best configuration; consequently, core power decrease following ATWS is expected. Unluckily, other system parameters may also not have their most effective value as far as core power removal function is concerned (e.g., low coolant flow following MCP-trip or low heat sink capability following LOFW). Thus ATWS puts a challenge to the integrity of the system. The overall capability of relief valves in PS and SS (including in the latter case the condenser dump) is designed to remove maximum thermal power produced in the core. Therefore, the opening of relief valves ensures protection for the structural integrity of the RCS A specific issue associated with ATWS in the case of BWR consists in the overheating of the PSP water. Following an assigned PIE and the ATWS condition, MSIV closure occurs and SRV operate within design conditions: the issue arises because SRV discharge the two-phase fluid into the pool and the same pool is used as suction from ECCS pumps. Average heating of the pool above a given threshold (typically 80°C) induces problems for the availability of pumps. Then (only) a few dozen seconds are allowed for the operation of SRV with core at nearly nominal power. Doppler effect or total power reactivity coefficients constitute intrinsic design features mitigating ATWS scenario. Additional (independent) scram including devoted high-pressure boron tank constitute typical ESF to protect or to mitigate the risk associated with ATWS.

Notwithstanding the availability of ATWS-specific ESF, ATWS analyses are performed and may be requested within the licensing framework.

15.4.10.2 Quantitative accident scenario: variable trends and TSE—ATWS An overview of typical ATWS scenarios can be derived from the TSE in Table 15.21, and Fig. 15.32, ABWR, taken from Ferng et al. (2010) (row 1 in Table 15.1) and Fig. 15.33, ABWR, taken from Huang et al. (2007) (row G in Table 15.5). Table 15.21 Typical ATWS in water cooled nuclear reactor: (selected) imposed and calculated events N

Event

I/R

Time (s)

Notes

1

Assigned PIE occurrence

I

–to*

2 3 4

Scram signal generated PORV/SRV first opening MCP-trip

I R I

0 10 15

5

Independent scram actuation

I

30

*to can be any time before “0,” typically less than few tens seconds Scram does not occur At the end of MCP coast-down If beneficial (e.g., case of BWR and PHWR) For example, boron injection (occurring at 12,500 s in Fig. 15.32)

Accident scenarios and phenomena in WCNR

Core power (%)

100

16

1035

Core Power w/o EOP I-16-NTF1

80 60 40 20

14

0 0

14

2500

5000 7500 Time (s)

10,000

12,500

Pressure w/o EOP Level w/o EOP 13 Level with EOP

11 10

10

8

9

6

8 7

I-6-CLDO I-16-NTF1 S-47-PS2

4

RPV water level (m)

RPV pressure (MPa)

12 12

6

2

5

0 0

2500

5000

7500

10000 Time (s)

12500

15000

17500

4 20000

Fig. 15.32 ATWS in ABWR and phenomena characterization: RPV pressure and level. Detail core power.

Table 15.21 does not refer to any specific ATWS. Rather it provides an idea for timings expected for ATWS scenarios. The data in Fig. 15.32 (not resulting from a three-dimensional coupled neutron physics thermal-hydraulic analysis) show ATWS duration of the order of 3 h, although core power is lower than the nominal value. One purpose for the concerned calculation is to determine the time available to operator to delay the boron injection. The data in Fig. 15.33 are related to what has been called an ATWS-similar accident scenario. The reason is that scram occurs about 10 s after the stop in FW flow (detail in Fig. 15.33). The motivation for Fig. 15.33 is the availability of curves showing different performances for RPV internal pumps.

15.4.10.3 Phenomena connection with accident scenario: ATWS The phenomena listed in the second and the third columns of Table 15.22 are connected with LOOSP from the cross-link process in Tables 15.3–15.5. Those phenomena are associated with variables in Figs. 15.32 and 15.33, as reported in the fourth column of Table 15.22. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible):

1036

Thermal Hydraulics in Water-Cooled Nuclear Reactors

2

160 S-30-IPU S-55-SDR

140

80 140 Percent of rated (%)

Flow rate (%)

100

60 40 20

1

0.5

Feedwater flow Neutron flux

105

1.5

Level (m - Sep - Skirt)

Core inlet flow Pump flow Gr 1 Pump flow Gr 2 Pump flow Gr 3 Level NR level WR level

120

70

0

35 0 0

5

10

15

20 25 Time (s)

30

35

40

−0.5

0 0

5

10

15

20 Time (s)

25

30

35

40

Fig. 15.33 ATWS type of condition in ABWR and phenomena characterization: RPV levels and flow from different groups of internal pumps. Detail: core power and FW flow.

Table 15.22 Phenomena visualized by variables representative of the accident scenario ATWS Phenomenon ID for: ATWS No.

Acronym

Description

Fig. no.

Notes

1

I-6-CLDO

Collapsed-level behavior in downcomer

15.32

See also phenomenon at row 4 below

2

S-30-IPU

15.33

3

I-16-NTF1

4

S-47-PS2

5

S-55-SDR

Internal pump behavior (specific geometry) Nuclear thermal-hydraulics feedback and spatial effect Phase separation/vertical flow with and without mixture level—Downcomer Steam dryer behavior

15.32* 15.32

15.33

Also in detail. *See also I-29-NTF2 See also Fig. 15.13

Accident scenarios and phenomena in WCNR

1037

(1) Collapsed-level behavior in downcomer (visualized, Fig. 15.32). Collapsed liquid level typically occurs in conditions where pressure is increasing. The prediction of the level is important because connected with actuation of various NPP signals (2) Internal pump behavior (visualized Fig. 15.33). Different internal pumps may rotate at different speeds inducing further three-dimensional phenomena inside RPV. (3) Nuclear thermal-hydraulics feedback and spatial effect (visualized, Fig. 15.32). Nuclear thermal-hydraulic feedback can be observed from any ATWS calculation-based analysis. Spatial effects visualization need a three-dimensional calculation (e.g., by Kliem et al., 2009, at row 3 in Table 15.1). (4) Phase separation/vertical flow with and without mixture level—Downcomer (visualized Fig. 15.32). Phase separation is connected with (mixture and collapsed) level formation and entrainment-de-entrainment processes. Mixture-level formation is properly visualized from results of LOCA analyses (e.g., Fig. 15.13). (5) Steam dryer behavior (indirectly visualized, Fig. 15.33). Steam dryer performance is affected by level formation, specifically when vaporization processes occur owing to depressurization. Steam dryer (and separator) behavior affects the formation of level in downcomer due to pressure drops occurring inside. Steam dryer integrity is challenged during accidents like MSLB, as discussed by Yan and Bolger (2009).

15.4.11 The AM procedures and selected related accident scenarios Accident management and AMPs constitute important fields for the application of nuclear thermal-hydraulics. This shall be seen as the key motivation for the present section. Basically, AM and AMP can be characterized by the following two statements: (a) they come into play during an accident in NPP following unexpected events or failures of systems and components including multiple failures of ESF; (b) they aim at ending further progression of any accident condition throughout prevention and mitigation actions, by adopting any resource available at NPP (including human resources), minimizing at the design-level (case of AMP) modifications needed to the hardware of existing NPP. Historically: l

l

l

l

AM studies started after the TMI-2 event in 1979 (e.g., Petrangeli et al., 1993; D’Auria et al., 1997a). AMP have been interpreted as time follow-up of EOP, although AMP and EOP constitute integrated set of actions. Within the context at the item above, AMP are associated with BDBA and EOP with DBA in licensing nuclear safety technology. Two areas have been distinguished for AM and AMP: out of DBA envelope but before core degradation and after core degradation. In the latter case the term SAMP is introduced.

In the present section only AM and AMP dealing with the “before-core-degradation” conditions are considered.

1038

Thermal Hydraulics in Water-Cooled Nuclear Reactors

15.4.11.1 Qualitative accident scenarios: needing AM or following AM actions The actuation of AMP is part of accident scenarios discussed in other sections of this document (see, e.g., LOOSP, Fig. 15.19 and SGTR with consequential MSLB, Fig. 15.53). Namely, the reference databases for AM in Table 15.1 include rows 12, 31, 34, and 40: related sections (other than the present one) deal with AM information. An idea of AM actions and issues having connection with nuclear thermalhydraulics can be drawn from the following statements: A. PS depressurization (e.g., Petrangeli et al., 1993). This is common practice in BWR and part of design-related EOP: e.g., any SBLOCA scenario may end-up with opening of ADS valves allowing the intervention of LPCI and LPCS and the recovery of the system. The same action may reveal troublesome in PWR because PS depressurization through RPV is not feasible and depressurization through PRZ causes the coolant accumulation in the PRZ owing to the CCFL occurrence in the surgeline. B. SS depressurization (e.g., Madeira et al., 2003). This is an effective AM for PWR provided it is performed under proper RCS conditions: (i) depressurization of a steam filled SG may not bring-down the PS pressure owing to ineffective heat transfer across U-tubes and it is only effective if subsequent SG flooding is possible; (ii) depressurization of a liquid full SG in case PS is empty is also not effective owing to ineffective heat transfer across U-tubes and, in addition, may take long time not consistent with needs from core cooling (i.e., to prevent or to mitigate RST excursions). C. Combination of PS and SS depressurizations (can be called consequential RCS depressurization) (e.g., Muellner et al., 2007). This may reveal the best action to prevent accident progression and to avoid core damage. The use of consequential depressurization stages combined with passive cooling of RCS is at the basis of the safety technology for the AP1000 reactor. D. Coolant delivery from FW tank (e.g., Madeira et al., 2003). In all NPP as part of BOP, an FW tank, also called degasser-condenser, is installed downstream the main condenser and upstream the main feedwater pumps. The tank (or the group of tanks) has a volume of the order of 1000 m3 (in a 1000 MWe unit). The pressure is kept, during NPP operation at nominal power, at around 0.5 MPa and slightly subcooled liquid is present. Feeding the FW tank into a depressurized RCS (either PS or SS) reveals effective in removing decay power for a few hour without the need of external power. E. Core Exit Thermocouples (CET). CET signal may be necessary to trigger the actuation of AMP (Toth et al., 2010): 400–600°C is the typical range of CET signals for entering the BDBA virtual region and start of the AM devoted systems. The (thermal-hydraulics) issue is that superheated steam temperature at core outlet is not easily or timely detectable owing to the influence of thermal inertia of solids present in the UP (e.g., the CRGT) and of phenomena like de-entrainment and deposition of droplets which may overshadow the fluid temperature at the location of the thermocouple. Refinement of measurement techniques is in progress allowing a straightforward connection between RST at PCT location and CET signals.

Complex AMP involving action with large asymmetries among performances or conditions of loops (PS and SS) have recently been the topic of experimental programs (see e.g., Mull et al., 2007).

Accident scenarios and phenomena in WCNR

1039

15.4.11.2 Quantitative accident scenarios: needing AM or following AM actions No typical TSE can be defined for an AMP or for an accident scenario where AM occurs. However, AM and AMP imply importance for the human (operator) actions; therefore time duration for accident scenarios involving AM are expected in the range of several minutes or hours rather than seconds. Typical results from thermal-hydraulics analyses of transient with AMP can be derived from Fig. 15.34 dealing with CANDU, taken from Mehedinteanu (2009) (row 12 in Table 15.1), and Fig. 15.35 dealing with VVER-1000 taken from D’Auria et al. (2006) (e.g., row 40 in Table 15.1). In the case of CANDU, Fig. 15.34, the AM based upon an innovative system allows the prevention of core damage following scenarios which in the absence of such a system evolve into severe accidents. The concerned procedure involves: (a) the PS depressurization through the pressurizer with the two-phase mixture discharge into the degasser-condenser; (b) the subsequent “re-pumping” of the water into the RCS by the pressure difference intrinsically created. A study was performed to confirm the applicability of AMP developed for PWR with UTSG to the VVER-1000 (Fig. 15.35). Among various findings, it was demonstrated that core integrity (so-called grace-period) in case of LOOSP may be shifted for the Balakovo-3 Unit from less than 3 h (and possible RPV failure at high pressure) to more than 12 h (and possible RPV failure at low pressure) (see also Cherubini et al., 2008b). No major modification for the NPP is needed: the combination of PS and SS depressurizations allows the FW tank liquid to enter the SG keeping cooled the core. The execution of the VVER-1000 AM) study also involved an experimental program (Bucalossi et al., 2012) and contributed to the creation of a database for code validation and uncertainty in system thermal-hydraulics (Petruzzi and D’Auria, 2016). 12 10

2000

Temperature

1600 8 1200 6 800

4

400

2 0 0

2200

4400

6600

8800

0 11,000

Time (s)

Fig. 15.34 Accident scenario with AM action in CANDU: PS pressure and RST.

Rod surface temperature (K)

Primary system pressure (MPa)

Pressure S-28-HT3 I-12-NC3

1040

Thermal Hydraulics in Water-Cooled Nuclear Reactors 19 18 17 16 15

Failure pressure (MPa)

14 13 12 11 10 9 8 7 6 5 4 3 2 1 0

5000 10000 15000 20000 25000 30000 35000 40000 45000 50000 55000 60000 65000 Failure time (s)

Fig. 15.35 Accident scenario with AM action in VVER-1000: reactor failure map following different AM actions.

15.4.11.3 Phenomena connection with accident scenarios: needing AM or following AM actions The phenomena listed in the second and the third columns of Table 15.23 are connected with LOOSP from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables in Fig. 15.34 as reported in the fourth column of Table 15.23. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): Table 15.23 Phenomena visualized by variables representative of accident scenario involving actuation of AMP Phenomenon ID for: scenarios needing AM actions No.

Acronym

Description

Fig. no.

1

S-28-HT3

15.34

2

I-12-NC3

HT [condensation]—SG structures NC core bypass, hot and cold bundles

15.34

Notes

BWR, CANDU, and RBMK

Accident scenarios and phenomena in WCNR

1041

(1) HT [condensation]—SG structures (indirectly visualized, Fig. 15.34). The HT with condensation in SG occurs during various accident scenarios. In case of the concerned AM-related scenarios, the PS depressurization causes temperature decrease in PS which brings steam condensation in the SS. (2) NC core bypass, hot and cold bundles (indirectly visualized Fig. 15.34). Different channels in the core need to be modeled in a variety of accident scenarios for reactors like BWR, CANDU, and RBMK. In the case of long-lasting accident scenarios relevant to AM, different flow conditions occur in various channels typically driven by NC.

15.4.12 Additional accident scenarios and related topics Accident scenarios in addition to those considered in Sections 15.4.2–15.4.11 are analyzed here together with related topics, with the objective to complete the characterization of phenomena listed in Table 15.2. The words “related topics” include subjects of special interest to thermal-hydraulics which constitute phenomena, phenomenological windows or system, and components, part of accident scenarios. Six topics are considered: (a) Critical Heat Flux (CHF) in Section 15.4.12.1. CHF is also part of phenomenon S-26-HT1 in Table 15.2 and is discussed with suitable detail in Chapter 7. (b) Containment performance in Section 15.4.12.2. Containment is a key safety feature for NPP. The system performance is at the basis of phenomena I-18-PRB and S-42-NCOC in Table 15.2. (c) PTS in Section 15.4.12.3. PTS involves the integrity of the RPV (in water cooled reactor equipped with RPV) and implies an integrated analysis by thermal-hydraulics, structural mechanics and material damage by neutron fluence. Only thermal-hydraulics aspects are considered here. PTS-related phenomena are S-22-GM2, S-34-LVM2, and S-35-LVM3 (in the last case in relation to CL) in Table 15.2. (d) Natural Circulation (NC) in Section 15.4.12.4. NC constitutes a PHW part of different accident scenarios where pump trip occurs. It is discussed by phenomena I-10-NC1 to I-15-NC6 and A-11-NC in Table 15.2. (e) Boron dilution in Section 15.4.12.8. Boron mixing, transport, and stratification are considered by phenomena S-2-BO and A-12-BO in Table 15.2. In Section 15.4.12.8, attention is given to the boron dilution event, part of the boron transport. Consequences of boron dilution upon the fission power are mitigated by boron mixing. (f ) Nuclear fuel in Section 15.4.12.12. Nuclear fuel constitutes the motivation for safety analyses. The words “nuclear fuel” include the fuel rods, the Fuel Assemblies (FA), and the overall core. Thermal- or thermodynamics-related properties, like conductivity and heat capacity, gap conductance, quantities connected with reactor physics like enrichment, composition (e.g., MOX), and neutron cross-sections, material properties and related interactions like density and radiation-induced swelling, structural mechanics effects like ballooning, and system effects like burn-up, crud and oxide formation are of interest and have a connection with thermal-hydraulics.

The description of accident scenarios and related topics is provided according to the structure proposed in Section 15.4.1. Furthermore, AM could also have been part of this list. However, overall accident scenarios are involved. Thus AM is discussed in Section 15.4.11.

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

15.4.12.1 Calculation of CHF aimed at core design optimization CHF is at the center of the attention in Chapter 7 and in several accident scenarios, see also the report (IAEA, 2001). Moreover, heat transfer and CHF are visualized in a number of diagrams in this chapter (e.g., Figs. 15.1, 15.9, 15.13, and 15.22). The purpose here is not to duplicate any discussion, rather, to confirm the relevance of the CHF phenomenon in nuclear thermal-hydraulics.

CHF topics and facts Selected key topics associated with CHF phenomenon and CHF evaluation are: l

l

l

l

l

CHF value is one key parameter at the basis of the design, that is, the diameter and the length, of fuel rods part of the core of water cooled reactors. Maximum linear power (q0 max), then attainable core power, and margin to dry-out are affected by CHF. Gathering scale-one experimental data for fuel assembly constitute the best practice to determine q0 max and margin to dry-out. Therefore any nuclear fuel vendor/ designer has access to an SETF to optimize parameters like number and configuration of spacer grids (specific CHF grids may be added to increase CHF at nominal operating conditions, as well as CHF expected following a transient), rod-to-rod distance and fluid velocities. CHF occurrence affects PCT value and related time following LBLOCA. In this connection, transient CHF value may differ from CHF determined under steady-state conditions. Complex phenomena and occurrences, not necessarily part of thermal-hydraulics, affect the CHF values like burn-up and associated swelling for the nuclear fuel pellets with changes in fuel rod gap size, ballooning, crud deposit, oxidation, etc. It is difficult to ensure the prediction of CHF value with an error less than 10% (or, any CHF prediction is affected by an error of 10% or greater).

Phenomena connection with accident scenario: CHF CHF phenomenon-related information can be derived from Fig. 15.36 taken from PWR-core basic experiment (Visentini et al., 2014) (row 5 in Table 15.1), including the transient CHF formula (right side of the diagram) proposed by Sakurai et al. (1990). The phenomenon identified in the second and the third columns of Table 15.24 is connected with CHF from the cross-link process in Tables 15.3 and 15.4. The phenomenon is associated with variables in Fig. 15.36. Only one phenomenon is outlined in this section and visually described through the use of experimental data (time trends) in basic facilities: (1) HT [NCO, FCO, SNB, SANB, CHF/DNB, post-CHF]—Core, SG, structures (partly visualized, Fig. 15.36). The phenomenon deals with convection heat transfer and, namely, with boiling heat transfer. The phenomenon constitutes the subject for Chapter 7 and is visualized by a number of time trends related to accident scenarios, as already mentioned. In the paper by Visentini et al. (2014), several heat transfer regimes are considered, too. Noticeably, in the case of post-CHF (or film boiling conditions), radiation hat transfer may play an important role.

Accident scenarios and phenomena in WCNR

1043

100

1000

4

750

3

TW – TL

50

500

25

250

Heat flux (W/m2) × 105

TW –TL (K)

S-26-HT1

Power (W)

Power

75

2

qCHF = qready [1 + 0.21τ –0.5]

1

PT Sat 0

0 0

2.25

4.5

6.75

9

0 10

20

Time (s)

30 40 TW –TCHF (K)

50

60

Fig. 15.36 CHF in basic test facility: transient conditions. Left: wall temperature and imposed power. Right: heat flux and wall temperature in nucleate boiling and film boiling (the legend “PT Sat” implies data in saturation conditions).

Table 15.24

Visualization of CHF phenomenon Phenomenon: CHF

No.

Acronym

Description

Fig. no.

1

S-26-HT1

HT [NCO, FCO, SNB, SANB, CHF/DNB, post-CHF]—Core, SG, structures

15.36

Notes

15.4.12.2 Containment performance As in the case of CHF, containment is at the center of the attention in nuclear reactor safety (although no specific chapter of the book is devoted to containment), and in several accident scenarios, see also the reports (OECD/NEA/CSNI, 1989a, 1999, 2014; IAEA, 2009a). Moreover, containment-related phenomena are visualized in a number of diagrams in this chapter (e.g., Figs. 15.3 and 15.8). The purpose here is not to duplicate any discussion, rather, to confirm the relevance of the containment phenomena in nuclear thermal-hydraulics.

Containment topics and facts Selected key topics associated with containment performance and containment phenomena are: l

l

Containment is at the same time a dynamic system, a safety component, and a barrier to protect the environment and to mitigate the consequences of accidents. Containment can also be defined a passive system. Classification of containment systems is beyond the purpose for the present document. However, one may briefly recall the existence of full pressure and PSP or BC type containments

1044

l

l

l

l

Thermal Hydraulics in Water-Cooled Nuclear Reactors

(see e.g., Blinkov et al., 2012), of vented containments also called confinements, namely in the case of VVER-440 and RBMK NPP. Containment systems equipped with PSP and BC are used in BWR and VVER-440, respectively. In both cases, the objective is to reduce the maximum pressure which builds-up in the containment following an LOCA with an assigned (containment) building volume. Otherwise, the presence of BC and PSP is at the origin of possible oscillations and consequent loads upon structures caused by steam condensation processes. The importance of containment venting (venting is available in VVER-440 and RBMK NPP) resulted from the Fukushima event, see also Chapter 16. Containment systems are expected to withstand mechanical loads originated by pipe-whip, jet thrust, jet impingement, H2 burning and deflagration, other than by pressurization, temperature gradients in the structures and by steam condensation into liquid pools. In all cases thermal-hydraulic analyses are necessary. A variety of thermal-hydraulic phenomena are expected to occur in containment systems during accidents in water cooled nuclear reactors. Examples of key phenomena are, in addition to steam condensation in PSP and BC subsystems, film-wise and drop-wise condensation in structures, natural convection in wide spaces, temperature stratification in pools and in steam-gas spaces, H2 transport, and formation of liquid pools (e.g., in the sump). Furthermore the natural convection motions are also governed by pressure drops at geometric discontinuities having different three-dimensional shapes.

Phenomena connection with accident scenario: Containment The containment-related phenomena for water cooled nuclear reactors considered in the present section can be derived from Fig. 15.37 taken from ABWR calculation (Chen and Yuann, 2015) (row 7 in Table 15.1) with the addition of the upper detail dealing with the check valve performance, taken from Hammersley et al. (2000) (row A in Table 15.5), Fig. 15.38, dealing with density lock performance, taken from Tasaka et al. (1992) (row B in Table 15.5), and Fig. 15.39, taken from AP1000 calculation (Hung et al., 2015) (row 6 in Table 15.1). Related to the ABWR calculation in Fig. 15.37, the WDVB leakage is modeled by a flow path connecting the lower dry-well and the wet-well airspace. There are eight WDVB installed in the containment. The key result is that three WDVB are enough to provide sufficient venting area to mitigate the negative pressure difference (allowed design threshold is 13.7 kPa); otherwise, if only one WDVB is operable, the peak value in the case of steam line accident (not shown in the figure) is expected to be greater than the design value. The performance of an individual check valve in case of a separate effect experiment can be deduced from the upper detail. A 200 check valve having 40 ms closing time is installed downstream a pressure vessel (initial pressure set at 12 MPa). The measured pressure response at a position downstream the valve is given in the figure. The peak observed pressure following check valve closure is seen to be 29 MPa. In the case of PIUS (Fig. 15.38), the primary loop is connected to a neutron poison pool (borated water) through a lower and a upper vertical section, termed density lock, in which stable density stratification is established during power operation. The penetration of the poison water into the primary loop is prevented by keeping a balance

Accident scenarios and phenomena in WCNR 30

14 Pressure (MPa)

Test pressure

12 Pressure difference (kPa)

1045

10

25 20

A-1-CHV

15 10 5 0

8

0

0.02

0.04

0.06

0.08

0.1

Time (s)

A-1-CHV I-18-PRB

6

Press. Diff. 1 Press. Diff. 2

4 S-12-CO4, Section 15.4.12.10

2 0 1

2

3

4 Nr. of WDVB (–)

5

6

7

Fig. 15.37 ABWR containment performance following FWLB: dry-well/wet-well pressure difference with various operable WDVB. The upper detail shows the pressure differential across a check valve during an SETF experiment.

1 Pump speed (Hz) – power (kW)

35

A-4-DL

30

0.8 0.6

Prim pump speed control Core power (7.5 kW max) Diff. press. across lower H Comb Flow rate through lower H Comb

25

0.4 0.2 0

20

–0.2 15

–0.4 –0.6

10

–0.8 5

Diff. pressure (Pa) – flow rate (kg/s)

1.2

40

–1

0 0

300

600

900 1200 Time (s)

1500

1800

–1.2 2100

Fig. 15.38 Experimental simulation of PIUS-related density lock behavior: various system parameters during a start-up transient.

1046

Thermal Hydraulics in Water-Cooled Nuclear Reactors

450

31

400

Pressure (kPa)

Pressure Level

300

29

250 28 200

A-2-CC S-42-NCOC I-18-PRB

150

27

Water level (m)

30 350

100 26 50 0 0

2000

4000

6000

8000

25 10,000

Time (s)

Fig. 15.39 DEHL break in AP1000: containment pressure and sump level.

between the pool static head and the loop differential pressure between the two density locks. Any perturbation in the primary loop causes inlet of borated liquid. The feasibility design of the system implies the demonstration of start-up process without inlet of borated water into the loop. The feasibility of the density lock-based system is confirmed from experimental data (e.g., time trends reproduced in Fig. 15.38). Typical time trends for pressure and liquid level for containment during a DEHL break in AP1000 are reported in Fig. 15.39. The formation of liquid pool in the sump allows long-term core cooling for any PWR. The phenomena listed in the second and the third columns of Table 15.25 are connected with containment performance from the cross-link process in Tables 15.3–15.5. Those phenomena are associated with variables given in Figs. 15.49–15.51 in the fourth column of the table. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Behavior of check valves (visualized, Fig. 15.37). Check valves can be associated with different systems in an NPP. The association of check valves and BWR containment derives from the available paper (i.e., row 6 in Table 15.1). It is shown that check valves (smooth) operation controls the differential pressure loading upon internal structures. In addition (detail in the figure) large pressure waves can be expected from the dynamic operation (e.g., fast closure) of check valves. (2) Behavior of containment emergency systems, PCCS (indirectly visualized, Fig. 15.39). PCCS affects the containment pressure in AP1000. (3) Behavior of density locks (visualized, Fig. 15.38). Density locks shall be considered as a historic device (giving origin to the phenomenon in concerned documents). Density locks

Accident scenarios and phenomena in WCNR

1047

Table 15.25 Phenomena visualized by variables representative of the containment performance Phenomenon ID for: Containment No.

Acronym

Description

Fig. no.

Notes

1 2

A-1-CHV A-2-CC

15.37* 15.39

*Including detail

3 4

A-4-DL S-42-NCOC

15.38 15.39

PIUS reactor design

5

I-18-PRB

Behavior of check valves Behavior of containment emergency systems (PCCS) Behavior of density locks Natural convection and H2 distribution Pressure-temperature increase and boiling due to energy and mass input

15.37 and 15.39

PD dry-well to wet-well affected by check valves, item (1)

were proposed for the PIUS reactor concept after the Chernobyl accident, and experiments were performed in Japan. Density locks may ensure what was called “intrinsic” safety: any operation of the primary system outside the pressure range imposed by the height of the honeycomb (i.e., the physical component/structure which forms the density lock) causes highly borated liquid inlet into the core region and stop of the fission reaction. The PIUS concept was abandoned before the end of the past century, and density lock thermal-hydraulic performance is documented in the cited reference. (4) Natural convection and H2 distribution (indirectly visualized, Fig. 15.39). NCO and H2 distribution contribute to the containment pressure. Direct visualization would imply the availability of local values of fluid velocities or concentration. (5) Pressure-temperature increase and boiling due to energy and mass input (visualized, Fig. 15.37). Containment pressure and pressure differentials between compartments are affected by energy and mass exchanges. The concerned time trends deal with differential pressure build-up originated by steam flows and condensation in the PSP of an ABWR.

15.4.12.3 Pressurized thermal shock (PTS) PTS topics and facts An idea of thermal-hydraulic aspects which characterize Pressurized Thermal Shock can be derived from Fig. 15.40 (bases for the work by Araneo et al., 2011), Fig. 15.41 (Araneo et al., 2011; Kinoshita et al., 2012), and Fig. 15.42 (Araneo et al., 2011). Selected relevant PTS topics are outlined hereafter: A. PTS constitutes a possible mechanism which causes the loss of integrity of RPV. As already stated the combination of pressure, temperature, and neutron fluence loads may trigger the propagation of a microfracture inside the RPV leading to its rupture in a brittle mode. Zones of the RPV close to the irradiation region (the core), affected by flaws (sometimes, unknown, or undetected) greater than a critical value are prone to PTS. Welding regions with impurities at an elevation close to TAF possible origin of PTS concern.

1048

Thermal Hydraulics in Water-Cooled Nuclear Reactors

16.0

DBA trajectories

14.0

UP pressure [MPa]

12.0 P-T Limit

10.0 8.0 6.0 4.0

Saturation pressure curve

2.0 0.0 300.0

350.0

400.0 450.0 500.0 CL1 fluid temperature [K]

550.0

600.0

Fig. 15.40 PTS thermal-hydraulic aspects: identification of the worst transient based on trajectories in the phase space and limit thresholds at system level.

Fig. 15.41 PTS thermal-hydraulic aspects which need CFD. Left: results from liquid mixing for the estimation of PTS; Right: liquid (red) and air (blue) interacting in the HL (CCFL numerical simulation). B. PTS may occur in the presence of single-phase, noticeably MSLB, and two-phase conditions, noticeably SBLOCA or LBLOCA. In the case of PTS occurring during LBLOCA pressure may have a small role and key triggering parameters for stresses the temperature gradient into the walls of the RPV. C. PTS is of low interest in the case of BWR because of the large downcomer and the low fluence expected for the vessel. D. The capability to calculate locally liquid-liquid mixing (e.g., liquid streams having different temperatures and velocities) and liquid-gas interactions (e.g., including mixing and phase

Accident scenarios and phenomena in WCNR

1049

4000 t = 200 s

3500

t = 150 s

(N/mm^1.5)

3000 t = 250 s

2500

t = 300 s

2000

t = 400 s

t = 100 s

1500

t = 50 s

1000

Klc (RTNDT = 50° C) 7 mm

500 0 0

50

100

150

200

250

300

Temperature (°C)

Fig. 15.42 PTS thermal-hydraulic aspects: calculation of safety margin for RPV considering limit thresholds at local level. changes in liquid and gas streams flowing cocurrently and counter-currently) is needed for a realistic estimation of the PTS loads (see the color pictures in Fig. 15.41). E. Various procedures are available from the literature to evaluate the PTS loads and the risk. Those procedures typically include several steps (e.g., Araneo et al., 2011; see also the paper at row 28 in Table 15.1; Jang, 2007). Three of steps directly connected with thermalhydraulics are outlined with the support of Figs. 15.40–15.42: - Search for the worst DBA transient (Fig. 15.40). - Evaluation of local fluid temperature (Fig. 15.41). This needs the application of CFD codes (discussed in Chapter 12). - Evaluation of the safety margins quantified by the minimum distance between the trajectory corresponding to the worst accident scenarios and a limit curve derived from material and stress status of the vessel (Fig. 15.42).

Phenomena connection with accident scenario: PTS The PTS scenario for a PWR equipped with UTSG can be derived from Figs. 15.43 and 15.44 taken from SGTR, MSLB, and SBLOCA calculations (Jang, 2007; rows 28, 39, and 42 in Table 15.1). The three typical transients, possibly at the origin of PTS, are considered in a PWR with UTSG. The accident scenarios are the MSLB, the SGTR, and the SBLOCA. Reference, plausible average temperature, and pressure time trends are reported. The MSLB and the SGTR scenarios are characterized by repressurization during the transient with moderate final coolant temperature. The SBLOCA-type transient results in the lowest final coolant temperature but the relatively low pressure. From temperature and pressure variations shown in the figures, it is presumed that the pressure effect is dominant in the MSLB-type transient and the temperature effect is dominant in the SBLOCA-type transient. The phenomena listed in the second and the third columns of Table 15.26 are connected with PTS from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables given in Figs. 15.42 and 15.43 in the fourth column of the table.

1050

Thermal Hydraulics in Water-Cooled Nuclear Reactors

350 MSLB SGTR SBLOCA

300

Temperature (°C)

250

S-9-CO1 S-35-LVM3

200

S-56-STR 150 100 50 0 0

20

40

60 Time (min)

80

100

120

Fig. 15.43 PTS in PWR: average fluid temperature in RPV downcomer following MSLB, SGTR, and SBLOCA. 17.5 15 MSLB SGTR SBLOCA

Pressure (MPa)

12.5 10

S-22-GM2 S-34-LVM2

7.5 5 2.5 0 0

20

40

60 Time (min)

80

100

120

Fig. 15.44 PTS in PWR: pressure in RPV following MSLB, SGTR, and SBLOCA.

The calculation of average temperatures and pressures (Figs. 15.43 and 15.44) focusing on PTS evaluation, implies the calculation of local phenomena as discussed at item D. earlier (Fig. 15.41). Therefore, phenomena associated with PTS in Table 15.26 (phenomena at rows 1, 3, 4, 5, 6, and 7) are assumed to be indirectly visualized in Figs. 15.43 and 15.44.

Accident scenarios and phenomena in WCNR

1051

Table 15.26 Phenomena visualized by variables representative of the phenomenon PTS Phenomenon ID for: PTS No.

Acronym

Description

Fig. no.

Notes

1

S-9-CO1

15.43

See also Fig. 15.41, right

2

S-10-CO2

3

S-11-CO3

4

S-22-GM2

5

S-34-LVM2

6

S-35-LVM3

7

S-56-STR

Condensation in stratified conditions—horizontal Pipes Condensation in stratified conditions—PRZ Condensation in stratified conditions—SG-PS Global multi-D fluid temperature, void, and flow distribution—downcomer Liquid-vapor mixing with condensation—Downcomer Liquid-vapor mixing with condensation—ECCI in HL and CL Stratification in horizontal flow—pipes

15.19 15.51 15.44

See also Fig. 15.41, left

15.44 15.43

See also Fig. 15.41, right

15.43

See also Fig. 15.41, right

The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Condensation in stratified conditions—Horizontal Pipes (indirectly visualized, Fig. 15.43). This phenomenon can be combined with phenomena (6) and (7) in the present list. Injection from ECCS is at the origin of the phenomenon and incomplete mixing and condensation create more severe conditions for PTS. In the HL the phenomenon may occur because of simultaneous presence of steam flowing from core to SG and condensed liquid flowing counter-currently from SG to core (i.e., not necessarily connected with ECCS injection), as from the sketch on the right side of Fig. 15.41. (2) Condensation in stratified conditions—PRZ (indirectly visualized, Fig. 15.19). PRZ level rise in the condition of constant pressure or increasing pressure implies condensation at the liquid-steam interface. (3) Condensation in stratified conditions—SG-PS (indirectly visualized, Fig. 15.51). The phenomenon is expected with liquid-level formation inside U-tubes. The siphon-condensation mode of NC is controlled by this phenomenon. (4) Global multi-D fluid temperature, void, and flow distribution—Downcomer (indirectly visualized, Fig. 15.44). Namely the global multi-D fluid temperature distribution can be “directly” observed in the supporting diagram of Fig. 15.41. This phenomenon can be combined with the phenomenon (5) in the present list, with main reference to the global multi-D void distribution.

1052

Thermal Hydraulics in Water-Cooled Nuclear Reactors

(5) Liquid-Vapor mixing with condensation—Downcomer (indirectly visualized, Fig. 15.44). See also phenomenon (4) in the present list. This phenomenon deals with two-phase conditions in the downcomer (e.g., expected in the case of SBLOCA rather than MSLB or SGTR). (6) Liquid-Vapor mixing with condensation—ECCI in HL and CL (indirectly visualized, Fig. 15.43). See also phenomenon (1) in the present list. (7) Stratification in horizontal flow—Pipes (indirectly visualized, Fig. 15.43). See also phenomenon (1) in the present list.

15.4.12.4 Natural Circulation Natural circulation is a phenomenon or a phenomenological window part of accident scenarios and it is not an accident scenario. It is inserted into the list of Sections 15.4.2–15.4.12.12, which deal with accident scenarios owing to the following reasons: (a) NC is at the basis of the thermal-hydraulic design and the safety demonstration of water cooled reactors; NC is important for the thermal-hydraulics discipline, too. (b) NC focused studies, including experiments, have been performed in order to identify and characterize various phenomena which otherwise are expected to occur in different accident scenarios (primarily SBLOCA, but NC conditions are expected during each accident scenario where MCP is stopped).

Furthermore, NC constitutes the key topic of documents and reports already considered for deriving the list of phenomena in Table 15.2 (i.e., OECD/NEA/CSNI, 1996b; IAEA, 2001, 2002b, 2005b, 2009a, 2012). It is not the objective here to duplicate the content of those reports. The structure of the following discussion is according to Section 15.4.1 and specific for Sections 15.4.12.1–15.4.12.4, 15.4.12.8, and 15.4.12.12.

NC topics and facts Reference is made hereafter to the documents at row 30 in Table 15.1, namely, IAEA (2005b). Three elements characterize the NC: (I) the presence of gravity; (II) the presence of a heat sink and a heat source; and (III) the geometrical configuration of the concerned NC system.

This is shown in Fig. 15.45 where (to the aim of NC analysis in nuclear systems): -

the top-left sketch is basically applicable to PWR, PWR-O, PWR-V, PHWR and to passive systems like IC and PRHR; the top-middle sketch is basically applicable to BWR, RBMK and to the SS of UTSG; and the top-right sketch is basically applicable to CANDU.

Furthermore, sketches characterizing possible multiple NC flow paths are also given in Fig. 15.45. -

The bottom-left sketch is related to all reactors with UTSG (namely PWR, PHWR, and CANDU) and to PWR-V: multiple NC flow paths are possible between the heat source (core) and the heat sink (SG). This also includes flow reversal or NC stop in groups U-tubes and of horizontal tubes (experimentally observed in the case of UTSG).

Accident scenarios and phenomena in WCNR

1053

IHE

Fig. 15.45 Loop configurations of nuclear systems for NC.

-

-

-

The bottom-right sketch is related to core NC in case of CANDU: NC is possible with moderator tank constituting the heat sink. In this case unstable NC paths may establish in a channel-to-channel mode, where the loop is closed by the header and the heat sink is constituted by the moderator. The top-central sketch shows multiple NC paths possible in BWR: one of the central channels in the sketch may be interpreted as the core bypass region or the space between fuel boxes where liquid moderator flows upward in nominal operating conditions. NC may establish between core assemblies and bypass (with downward liquid flow). Unstable NC may also establish between high-power assemblies and low-power assemblies. The bottom-central sketch shows a series of NC loops. Namely power produced in the core can be removed to the environment by three NC loops (present case). This has also been called dual NC system; specific models have been developed (e.g., D’Auria et al., 1998). Among the other things, (a) the optimum position for the Intermediate Heat Exchanger (IHE) can be determined: a high elevation for the IHE improves NC in the loop with core and degrades the NC in the downstream loop, so optimization is needed; (b) the stability performance of the overall system, tightly coupled loops must be determined; furthermore, a perturbation (instability occurrence, see later) in the last loop may reflect on the first one.

The complexity of the NC phenomenon already derives from the geometry outline. Additional NC topics considered in IAEA (2005b) are: A. The NC flow regimes expected for PWR UTSG at various PS coolant mass inventories (Fig. 15.46). The following flow regimes are characterized and can be considered phenomenological windows in case of SBLOCA scenario with continuous coolant mass decrease in PS: - One-phase NC: The entire PS is in single phase and PRZ is losing mass to compensate the break flow. The heat sink of SG is available. Flow reversal in U-tubes may occur (De Santi et al., 1986), causing an increase of pressure drops (because of reduced cross flow area) for the overall system under natural circulation. The phase ends when PS

1054

Thermal Hydraulics in Water-Cooled Nuclear Reactors

2

6 W core %

5 Two phase

1.5

1

3

2

Reflux condenser

Single phase

Gc

4 GcW %

Dryout

Two phase instability

0.5

1 1

0 100

2

4

3

5

0 75

50

25

MI %

Fig. 15.46 NC flow regimes in PWR with UTSG at decay power.

-

-

-

achieves the saturation pressure corresponding to the HL fluid temperature in nominal conditions. Two-phase NC: This includes the rising part of the curve, the region of the maximum value, and the first part of the descending region in Fig. 15.46. The phase starts with void formation due to simultaneous boil-off in the core and break induced depressurization. The rising part is due to a positive balance between driving head (i.e., the difference in average density of the coolant in the hold-ascending and the cold-descending loop regions) and pressure drops when MI decreases; then, a zero balance results in the peak region; further mass reduction causes a negative balance between driving head and pressure drops. Flow reversal in U-tubes may occur (see e.g., D’Auria, 2012) with same consequences (pressure drop increase in the overall loop) outlined for one-phase NC. Siphon condensation: CCFL at U-tube entrance, liquid formation inside the U-tubes due to condensation (or heat removal from SS), liquid level rise till the top in the concerned U-tubes, sudden liquid draining caused by the “siphon-effect,” are at the origin of siphon-condensation oscillations in RCS of PWR (D’Auria and Galassi, 1990b). Phase shift may occur among different groups of U-tubes. Oscillations are reflected all along the loop and are associated with the presence of the U-tube. Thus, siphon condensation oscillations are expected in PWR equipped with UTSG. Reflux condensation: When MI is further reduced a new stable NC regime establishes, called reflux condensation or reflux condensing mode. In this situation flowrate at core inlet is almost zero. At core outlet, upward two-phase mixture and downward liquid flow occur simultaneously: two-phase mixture produced in the core from liquid vaporization flows to the SG; condensate flows back to the core. The mechanism is consistent with flooding curve and CCFL at UTP, at the UP-HL connection, at the HL bend, and at

Accident scenarios and phenomena in WCNR

1055

25 3% 2.5%

20

Upper envelope

3%

G/P (kg/MWs)

Lower envelope

15

RBMK BWR

10 % = % of nominal power G = core flow-rate P = corepower V = volume of primary loop

5

0 900

725

550 MI/V (kg/m3)

375

200

Fig. 15.47 NC flow map in PWR with UTSG at decay power with consideration of BWR and RBMK. U-tubes connection with SG inlet chamber. The mechanism allows core power removal at MI values as low as 40% the nominal value. B. The NC flow map derived from collecting experimental data from several PWR-ITF experiments (Fig. 15.47; Cherubini et al., 2008a; proposed by D’Auria et al., 1994). Basically, upper and lower boundaries have been found for NC flowrate (i.e., the ratio G/P [NC flowrate at core inlet over core power] vs. MI/V [mass inventory in PS over coolant mass volume]) in PWR considering several experimental conditions, including scaling for ITF design. The flow map allows the evaluation of the NC performance of different nuclear systems by comparing with the spectrum of experimental data valid for PWR UTSG. The following key remarks apply (i.e., making reference to the NC core flow as a function of PS MI with core power at decay values around 3% of the nominal value): - NC calculated flowrate for any PWR UTSG is bounded by the upper and lower curves with values (as expected) around the mean region of the map. - PWR with OTSG (here called PWR-O) are characterized by low NC performance in two-phase region because of the presence of the so-called candy-cane in HL: this causes degradation of NC flow (compared with U-Tube PWR) when MI is below 90% the nominal value. - PWR with HOSG (VVER, here called PWR-V) are characterized by suitable NC performance, i.e., there is no important difference related to PWR with UTSG; in the case of VVER-440 additional complexity arises in the curve because of the presence of loop seal in the HL (in addition to loops seal in CL). - CANDU reactors are characterized, as expected, by low NC performance compared with PWR equipped with UTSG. The reason is the presence of long pipes upstream and downstream each horizontal channel which causes large increase in friction pressure drops. This is true notwithstanding higher (potential) driving head owing to the relative elevation between the core and the UTSG (compared with PWR). The low NC performance in the case of CANDU is derived from considering “only” the flow between core and SG:

1056

Thermal Hydraulics in Water-Cooled Nuclear Reactors

the NC induced by channel to channel flow (bottom-right sketch in Fig. 15.45, where the moderator tank constitutes the heat sink) is not accounted in this evaluation. - BWR and RBMK loop configurations are different from PWR. Furthermore, (a) NC is part of the BWR normal operation conditions; (b) NC occurs between core and bypass and between core and DC; (c) only two-phase conditions in the core with RCS pressure and DC (or steam drum in the case of RBMK) level close to nominal can be compared with data in Fig. 15.47. This signifies one point in the map for each of BWR and RBMK (D’Auria and Galassi, 2014). The points show high values for NC flows in both cases. C. The stability of NC flows (Fig. 15.48; Vijayan, 2014). NC flows in industrial systems including NPP are characterized by low driving forces, i.e., compared with systems where pumps are installed. This makes the NC systems “prone to instability” namely when boiling and condensation processes occur. Type-I and Type-II instabilities are distinguished in Fig. 15.48, left diagram as originally proposed by Fukuda and Kobori (1979), for DWI-related oscillations; low-flow and low-power characterize Type-I while (relative) high-power and high-flow characterize Type-II. A typical stability map is shown in the right side of Fig. 15.48. The effects of various system and BIC-related parameters are studied and affect the stability (or the instability) region. Examples are the length of cooler and heater, the L/D of piping, the orientation (related to the gravity force) of cooler and heater, the inlet (stabilizing if increased) and the outlet (destabilizing if increased) local pressure drop for the heat source, the time constants for conduction heat transfer (namely for heaters). D. The reliability evaluation for passive systems and the NC phenomenon (no supporting figure) (D’Auria et al., 2014). In Chapter 1, the subject of reliability of passive system has been excluded from the scope of the book. Nevertheless, a few notes are provided hereafter. The design of passive systems including IC and PRHR is based upon the NC phenomenon. The NC (as already stated) depends upon the existence of a heat source and a heat sink at different elevations in a gravity environment. The reliability of the TH phenomenon appears equal to one. This is not the case, specifically when boiling and condensation phenomena are involved in an NC system. Furthermore, a passive system suitable for applications in nuclear reactor safety may never operate at steady-state conditions: either the core power (decay heat) or the fluid temperatures or the levels in pool, where a heat exchanger of a passive system is immersed, change with time. This is in addition to possible changes in pressure or in other operating conditions (position of the stem of a valve) associated with a specific accident scenario. Thus, an infinite number of combinations (time functions) occur to generate the driving force in an NC system. One may expect a zero driving force 1000

100

ΔP Pressure

600

60

400

40

200

0.6

80

Type-II Unstable

0.4 NPCH

Type-II instability

Pressure (bar)

ΔP (Pa)

800

Stable

0.2

20 Type-I Type-I instability

0 0

4000

8000 Time (s)

12,000

0 16,000

0 0.01

0.03

0.05 Nsub

0.07

Fig. 15.48 NC flows: the stability issue. Left: characterization of “type-I” and “type-II” instability. Right: typical stability map for DWI.

Accident scenarios and phenomena in WCNR

1057

occurring in some cases which leads to stop in NC flow. This subject has been extensively studied (references can be found in Chapter 1) and reliability values less than unity have been found for NC systems.

NC stop constitutes one example of situations challenging the reliability of the phenomenon. When boron dilution is involved, NC restart (after the stop) may put at risk the core integrity (further discussed in Section 15.4.12.8).

Phenomena connection with accident scenario: NC The NC scenario for a PWR equipped with UTSG can be derived from Fig. 15.49 taken from PWR calculation (IAEA, 2005b) (row 30 in Table 15.1, data elaboration by D’Auria and Galassi), Fig. 15.50 (Cho et al., 2009) (row J in Table 15.5, ATLAS experimental data), and Fig. 15.51 (Takeda et al., 2013) (row K in Table 15.5, LSTF experimental data). The NC scenario calculated for a PWR with UTSG can be derived from the core flowrate shown in Fig. 15.49. Starting from a situation PS full of coolant in liquid phase (one-phase NC) at time zero, PS MI inventory is decreased from 100% till about 40% at 20,000 s. The resulting NPP scenario is consistent with the description of NC flow regimes related to Fig. 15.46 (one-phase, two-phase, siphon condensation oscillations, and reflux condenser modes can be observed). The scenario of NC between core and DC can be derived from Fig. 15.50 and includes the QF progression during reflood. Data measured in the ATLAS facility related to LBLOCA simulation in APR1400 are reported in the figure. Reflood is caused by the injection of accumulators (here called SIT). Vent valves between core and 2000 I-4-NCBC

1750

Flow rate (kg/s)

1500

B-5-IF1 B-6-IF2

G-core

1250 1000 750 500

B-1-COH, Section 15.4.12.9

I-22-RCM

250 0 –250 0

2500

5000

7500

Fig. 15.49 NC in PWR: core flowrate.

10,000 Time (s)

12,500

15,000

17,500

20,000

1058

Thermal Hydraulics in Water-Cooled Nuclear Reactors

4 I-14-NC5

Core

3.5

Downcomer Quench front elevation

3 TAF

Level (m)

2.5 2 1.5 1

BAF

0.5 SIT injection

0 1250

1350

1450

1550 Time (s)

1650

1750

1850

Fig. 15.50 NC in ATLAS test facility: core and DC levels and quench front.

14 T1 T2 T3 T4 T5 T6

S-43-NCG

12 U-tube level (m)

S-11-CO3, Section 15.4.12.3

10 8 6

S-6-CCF4, Section 15.4.3 S-19-ED5, Section 15.4.3

4 2 0 0

2000

4000

6000 Time (s)

8000

10,000

12,000

Fig. 15.51 NC in LSTF test facility: levels in SG tubes (loop B).

downcomer are open at about 600 s into the transient. During the early phase of reflood the bottom-up QF progression is delayed related to the liquid-level rise in the core (IAFB condition typical of high reflood rates). Top-down QF progression can also be observed from the figure. The performance of SG tubes during NC in the presence of noncondensable gases can be derived from Fig. 15.51. Data measured in different SG tubes are shown. The concerned transient is an SBLOCA with AM. The NC mass flow rate temporarily

Accident scenarios and phenomena in WCNR

1059

decreased following ACC flow increase due to SG depressurization. The gas entered the primary system at the end of liquid delivery from ACC at about 2500 s. The consequence of gas entrance was a decrease in heat transfer between PS and SS (the PS depressurization rate decreased when gas entered the system). The noncondensable gases accumulated differently in the two loops of the facility. This contributed to the creation of a nonuniform performance of U-tubes which occurred (in the case of loop B, shown in the figure) till about 7500 s into the transient. The phenomena listed in the second and the third columns of Table 15.27 are connected with NC from the cross-link process in Tables 15.3–15.5. Those phenomena are associated with variables given in Figs. 15.49–15.51 in the fourth column of the table. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): Table 15.27 Phenomena visualized by variables representative of the phenomenon, or PHW, NC Phenomenon ID for: NC No.

Acronym

Description

Fig. no.

Notes

1

I-4-NCBC

15.49

2

I-5-BC

Boiler condenser mode (of NC) Channel and bypass axial flow and void distribution

3

B-5-IF1

15.49

4

B-6-IF2

5

I-9-INC

6

I-10-NC1

7

I-11-NC2

Interfacial friction in horizontal flow Interfacial friction in vertical flow Intermittent two-phase NC NC, one-phase, and two-phase—PS and SS NC core and downcomer

8

I-13-NC4

9

I-14-NC5

10

S-43-NCG

11

I-22-RCM

Equivalent to two-phase NC Occurring in BWR core, row D in Table 15.5 In any two-phase system In any two-phase system Equivalent to siphon condensation *Focusing on SS (Fig. 15.49 for PS) Occurring in BWR RPV *Focusing on VVER1000 Vent valves installed in APR1400 Experimental data show the effect of noncondensable gas in PS Stable NC regime in PWR with UTSG

15.14

15.49 15.49 15.17* 15.13

NC core gap, downcomer, and dummy elements NC core, vent valves, downcomer Noncondensable gas effect including effect on condensation HT in RCS

15.18*

Reflux condenser mode and CCFL

15.49

15.50 15.51

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

(1) Boiler condenser mode of NC (visualized, Fig. 15.49). This is basically equivalent to two-phase NC discussed under item A. earlier; see also Fig. 15.46. (2) Channel and bypass axial flow and void distribution (indirectly visualized, Fig. 15.13). This is related to the BWR core during SBLOCA. Additional information, not in the form of time histories, can be derived from fundamental research documented in the papers at row D of Table 15.5, Kok et al. (1997), and Yang et al. (2012b). (3) Interfacial friction in horizontal flow (indirectly visualized, Fig. 15.49). The transient evaluation of interfacial frication in horizontal flow is needed to calculate fluid velocities at any location of a thermal-hydraulic system (e.g., RCS and containment) in two-phase conditions. The phenomenon occurs in any accident scenario. It has been (arbitrarily) assigned to NC and is visualized through the core inlet flowrate (i.e., also the HL flowrate where horizontal flow is encountered). (4) Interfacial friction in vertical flow (indirectly visualized, Fig. 15.49). The transient evaluation of interfacial friction in vertical flow is needed to calculate fluid velocities at any location of a thermal-hydraulic system (e.g., RCS and containment) in two-phase conditions. The phenomenon occurs in any accident scenario. It has been (arbitrarily) assigned to NC and is visualized through the core flow-rate (see also Melikhov et al., 2011) for interfacial friction effect in HOSG SS. (5) Intermittent two-phase NC (visualized, Fig. 15.49). This is basically equivalent to the siphon condensation NC flow regime discussed under item A. earlier; see also Fig. 15.46. (6) NC, one-phase, and two-phase—PS and SS (NC PS visualized, Fig. 15.49; NC SS indirectly visualized, Fig. 15.17). The phenomenon includes NC occurring in PS and SS of PWR. The PS NC is described under item A., earlier. The SS NC is at the basis of SG performance in nominal operating conditions and in case of any accident scenario. (7) NC core and downcomer (indirectly visualized, Fig. 15.13). This is related to the BWR RPV during various accident scenarios (e.g., SBLOCA). (8) NC core gap, downcomer, and dummy elements (indirectly visualized, Fig. 15.18). This is related to the VVER RPV and core during various accident scenarios (e.g., SBO). (9) NC core, vent valves, downcomer (visualized, Fig. 15.50). The NC inside RPV equipped with vent valves constitutes the focus for this phenomenon. This is of interest for reflood PHW following LBLOCA and also considered in Section 15.4.2 (see Fig. 15.5 and related discussion). (10) Noncondensable gas effect including effect on condensation HT in RCS (visualized, Fig. 15.51). The injection of noncondensable gas in PS (e.g., coming from ACC) enhances the disuniformity in the performance of SG tubes. (11) Reflux condenser mode and CCFL (visualized in Fig. 15.49). This is a stable NC flow regime discussed under item A. earlier; see also Fig. 15.46.

Furthermore, phenomena S-6-CCF4 and S-19-ED5, discussed in Section 15.4.3, related to SBLOCA are visualized in Fig. 15.51.

15.4.12.5 MSLB combined with SGTR Two key motivations for this section shall be mentioned: l

l

After any assigned accident scenario is triggered, the probability of another (independent) accident to occur may remain the same (or increase, see next item) as in nominal operating conditions: in other words, following any event a new DBA framework could be considered. In some cases an accident can be the consequence of a previous event. Examples of consequential accidents are:

Accident scenarios and phenomena in WCNR

-

1061

SBO following LOCA owing to weakness of the electrical net; SBLOCA (e.g., due to PORV stuck open) following LOFW or SBO; LCC/SW following SBO also inducing SBLOCA due to break in MCP seal circuit; LOCA following earthquake (in this case the trigger is an external event); ATWS following some AOO (only because scram is requested); SBLOCA following CRE which causes an opening in the RPV; MSLB induced by a TT in PWR which causes SRV stuck open in SG; and SGTR (or PRISE) following MSLB because of induced fluid-dynamics loads including pressure wave loads.

A systematic listing of consequential accidents is well beyond the framework for the current document. The characterization of phenomena in consequential accidents can be based upon the cross-link in Table 15.3 (i.e., “summing” phenomena expected from two accident scenarios). However, additional specific studies may be needed for characterizing phenomena during combined accident scenarios In the present section the attention is focused upon SGTR induced by MSLB in PWR.

Qualitative accident scenario: SGTR with consequential MSLB Qualitative accident scenarios for SGTR and MSLB are discussed in Sections 15.4.7 and 15.4.8 and not repeated hereafter. Combination of the two scenarios, actually SGTR can be a consequence of MSLB (SGTR at the origin of MSLB is also possible in case of stuck open SRV in the affected SG), does not bring to phenomena not identified in Table 15.3, specifically when a single tube is concerned. However, the recovery of the NPP shall be different in MSLB and in MSLB with consequential SGTR. Furthermore, MSLB combined with SGTR and multiple SGTR accident scenarios have similar features from a thermal-hydraulics view point: namely, multiple SGTR causes the opening of SRV, which (for short time) is equivalent to a rupture of the steam line.

Quantitative accident scenario: variable trends—SGTR with consequential MSLB SGTR with consequential MSLB is discussed in the following. No TSE is provided because it is largely affected by peculiarities of individual NPP and by calculation assumptions for the event including the number of broken tubes. The SGTR with consequential MSLB for a three-loop PWR equipped with UTSG can be derived from Figs. 15.52 and 15.53, taken from Jimenez et al. (2013) (row 29 in Table 15.1). Temporarily stuck open SRV in the affected SG is considered in the former figure, which includes two additional sensitivity calculations (only in the right diagram of the figure) dealing with the time-period of SRV stuck open. Failed stuck open SRV in the affected SG is considered in the latter figure, which includes one additional sensitivity calculation (only in the right diagram of the figure) where no cooldown of intact SG is actuated (black line). Making reference to Fig. 15.53, it shall be noted that whatever is condition for cooldown of intact SG, the release of radioactive fluid to the environment cannot be stopped without direct depressurization of PS via the PRZ PORV. However, the situation without cooldown results in a lower narrow level in the affected SG, thus lower mass release from the stuck open SRV.

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

50

17

40

RCS

Pressure (MPa)

13

SG 3 (dam.)

35

SG 1 & 2 (int.)

11

30 min. SG PORV open

30

20 min. SG PORV open

9

25

11 min. SG PORV open

20

7

15 5 10 3

Integrated mass flow (kg) × 103

45

15

5 0

1 4800

6000

7200

8400

Time (s)

Fig. 15.52 SGTR with temporarily stuck open SRV in the affected SG in PWR with UTSG: PS and SS (damaged and intact) pressure and mass lost from affected SRV (three sensitivity cases).

17

100 RCS S-51-SEP SG 3 (dam.) SG 1 & 2 (int.) SG 3 (dam.) Level SG 3 (dam.) SD Level

Pressure (MPa)

13

90 80 70

11

60

9

50

7

40 30

5

SG narrow range level (%)

15

20 3 1 4800

10 5500

6200

6900 Time (s)

7600

8300

0 9000

Fig. 15.53 SGTR with stuck open SRV in the affected SG in PWR with UTSG: PS and SS (damaged and intact SG) pressure and mass lost from affected SRV (three sensitivity cases).

Accident scenarios and phenomena in WCNR

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Table 15.28 Phenomenon visualized by variables representative of the accident scenarios during SGTR induced by MSLB Phenomenon ID for: SGTR with consequential MSLB No.

Acronym

Description

Fig. no.

Notes

1

S-51-SEP

Separator behavior (and* flooding, steam penetration, liquid carry-over)

15.53

*BWR related

Calculated scenarios in both figures show the importance of a timely actuation of PS depressurization for minimizing the release of radioactivity to the environment (this is evaluated in the various sensitivity calculations in the paper by Jimenez et al., 2013). The complexity of the scenario also results from the reported variables as well as the importance of predicting correctly the phenomenon “separator behavior.”

Phenomena connection with accident scenario: SGTR with consequential MSLB The phenomenon identified in the second and the third columns of Table 15.28 is connected with SGTR with consequential MSLB from the cross-link process in Tables 15.3 and 15.4. This phenomenon is associated with variables in Fig. 15.52, as reported in the fourth column of the table. The following phenomenon is outlined in this section and visually described through the use of calculated accident scenarios and related time trends: (1) Separator behavior and flooding, steam penetration, liquid carry-over (visualized Fig. 15.53). The separator and the steam-liquid separation process are directly connected with the (narrow) level formation at the top of the downcomer in the secondary side of SG. The subset of phenomena flooding, steam penetration, and liquid carry-over reported for S-51-SEP of importance for BWR is also important in the current scenario.

15.4.12.6 Fuel channel blockage The blockage of a fuel channel (e.g., caused by deposition and chemical attack or by pieces of solid material present in the coolant) may occur at the inlet of any fuel assembly. Closure or flow obstruction of a valve eventually installed upstream or downstream a power generation channel may also generate an FCB event. The issue may become of safety concern in the case of channel-type reactors (discussed hereafter) and research reactors (Adorni et al., 2005; Adu et al., 2015).

Qualitative accident scenario: FCB Selected topics for FCB transients of interest to thermal-hydraulics are: A. The blockage may be partial or total. In case of partial blockage, the target of analysis may be the characterization of the “amount of blockage,” which causes ONB (case of research reactors), or CHF or damage to the fuel.

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B. In case of total blockage (or blockage section higher than a threshold value), the damage of the fuel cannot be avoided. C. The item above occurs because early detection is difficult or impractical (possible detection is discussed by D’Auria et al., 2008c, in relation to RBMK, see also Section 15.4.12.7). D. Coolant (fast) vaporization in the affected channel and radiation heat transfer toward the channel walls are important phenomena. E. Propagation of the damage to neighboring channels is a safety concern. The FCB accident constitutes one plausible precursor event for the MPTR discussed in Section 15.4.12.7.

Quantitative accident scenario: variable trends and TSE – FCB The FCB scenario for an RBMK from Table 15.30 (in Section 15.4.12.7 related to MPTR originated by FCB, not repeated here) and Figs. 15.54 and 15.55, taken from D’Auria et al. (2005) (row 9 in Table 15.1). Typical timing for the event can be seen in Figs. 15.54 and 15.55 and in Table 15.30. Other than radiation heat transfer phenomenon, nuclear fuel material performance (including possible ballooning and H2 production) and mechanical resistance of PT at high temperature (creep phenomenon) are of interest during the transient. Fuel and PT failures shall be determined based on variable time histories reported in the previous diagrams. Moreover, during the event a local (channel related) neutron flux perturbation is expected due to coolant vaporization (coolant is not moderator in the RBMK) and to increase in fuel temperature.

Phenomena connection with accident scenario: FCB The phenomenon identified in the second and the third columns of Table 15.29 is connected with FCB from the cross-link process in Tables 15.3 and 15.4. This phenomenon is associated with variables in Fig. 15.54, as reported in the fourth column of the table. 1670

Temperature (K)

1470

S-27-HT2

1270 1070 Case 1

870

Case 2 Case 3

670 470 270 –100

0

100

200 300 Time (s)

400

500

600

Fig. 15.54 FCB in RBMK and phenomena characterization: RST at different axial elevations.

Accident scenarios and phenomena in WCNR

1065

1170 1070

Temperature (K)

970 870 770

Case 1 Case 2

670

Case 3

570 470 370 270 –100

0

100

200 300 Time (s)

400

500

600

Fig. 15.55 FCB in RBMK and phenomena characterization: PT wall temperatures at different axial elevations. Table 15.29 Phenomenon visualized by variables representative of the accident scenarios during FCB Phenomenon ID for: FCB No.

Acronym

Description

Fig. no.

1

S-27-HT2

HT [radiation]-core

15.54

Notes

The following phenomenon is outlined in this section and visually described through the use of calculated accident scenarios and related time trends: (1) HT [radiation]-core (visualized, Fig. 15.54). Fuel rod surface temperatures above about 600°C (or 900 K) imply radiation heat transfer. Absorbing media are the coolant (steam and droplets) and mainly the PT walls which is heated up by radiation (Fig. 15.55). Practical problems associated with the characterization of the radiation HT phenomenon are the complexity of the geometry involved, the status of irradiating and absorbing surfaces, and the difficulty in distinguishing between radiation and film boiling contributions to HT.

15.4.12.7 MPTR Multiple pressure tube rupture is a potential catastrophic accident in pressure tube reactors: the rupture of one pressure tube (possible) propagates to neighboring tubes destroying the entire core. MPTR may be triggered by loss of integrity of the wall caused by pressure load (e.g., propagation of a defect in the material), by combined

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temperature and pressure load (e.g., overheating of one fuel assembly due to channel blockage), and by temperature only (e.g., pressure may be low and fuel melt may occur). MPTR must be avoided at the design stage level of the reactors. Making reference to the pressure tube water cooled reactor: (a) MPTR is excluded in PHWR, e.g., Atucha type reactor mostly because pressure difference across one single tube is “low” (i.e., less than 0.1 MPa); (b) MPTR is possible in CANDU; however, the presence of water in-between the pressure tubes makes realistic (although with negligible probability), the event only in case of fuel melt in one channels with propagation of the rupture only in the gravity force direction; (c) MPTR may occur in RBMK following the rupture of one pressure tube because the surrounding graphite blocks may not absorb the pressure and the displacement loads and may transmit those loads neighboring tubes. Attention is focused hereafter to RBMK.

Qualitative accident scenario: MPTR The sketch in Fig. 15.56, taken from D’Auria et al. (2005), see also D’Auria et al. (2008b), gives an idea of the MPTR. Geometric details and operating conditions for RBMK core can be found in the cited references. Selected topics for MPTR transients of interest to thermal-hydraulics are: A. “Hot” and “cold” graphite stacks surrounding pressure tubes are part of the core. Namely, coolant in the channels where CR is located is kept below the boiling point by suitable circulation flow. Temperature of graphite is a function of the radial and axial coordinates. B. The thickness of the gap between adjacent graphite blocks and upon neutron fluence. C. The gas, helium or nitrogen or a mixture helium-nitrogen, circulates inside the gaps among graphite stacks. D. Mechanical, thermal, and neutron-related properties of materials constituting the core are largely a function of core life. E. The deformation characteristic of the fuel channels and associated graphite stack depend upon the type of load (uniform, punctual, etc.) and upon the location of the triggering rupture. Furthermore, other than the pressure tube that constitutes the most resistant part of the ensemble, the graphite blocks, the fuel bundles, and the central bar inside each fuel assembly contribute to the overall stiffness. F. The detection of the event is not immediate and so the occurrence of scram.

The outline at the earlier listed items and Fig. 15.56 shows that MPTR is a multidisciplinary scenario where thermal-hydraulics, neutron physics, materials, and structure mechanics play a role.

Quantitative accident scenario: variable trends and TSE—MPTR MPTR is typically triggered by the rupture of one tube. This may be originated by FCB discussed in Section 15.4.12.6. The single-tube rupture may evolve into an MPTR. Bases for MPTR analyses in RBMK can be derived from the TSE in Table 15.30, and Figs. 15.57 and 15.58 taken from D’Auria et al. (2005) (row 25 in Table 15.1). The FCB accident scenario, plausible trigger for MPTR, is not easily detectable. Detection may occur after the rupture of the pressure tube: early detection is discussed by D’Auria et al. (2008c). Then, MPTR, if not prevented at a design level evolves in a few seconds.

Accident scenarios and phenomena in WCNR

1067

Initiating failure location

NOTE: corner-wise deformation patterns are not considered. Worst load direction to be considered in deterministic safety analysis

Pressure distribution around failed ch.

FC blockage

Selected row of resisting columns

NK feedback

RC pressurization & mech loads On graphite stacks

Fuel Failure

CHF

Fuel & Coolant overheating

PT & Brick Failure

H2 & FP release & diffusion

TH NK FPM & SM FP & H2 The ICM to prevent PT failure

MPTR The role of the reactor tank

Fig. 15.56 Sketch for modeling MPTR in the core of an RBMK. CFD, computational fluid dynamics; CHF, critical heat flux; FC, fuel channel; FP, fission products; FPM, fuel pin mechanics; ICM, individual channel monitoring; MPTR, multiple pressure tube rupture; NK, neutron kinetics; PT, pressure tube; RC, reactor cavity; SM, structural mechanics; TH, thermal-hydraulics.

Pressurization of the interassembly region after the pressure tube rupture (conditions for rupture calculated in Fig. 15.58, left) occurs with a large-axial gradient as indicated in Fig. 15.58, right. Thereafter, the possible MPTR occurrence becomes a matter of interest for structural mechanics: strain and stressed in graphite stacks and in intact pressure tubes are involved (sketch in Fig. 15.57).

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

Table 15.30 FC blockage event in RBMK at the potential origin of MPTR: (selected) imposed and calculated events N

Event

I/R

Time (s)a

Notes

1 2

Full blockage occurrence Dry-out occurrence

I I

0 1–40

3

Clad surface temperature achieves 700°C

I

20–70

4

Pressure tube heating rate before break (K/s) The pressure tube temperature achieves 700°C (tB)

R

10–12

R

71

Full blockage in 0.1 s Range given for the various axial locations Clad collapse occurs. Range given for the various axial locations Relevant for calculating tube rupture time Start of creep mechanism. Integrity of pressure tube challenged

Time of maximum load acting upon neighboring graphite stack Value of the maximum load on neighboring graphite stack (Ton) Scram occurrence

R

tB + 2

I

22

Including dynamic load

tB + 2.2

Pressure increase in RC ¼ 0.075 MPa

5

6

7

8 9 10 11

Quench of wall temperatures above break axis Max pressurization rate in gas-gap (MPa/s) Maximum stress in graphite before crack propagation (MPa)

tB + 5 0.6

0.04 MPa/s in the cavity

6.6

The value largely depends on parameters like fluence

a

Or quantity as defined in the second column.

Coupled three-dimensional thermal-hydraulic, neutron physics, and structural mechanical calculation allowed the conclusion that design features of RBMK core are such to prevent the MPTR (D’Auria et al., 2005). The steel cylinder surrounding the graphite core stacks constitutes the ultimate resistant element effective in avoiding the propagation of the rupture. This conclusion may not apply in the case of some elements located in the periphery of the core. In this situation MPTR occurrence depends upon a specific direction for the pressure tube rupture axis. Therefore, the probability of MPTR is calculated to be negligible.

Phenomena connection with accident scenario: MPTR The phenomenon identified in the second and the third columns of Table 15.31 is connected with MPTR from the cross-link process in Tables 15.3 and 15.4. This phenomenon is associated with variables in Fig. 15.58, as reported in the fourth column of the table.

700

600 1

800

900

Time after tube rupture

1

2

2

3

3

39.9 c

4

4

37.9 c

5

5

39.9 c

6

6 0

10

21 31

7

8

9

10

11

12

(A)

13 14

(B)

Fig. 15.57 MPTR accident scenario in RBMK: (A) tube deformation process after the blockage event; (B) “stabilized” situation at the end of the analysis.

Pressure Tube rupture temperature (K)

1350 Break occurrence time (t = 74 s)

1250 1150 1050 950 850 750

Experimental Limit curve (average) obtained for PT of RBMK Limit curve Sup. Limit curve Inf. Limit step curve used for RELAP

650 RELAP result

t= 0 s

550 0.00E+00 1.00E+06 2.00E+06 3.00E+06 4.00E+06 5.00E+06 6.00E+06 7.00E+06 Differential pressure across pressure tube wall (Pa)

Fig. 15.58 MPTR accident scenario in RBMK. Left: trajectory in the temperature-DP space to determine rupture time. Right: pressure in the gap along the axis of the affected channel.

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

Table 15.31 Phenomenon visualized by variables representative of the accident scenarios during MPTR Phenomenon ID for: MPTR No.

Acronym

Description

Fig. no.

Notes

1

B-8-PW

Pressure wave propagation

15.58

Also Fig. 15.66

The following phenomenon is outlined in this section and (visually) described through the use of calculated accident scenario and related time trends: (1) Pressure wave propagation (visualized, Fig. 15.58). The pressure tube rupture causes a pressure wave propagating from the PS at high pressure into the reactor tank of the RBMK core passing through the region which separates the graphite stacks. The region has a small thickness which causes the axial pressure gradient that can be observed in Fig. 15.58. Moreover the pressure wave: (a) can be visualized for tenths of a second and (b) is at the origin of loads on graphite stacks around the broken pressure tube as illustrated in Figs. 15.56 and 15.57.

15.4.12.8 Boron Dilution The motivations for a specific boron section have been given under item (e) in Section 15.4.12. Wide literature exists in relation to the use of boron in nuclear technology and in reactor safety (e.g., Ebert, 1995; Pla et al., 2010). Selected topics and connection with thermal-hydraulics are discussed hereafter.

Boron topics and facts Boron connected topics/facts, which have a relation with thermal-hydraulics, are: A. Dilution: Boron dilution is associated with the boiling and condensation processes. The steam is free of boron (distillation type of process) and the condensation of steam produces a boron diluted liquid. The boron diluted liquid may accumulate in loop seals in a time period when NC is close to zero and pushed toward the core when NC restarts (further discussion below in this section). B. Transport: Boron transport occurs at global level in the core. Any motion of borated fluid mass into the PS shall be called boron transport. C. Mixing: Boron mixing occurs locally in the primary system, noticeably in CL, DC, and LP when masses of fluid having different boron concentration enter in contact. D. Deposition: Boron deposition occurs in boiling regions (i.e., subcooled boiling length in PWR core). Boron deposition causes crud formation and may affect the local fission power production. E. Stratification: Boron stratification may be induced by liquid temperature stratification (different liquid density and temperature cause differences in boron concentration) or, in the long term (weeks or months), by gravity effect in a stagnant tank at constant temperature. However, following Graffard and Goux (2006) and Nourbakhsh and Cheng (1995), the boron stratification is the phenomenon occurring in DC. This should be intended as a special situation of boron mixing (item C.). F. Crystallization: High boron concentration flowing across geometric discontinuities may cause deposit and obstructions of the flow section.

Accident scenarios and phenomena in WCNR

1071

G. Neutron reactivity excursion: Neutron flux excursion induced by boron may occur when a liquid diluted born plug arrives in the core or when boron crud detaches from fuel rods. H. System quantities associated with boron can be emphasized as follows for a large UTSG PWR unit: - Boron concentration at BOL/BOC conditions: up to 2000 ppm. - Volume occupied by coolant/moderator in the core: 20 m3. - Reactivity worth associated with boron in the core at BOC/BOL: up to $40. This is the reactivity that could be inserted in the case all borated liquid is substituted by unborated liquid. - Reactivity associated with all control rods: <$20. - Mass of liquid (potentially unborated) in each loop seal: 10 m3.

In Section 15.4.12.4, the possibility of NC stop is mentioned. During the NC stop (typically short, i.e., a few dozen seconds) boron diluted plugs of liquid may form in the loops seals. Once NC restarts (e.g., due to positive imbalance between ECCS and break flowrates) the diluted boron plug may be transported toward the core. The NC stop and restart is illustrated in Fig. 15.59 and the consequences of boron diluted plug motion are discussed hereafter.

Phenomena connection with accident scenario: Boron Boron phenomenon-related information can be derived from Fig. 15.60, left, taken from PWR-core basic calculation by Jimenez et al. (2015) (row 4 in Table 15.1). The boron mixing and stratification in downcomer can be (color plot) observed at a representative time snap-shot in the right side of the figure, taken from PWR calculation by Graffard and Goux (2006) (row 4 in Table 15.1). It shall be noted that the right picture and the left time trend (actually obtained from different and not connected PWR calculations) in Fig. 15.60 can be considered as results of consequential analyses, started with the movement of the possible unborated boron plug transported toward RPV by NC restart in Fig. 15.59. 1

Core flow rate (–)

0.8 0.6 0.4 0.2 0 –0.2 0

1000

2000

3000 4000 Time (s)

5000

6000

7000

Fig. 15.59 NC stop and restart calculated for a PWR with UTSG: the NC restart may cause the motion of boron diluted plug toward the RPV.

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

1.E+05

Power (MW)

1.E+04 S-2-BO

1.E+03 1.E+02 1.E+01

1.E+00 4

4.5

5

5.5

6 6.5 Time (s)

7

7.5

8

Fig. 15.60 Effect of transport of boron diluted plug into the RPV of a PWR with UTSG. Left: core power. Right: snapshot imagine for boron concentration in DC. Table 15.32

Visualization of boron-related phenomenon

Phenomenon: Boron No.

Acronym

Description

Fig. no.

Notes

1

S-2-BO

Boron mixing and transport

15.60

Stratification of Boron (A-2-BO) in Fig. 15.60, left

The phenomenon identified in the second and the third columns of Table 15.32 is the outcome of the cross-link process in Tables 15.3 and 15.4. The phenomenon is associated with variables in Fig. 15.60. The following phenomenon is outlined in this section and (visually) described through the use of calculated accident scenarios and related time trends: (1) Boron mixing and transport (indirectly visualized, Fig. 15.60). Three-dimensional analyses are needed for a comprehensive evaluation of boron-related phenomena. Namely transport of boron and formation of diluted boron plugs can be calculated by one-dimensional model. However, three- dimensional analyses are needed for mixing and stratification (or mixing in vertical components) (i.e., at the basis of results shown on the right-hand side of Fig. 15.60). The calculation of the effect of boron upon fission power also needs coupled threedimensional thermal-hydraulics and neutron-physics calculation (i.e., at the basis of results shown on the left-hand side of Fig. 15.60).

The statements at items C. and E. also apply to the phenomenon A-12-BO of Table 15.9. This is directly visualized in the right sketch of Fig. 15.60.

15.4.12.9 LOFA and MCP-trip and LCC/SW The MCP-trip accident scenario is part of group of four scenarios where similarities in system performance are expected (at least in relation to some periods and some phenomena), e.g., LOFA (or MCP-trip), LOFW, SBO, and LOOSP (respectively, present

Accident scenarios and phenomena in WCNR

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section, Sections 15.4.4–15.4.6). One may state that MCP-trip is related to the stop of one MCP and LOFA relates to the stop of all MCP. However, in the general cases the terms MCP-trip and LOFA are used interchangeably. The interest toward the Loss of Component Cooling Service Water (LCC/SW) derives from considering that the scenarios involve actions and events occurring at very uncertain times (OECD/NEA/CSNI, 2011). In this way a connection can be delineated in the present section between thermal-hydraulics and PSA. EOP are designed to account for MCP-trip or LOFA and LCC/SW.

Qualitative accident scenario: LOFA/MCP-trip and LCC/SW MCP-trip A partial LOFA is an AOO that may be caused by a mechanical or electrical failure in a pump motor, a fault in the power supply to the pump motor, or a pump motor trip caused by anomalies such as overcurrent or phase imbalance. The resulting decrease in reactor coolant flow, which occurs while the plant is at power, degrades the core heat transfer, and reduces the DNB margin. The following cases may occur: -

Partial loss of forced reactor coolant flow. Complete loss of forced reactor coolant flow. MCP rotor seizure. MCP shaft break.

RCS pressure increases and is limited by PRZ spray (actuation possible in case of availability of at least one MCP or of external pump). The margin to DNB is challenged. LCC/SW and PSA connection The loss of CCW and/or SW impacts the plant operation due to the loss of cooling of essential equipment, most notably, MCP and ECCS. When SW is lost, one of the consequences is the loss of CCW, although there are also some additional effects. Most notably, SW provides cooling water to DG, which become unavailable in loss of SW scenarios but not in loss of CCW due to other causes. Also, some containment systems like fan coolers can be affected by loss of SW but not by loss of CCW. The loss of SW has a frequency given by 1.88E-03 y-1 (Zion NPP, OECD/NEA/CSNI, 2011). Upon the loss of CCW/SW, the MCP is first affected. Even if the reactor coolant pumps are stopped, lack of water injection to the seals combined with loss of cooling water to the pump thermal barrier may lead to seal damage resulting in a seal LOCA (a particular case of small LOCA, conditioned by the loss of equipment). In the case that a seal LOCA occurs as a consequence of the loss of CCW, safety injection systems (SIS) are needed to compensate the loss of inventory. However, both HPIS and LPIS are unavailable while CCW is not recovered and ACC are the only available injection systems. Sequences without seal LOCA were not significant contributors to core damage since the intervention of the AFW system (failure probability 3.4E-05) was considered enough to prevent the exceedance of the cladding temperature limit, even without CCW recovery. The global frequency of sequences without seal LOCA and with AFW failure is about 5.05E-8 which is below the cut-off level. The analysis then focused on seal LOCA sequences.

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Thermal Hydraulics in Water-Cooled Nuclear Reactors

Following the seal LOCA event (originated by LCC/SW) a set of events or occurrences like manual scram, cooling of SG via SRV and CCW recovery, are considered together with assumptions like If the MCP is not stopped before high-temperature fluid reaches the seal package, all the seals are assumed damaged and reactor coolant is lost at the maximum rate. The failure of a sealing stage does not condition the failure of other stages. The failure of a sealing stage is assumed to occur simultaneously in all the four pumps.

-

Quantitative accident scenario: variable trends and TSE—LOFA/MCP-trip and LCC/SW MCP-trip Typical LOFA scenario can be derived from the TSE in Table 15.33, PWR with UTSG, and Fig. 15.61, EPR, taken from AREVA (2014) (row 20 in Table 15.1) and Fig. 15.62, VVER-440, taken from Groudev and Stefanova (2006) (row 24 in Table 15.1). It may be noted that the EPR information derives from an available licensing analysis (Fig. 15.61), and the VVER-440 time trends in Fig. 15.62 include comparison with measured data in the Kozloduy NPP unit 4. LCC/SW and PSA connection Typical LCC/SW scenarios, PSA relevant, can be derived from Fig. 15.63, PWR with UTSG, taken from OECD/NEA/CSNI (2011) (row 19 in Table 15.1). It was assumed that CCW system recovery occurs at the time of the seal LOCA, so that safety injection systems are available, if they do not fail for reasons different from LCC/SW. The depressurization of the SG SS, if not failed, is assumed to occur 600 s after the LOCA. Rather than providing a comprehensive documentation of the results (this can be found in the cited document), the main objective for Fig. 15.63 is to show the link between PSA and thermal-hydraulics.

Table 15.33 MCP-trip in PWR with UTSG: (selected) imposed and calculated events N

Event

I/R

1 2

MCP-trip occurrence Scram signal generation Scram occurrence MDNBR occurrence

I I

0 8

I R

8.5 9.0

Turbine trip Maximum PRZ pressure

I R

9.5 10

3 4 5 6

Time (s)

Notes

Failure of DG on demand and consequently of AFW Always greater than 1 (no CHF occurrence expected) Below PORV set-point

Accident scenarios and phenomena in WCNR

1075 2

120 DNBR (–)

1.8

Normalized flow rate (%)

100 80

1.6 1.4 1.2 1 0.8

S-29-IMPU

0

3

6 Time (s)

60

9

12

40 Core inlet Affected loop Unaffected loops

20 0 0

5

10

15 Time (s)

20

25

30

Fig. 15.61 Partial LOFA in EPR: flowrates in the primary loop and detail of DNBR time trend.

13

5500 Experim. data

12.8

Experim. data

5400 Water level (m)

Pressure (MPa)

RELAP calc.

12.6 12.4 B-2-COP

RELAP calc.

5300 B-2-COP

5200

12.2

5100

12 0

50

100 Time (s)

150

200

0

50

100

150

200

Time (s)

Fig. 15.62 Partial LOFA (trip of one MCP) in VVER-440. Left: PS pressure. Right: PRZ level.

Phenomena connection with accident scenario LOFA/MCP-trip and LCC/SW The phenomena listed in the second and the third columns of Table 15.34 are connected with LOFA/MCP-trip from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables in Figs. 15.61 and 15.62, as reported in the fourth column of Table 15.34. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible):

1076

Thermal Hydraulics in Water-Cooled Nuclear Reactors

2000 1800

Case 1 Case 2

1600

Case 3 Case 4

RST (K)

1400

Case 5 Case 6

1200

Case 7 Case 8

1000

Case 9 Case 10

800

Case 11

600 400 0

10,000 20,000 30,000 40,000 50,000 60,000 70,000 80,000 90,000 Time (s)

Fig. 15.63 LCC/SW sequences in PWR with UTSG: RST at PCT location from 11 PSA-related calculations.

Phenomena visualized by variables representative of the accident scenarios LOFA/MCP-trip and LCC/SW

Table 15.34

Phenomenon ID for: LOFA/MCP-trip and LCC/SW No.

Acronym

Description

Fig. no.

Notes

1

B-1-COH

15.49

See also SBLOCA

2

B-2-COP

15.62

3

S-29-IMPU

Condensation due to heat removal Condensation due to pressurization Impeller pump behavior

See also LOFW, SBO, and LOOSP Occurring in all accident scenarios

15.61

(1) Condensation due to heat removal (visualized, Fig. 15.49). The basic phenomenon occurs (at least) in all scenarios where SG SS removes thermal power from a PS where voids are present. All two-phase NC regimes satisfy this condition. Reflux condensing mode of NC and related flows in HL of UTSG PWR constitute an example of the effect of condensation. (2) Condensation due to pressurization (indirectly visualized, Fig. 15.62). The basic phenomenon occurs each time there is a pressure increase. Pressure increase is connected with PRZ level increase, typically in MCP-trip, LOFW, SBO, and LOOSP. (3) Impeller pump behavior (visualized, Fig. 15.61). Centrifugal pump performance constitutes a phenomenon occurring in all transients in all reactors (an exception could be an SMR without MCP). So-called homologous curves for centrifugal pumps, derived from experiments in one- and two-phase conditions, determine the predicted behavior of impeller pump.

Accident scenarios and phenomena in WCNR

1077

15.4.12.10 Turbine trip Turbine trip (TT) constitutes an expected (almost yearly) event during the life of NPP. It is not usually characterized with the word “accident.” Rather, the term AOO is adopted or even operational transient. The turbine is the key component of the BOP in an NPP. In this case “key” means the most expensive. Thermal efficiency of the plant is also affected by the turbine design. Materials of the wings and related stresses in nominal operation are the best available from current technology and the highest achievable, respectively. The turbine is installed on the same rotating axis as the alternator (i.e., an electromechanical component). Losing the load is a typical instantaneous process of electrical nets, where instantaneous means 103 s or less. In this situation, turbine over-speed may occur with mechanical damage. In order to protect the turbine a fast closing valve must be installed in the steam line upstream the machine. The fast closing of this valve is one main origin for thermal-hydraulics issues associated with turbine trip.

Qualitative accident scenario: Turbine trip Selected topics for turbine trip transients of interest in thermal-hydraulics are: A. Closing of turbine inlet valve after the “electric” event must be associated with fast opening of condenser dump valve. B. The opening of the condenser dump valve causes high-pressure (then, superheated) steam entering the condenser. Although condenser may have a capacity to remove thermal power higher than needed in nominal operating conditions, the entrance of superheated steam may locally decrease the condensation HTC. The associated thermal-hydraulic issue is addressed within the framework of the component design. C. The fast closing of the turbine inlet valve causes a positive pressure, wave which propagates upstream in the steam line. This causes condensation in the SG SS in a PWR and in the RPV of a BWR. Condensation in the SG causes a negative temperature perturbation for the PS coolant in a PWR, somewhat mitigated by the thermal inertia of the SG tubes and weakened by mixing before reaching the core. Otherwise condensation in the RPV of a BWR causes void collapse and has a direct impact on fission power which may exhibit an excursion after turbine trip. D. In the case of BWR, turbine trip is also associated with PSP temperature increase, because in the attempt to avoid a large decrease in thermal power production (to minimize xenon effects), SRV valves are open and deliver steam to the PSP. Liquid temperature in the PSP cannot increase above a certain threshold (typically below 80°C) to avoid endangering the function of ECC pumps in case these are called in operation. Thus, turbine can be kept in standby with core nearly at full power for a short-time period. E. Turbine axis mechanical inertia can be considered for a few seconds for keeping the electrical power supply to crucial RCS components like MCP. Noticeably, the Chernobyl event (Chapter 16) was originated by the planning of a test to confirm availability of electrical power supply after TT.

Quantitative accident scenario: variable trends and TSE—Turbine trip Typical turbine trip scenario can be derived from the TSE in Table 15.35, including PWR with UTSG and BWR and Fig. 15.64, SMR, taken from Haratyk and Gourmel (2015)

1078

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Table 15.35 Turbine trip in PWR with UTSG and BWR: (selected) imposed and calculated events N

Event

I/R

Time (s)

1 2

Turbine trip Turbine Isolation valve closure Condenser dump valve opening SRV opening Density perturbation reaching the core of BWR Density perturbation reaching the core of PWR Scram

I I

0 0.1

I

0.5

R R

1.0 1.0

R

2.0

I

10

4 5 6 7

Sec. system pressure (MPa)

Prim. syst. pressure (MPa)

Closure time in the order of tenths of second Opening time in the order of tenths of second

Not needed

8

17

16.5 S-12-CO4

16

15.5

6 8.5

4

2

Pressure (MPa)

3

Notes

0

20

40

60

Time (s)

80

100

S-12-CO4 S-12-CO4

7 6.5 6 0

0

15

8 7.5

0

10

20 Time (s)

2000

30

40

4000

6000

8000

10,000

Time (s)

Fig. 15.64 Turbine trip accident scenario in SMR. Left: PS pressure short term. Right: SS pressure, long term with focus on initial 40 s.

(row 44 in Table 15.1) and Figs. 15.65 and 15.66, BWR, taken from Bousbia-Salah and D’Auria (2002) (row 45 in Table 15.1). The turbine trip transient evolves in a few seconds, as from Table 15.35. Later on, specific EOP or operator actions affect the transient evolution (e.g., Fig. 15.64). In the case of BWR the positive pressure wave entering the RPV from the steam line propagates toward the core following two paths: (a) crossing steam dryers, then separator region, then upper plenum and entering the core from the top; and (b) entering DC downward separators, then lower plenum and entering the core from the bottom. The first path is shorter but wave propagation velocity is faster in the second path where liquid is present till core inlet. Both the two pressure waves contribute to the void collapse in the core and to the pressure rise. A three-dimensional RPV model is needed for predictions.

Accident scenarios and phenomena in WCNR

1079

7.3

300 I-30-VCP

250 Relative power (%)

Pressure (MPa)

7.2 7.1 Case 1 (base)

7

Case 2 Case 3

6.9

I-30-VCP

Case 1 (exp.) Case 2

200

Case 3

150 100 50 0

6.8 0

1

2

3

4

5

0

0.4

Time (s)

0.8 Time (s)

1.2

1.6

Fig. 15.65 Turbine trip accident scenario in BWR including measured NPP data. Left: RPV pressure with different assumptions about condenser dump valve history. Right: core power considering different condensation models.

7.3

Pressure (MPa)

7.2

0.0 s 0.2 s

7.1

0.4 s 0.6 s

7

1.0 s 2.5 s

6.9 6.8 6.7 6.6 0

20

40

60 80 Length (m)

100

120

140

Fig. 15.66 Turbine trip accident scenario in BWR. Pressure profile in Steam Line at time 0 (before the event) and times 0.2, 0.4, 0.6, 1.0, and 2.5 s after the event.

Phenomena connection with accident scenario: Turbine trip The phenomena listed in the second and the third columns of Table 15.36 are connected with Turbine Trip from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables in Figs. 15.64–15.66, as reported in the fourth column of the table. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) Condensation in stratified conditions—SG SS and BWR PSP (indirectly visualized, Fig. 15.64). Condensation in stratified conditions in SG SS and in PSP may occur in several

1080

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Phenomena visualized by variables representative of the accident scenarios during turbine trip

Table 15.36

Phenomenon ID for: Turbine trip No.

Acronym

Description

Fig. no.

Notes

1

S-12-CO4

15.64

PSP condensation in Fig. 15.37

2 3

I-25-SLD I-30-VCP

Condensation in stratified conditions— SG SS and BWR PSP Steam line dynamics Void collapse and temperature distribution during pressurization

15.66 15.65

Both left- and right-hand sides of the figure provide indirect visualization

scenarios, noticeably when pressure is increasing (SG SS) and SRV are discharging steam into the pool. Visualization in Fig. 15.64 is connected with the concerned phenomenon in SG SS. The PSP-related phenomenon is indirectly visualized in Fig. 15.37 and discussed with more detail by Wulff et al. (1992) (see also Chapter 16). (2) Steam line dynamics (visualized, Fig. 15.66). Steam line has a length of about 100 and a number of bends. Venturi nozzles and valves are installed in the line making complex the dynamics of the propagation of pressure wave. Reflection of pressure waves (geometric discontinuities and walls) is not within the capabilities of existing system thermal-hydraulic codes. (3) Void collapse and temperature distribution during pressurization (indirectly visualized, Fig. 15.66). The pressure and power pulse in the RPV are direct consequence or at the origin of void collapse.

15.4.12.11 Accidents during shutdown conditions NPP operate or are designed to operate the largest part of their life at full power conditions. However, nominal conditions include all needed status like start-up and shutdown. In those situations either decay power is produced or fission core power is varied to achieve decay power (shutdown period) or the full power (start-up procedure). Let us call all NPP nominal operational conditions not at full power as shutdown conditions. Then, shutdown conditions may arise at any pressure in the range from ambient pressure to full-nominal pressure and RCS may be tight (i.e., under pressure) or open to the containment with a variety of openings (i.e., RPV open for refueling). Owing to a number of reasons, not last the attention given to full power conditions, it has been clear since the 1990s that shutdown conditions may largely contribute to the overall risk of NPP or even may produce a contribution to risk higher than the full power operation. This started technological research in all areas including thermal-hydraulics. Before showing parameter time histories related to shutdown, let us note the following:

Accident scenarios and phenomena in WCNR l

l

1081

A key target (not the only one) for accident scenario calculations starting from full power is the demonstration of suitable design for ESF. For instance, in the case of LBLOCA, the analysis must demonstrate that the pressure of ACC and the volume of injected liquid are suitable to keep the RST below the licensing limit. This implied and implies the evaluation of phenomena like TPCF, CCFL in downcomer, and reflood, including NC during reflood and QF progression. A positive outcome of the thermal-hydraulic analysis is constituted by the demonstration that phenomena are understood, calculation is qualified and performance of the ESF is according to the design. In the case of shutdown conditions, the key target for the analysis of accident scenario is to evaluate what happens if one or more ESF fail. For instance, a typical target of calculation in the case of RPV open is to find the time when the core uncovers following a full stop of cooling water. The calculation does not need sophisticate computational tools and complex phenomena are not expected.

In the former case, thermal-hydraulic competences are needed to confirm the suitability of the concerned ESF or to improve its design. In the latter case, the only action possible is to improve the reliability of the concerned ESF (i.e., an action that has little or no connection with thermal-hydraulics). Furthermore, in some cases, accidents during shutdown periods are classified according to scenarios: noticeably, the event in Doel-2 (OECD/NEA/CSNI, 1988) is classified as SGTR, although it happened during a shutdown period. Nevertheless, shutdown conditions constitute a “fashion-interest” for thermalhydraulics.

Qualitative accident scenario: SHUTDOWN Any PIE and associated evolution like LBLOCA, SBLOCA, SGTR, MSLB, etc., may happen during shutdown. Phenomena expected during an assigned event starting at full power shall also be expected for the same event occurring during shutdown. For instance, the distinction between an SBLOCA occurring at full power and an SBLOCA occurring during shutdown is possible, but outside the context for the present chapter. Selected topics for shutdown transients of interest in thermal-hydraulics are: A. In some shutdown situations the barrier constituted by the RCS is not available and one may also imagine situations when both the RCS and the containment barrier are not available (to allow refueling, maintenance, etc.). This justifies the importance of shutdown transients for NPP risk studies. B. The evolution of some shutdown transients is expected at atmospheric pressure: this may put a challenge to some existing models. C. A shutdown transient may occur when more than one opening exists between RCS and containment. This implies a coupled containment-RCS calculation considering the possibility that natural circulation or natural convection motion established between the two openings. D. In case of stop in the operation of a service pump when the RPV is open, selected questions of interest to thermal-hydraulics are: - When flow is lost at core entrance? - When subcooling is lost at core outlet? - When core wide void appear? - When core uncovery occurs? - When RST reaches unacceptable and unrecoverable value after core uncovery?

1082

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Quantitative accident scenario: Variable trends—SHUTDOWN Typical SHUTDOWN scenarios can be derived from Fig. 15.67, PWR with UTSG, taken from Haste et al. (2010) and Fig. 15.68, APR1400, taken from Son and Shin (2007) (row 43 in Table 15.1 for both documents). Very different timings of events can be seen for the two considered transients: less than 1/2 h in Fig. 15.68 and more than 20 h in Fig. 15.68. In the former calculation, i.e., an LOCA during shutdown, the coupling between RCS and containment is evident. In the latter case the PSA-oriented calculation should be noted: the target is the evaluation of different actions (e.g., no actions, HPIS and Gravity Feed, GF, actuation) following a loss of flow. In the reported calculation two SG are unavailable, PZR head vent, reactor head vent, PRZ manway, and SG inlet manways are open. 0.4 0.35

Primary system

Pressure (MPa)

0.3

Containment (tot.)

A-9-HTCO

Containment (steam)

0.25 0.2 0.15 0.1 0.05 0 –200

0

200

400

600

800

1000

1200

1400

1600

Time (s)

Fig. 15.67 Shutdown accident scenario, LOCA in UTSG PWR: PS and containment pressures.

2000

0.15

Pressure (MPa)

0.14

HPSI press.

1750

GF press.

1500

No action temp.

0.13

0.12

HPSI temp.

1250

GF temp.

1000 750

I-2-ASY-D, Section 15.4.8

Temperature (K)

No action press.

I-23-SIP

500

0.11

250 0.1 0

2000

4000

6000

8000

10,000

0 12,000

Time (s)

Fig. 15.68 Shutdown accident scenarios, loss of cooling with recovery actions in APR1400. Left axis: PS pressure. Right axis: RST.

Accident scenarios and phenomena in WCNR

1083

RCS pressure increase when HPIS is adopted is due to pump head. However, HPIS tank is exhausted at a time well before emptying of the same tank by gravity. GF is better in a case where many openings are present. In different RCS configuration different use of available (water) resources are preferable. One expected problem occurring when recovery is attempted by gravity feed liquid is reported (by Son and Shin, 2007), when the number or the size of the RCS openings is not enough: even though the GF supplies RCS cooling water continuously, the amount of cooling water from the IRWST to the cold leg in each loop is not sufficient enough to prevent core damage because the pressurizer manway is not large enough to make large pressure difference to induce sufficient amount of cooling water from IRWST to the CL.

Phenomena connection with accident scenario: SHUTDOWN The phenomena listed in the second and the third columns of Table 15.37 are connected with SHUTDOWN from the cross-link process in Tables 15.3 and 15.4. Those phenomena are associated with variables in Figs. 15.67 and 15.68, as reported in the fourth column of the table. The following phenomena are outlined in this section or (visually) described through the use of calculated accident scenarios and related time trends (whenever possible): (1) HT condensation in containment structures, with or without noncondensable (indirectly visualized, Fig. 15.67). Pressure in containment in long-lasting transients is affected by heat transfer by condensation in the structures including the presence of noncondensable gases. (2) SG siphon draining (indirectly visualized, Fig. 15.68). Various openings in the PS cause a variety of emptying modes.

15.4.12.12 The nuclear fuel behavior The motivations for a specific “nuclear fuel” section have been given under item (f ) in Section 15.4.12. One may state that nuclear fuel and nuclear fuel performance constitute established technologies. Even a summary outline of those technologies (selected aspects are listed under item (f ) in Section 15.4.12) is well beyond the purposes here. Wide literature exists: fundamental materials-related issues can be

Phenomena visualized by variables representative of the accident scenarios during SHUTDOWN

Table 15.37

Phenomenon ID for: SHUTDOWN No.

Acronym

Description

Fig. no.

Notes

1

A-9-HTCO

15.67

Containment phenomenon

2

I-23-SIP

HT condensation in containment structures, with or without noncondensable SG siphon draining

15.68

Shutdown expected

1084

Thermal Hydraulics in Water-Cooled Nuclear Reactors

found in OECD/NEA/NSC (2015), and analyses involving transient nuclear fuel behavior are discussed by Adorni et al. (2011), Rozzia et al. (2011), and Lisovvy et al. (2015).

Nuclear fuel topics and facts Selected nuclear fuel topics/facts which have a relation with thermal-hydraulics are: A. The geometrical configuration of nuclear fuel rod, namely diameter and height (including the distance TAF to BAF) is strictly associated with linear power production (q0 , kW/ m), other than thermal neutrons motion and mean-free path inside the pellet. Related to (current) optimized pin design, an increase in diameter causes a decrease in the volumetric power density of the core, and a decrease in diameter creates a challenge for the structural integrity. B. Linear power production (q0 ) controls the maximum fuel temperature in nominal operating conditions, then the energy stored in the fuel. Therefore, PCT in case of LBLOCA and namely what is called blowdown-PCT is directly affected by q0 . C. Burn-up has a complex influence upon material, mechanical, thermal, and reactor physics properties of fuel pins. Burn-up increase above certain limits undermines the fuel capability to remain intact following events like CRE or LOCA. D. Formation of oxide and crud on the rod external surface increases RST in normal operation. The increase in LOCA PCT associated with oxide and crud can be of the order of 200 K. E. Ballooning is the main expected mechanism for clad rupture in case of LBLOCA. Other than release of (primarily) gaseous FP into the coolant, ballooning causes obstruction in the flow inside an FA. Obstruction may be at the origin of inadequate cooling for neighboring pins (see e.g., Ammirabile and Walker, 2014). F. Brittle rupture of fuel clad must be avoided and H2 produced by the chemical reaction between clad material and water must be controlled during any DBA accident. PCT and time of quench directly affect brittle fracture and H2 production. G. Spacer grid design and number have large influence in the interaction between coolant and fuel in nominal operating conditions and during accidents.

Nuclear fuel is part of any NPP calculation.

Phenomena connection with accident scenario: Nuclear fuel Specific nuclear fuel information can be derived from Fig. 15.69, taken from CANDU fuel-related experiments (Horhoianu et al., 1998) (row 11 in Table 15.1). Results from irradiation tests (energy deposition up to 265 cal/g) in research reactor performed in stagnant water at room temperature are shown in Fig. 15.69. Tests included the effects of initial element internal pressure and a range of energy deposition on the fuel element behavior. Cladding failure mechanism and the failure threshold have been established: the fuel failure mechanism is a burst type and is very similar to LOCA failure mechanism. The phenomenon identified in the second and the third columns of Table 15.38 is the outcome of the cross-link process in Tables 15.3 and 15.4. The phenomenon is associated with variables in Fig. 15.69.

Accident scenarios and phenomena in WCNR

1085

2.5 Rod initial pressure: 1,7 MPa energy deposition: 150 cal/g

850

2

750

Temperature

1.5

Pressure

650 550

1

I-29-NTF2

450 0.5

350 250

Rod internal pressure (MPa)

Cladding surface temp. (K)

950

0 0

0.5

1

1.5

2

Time (s) Clad diam. directional def. (mm)

13.5 Rod initial pressure: 0.2 MPa energy Deposition: 95 cal/g

13.4 13.3 I-29-NTF2

13.2 13.1 13 0

20

40

60

80

100

120

140

Axial position (mm)

Fig. 15.69 Nuclear fuel behavior. Top: clad internal pressure and RST. Bottom: clad diametric deformation along three axial directions Table 15.38

Visualization of nuclear fuel behavior

Phenomenon: Nuclear fuel No.

Acronym

Description

Fig. no.

Notes

1

I-29-NTF2

Thermal-hydraulics— Nuclear fuel feedback

15.69

Occurring in all NPP transients

The following phenomenon is outlined in this section and (visually) described through the use of calculated accident scenarios and related time trends: (1) Thermal-hydraulics—Nuclear fuel feedback (visualized, Fig. 15.69, top). Feedback between coolant/moderator thermal-hydraulic parameters and nuclear fuel occurs in any NPP transient. The snapshot outline at items A. to G. gives an idea of the interactions.

1086

15.5

Thermal Hydraulics in Water-Cooled Nuclear Reactors

Outcomes from the performed process

15.5.1 Resources for the cross-link between phenomena, accident scenarios, and time histories Accident scenarios in NPP are understood and described by the use of parameter (or variable or quantity) time trends (or time histories), which are typically calculated by system thermal-hydraulics codes (discussed in Chapter 10). Phenomena have been derived to prove the qualification level of those computer codes. A process has been completed in this document aiming at connecting accident scenarios, phenomena, and parameters resulting from NPP calculations. Items which are part of the present chapter of the book can be summarized as follows: l

l

l

l

l

l

l

l

l

12 water cooled reactor designs: PWR-UTSG, PWR-OTSG, PWR-HOSG (VVER-440 and VVER-1000), BWR, CANDU, PHWR, RBMK, plus AP1000, APR1400, EPR, SMR, and ABWR. 23 main accident scenarios (from Table 15.1) and connected sections in Chapter 4. 7 accident scenarios-related topics (e.g., NC, CHF, Boron, PTS, Fuel, AM, Containment). 45 + 2 accident scenarios [rows] and 47 + 21 leading reference documents in Tables 15.1 and 15.5 (“+” related to Table 15.5), to connect reactor designs and accident scenarios. 113 phenomena in Table 15.2. 5085 (¼45  113) boxes in Table 15.3 cross-connecting accident scenarios and reactor designs with phenomena. 15 sets of “homogeneous” variable time trends, resulting in about 30 individual time trends distinguished in the top row of Table 15.4. 1695 (¼15  113) boxes in Table 15.4 cross-connecting accident scenarios and phenomena with parameters. 158 documents part of the list of references, including 68 “process leading” documents cited in Tables 15.1 and 15.5.

One might also note the uncomfortable (for the reader) list of acronyms, including about 180 items, which was used to describe the process. The process starts with the development of Tables 15.1 and 15.2, respectively, from: (a) considering reactor designs and DBA analysis results: namely, Table 15.1 includes a list of accident scenarios supported by referenced NPP calculations; and (b) the list of phenomena already part of Chapter 6 and available from OECD/NEA/CSNI and IAEA devoted reports.

Table 15.3 is developed by cross-connecting the accident scenarios, part of Table 15.1, and the phenomena, part of Table 15.2. The “filling of Table 15.3” provides a demonstration that all phenomena are part of at least one accident scenario. Moreover looking at Table 15.3 one may note that LBLOCA and SBLOCA are the accident scenarios where the largest number of phenomena is expected to occur. The parameters used to characterize the accident scenarios are connected with phenomena in Table 15.4. Following the process of “filling of Table 15.4,” some phenomena (i.e., those listed in Table 15.5) could not be characterized by suitable parameters

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derived from NPP calculations. Thus Table 15.5 was developed including time histories referring to accident scenarios but not part of NPP calculations: available experimental and design analysis data were considered. The data in Tables 15.1–15.5 are at the origin of Chapter 4: the sections connected with accident scenarios in Table 15.1 are considered. As already mentioned in the Foreword, the process of connecting calculated variables and phenomena could be done using a couple of accident scenarios instead of 47 scenarios, having access to full calculation details (e.g., in-house calculation). Such a simplified process does not include the description of scenarios covering the DBA envelope and does not give an idea of the interest toward accident scenarios by the scientific community.

15.5.2 Key achievements The main outcome from the performed activity is a vision for the nuclear thermalhydraulics universe mentioned in Chapter 1. This is based on phenomena, reactor design features, and results from the calculations by system thermal-hydraulics codes: connection with the licensing and the PSA processes comes from the nature of considered NPP calculations (i.e., the DBA envelope). A confirmation of the complexity of nuclear thermal-hydraulics shall be associated with the main outcome. What else can be derived from the performed study? Phenomena (i.e., expected in case of accidents in water cooled nuclear reactors) have been identified and characterized, also as a result of Chapter 6. They are derived from experiments and expertise which are continuously updated. Variables, results of computer code application to NPP analyses, have been associated with phenomena: whether the phenomena are properly modeled to produce a calculated accident scenario constitutes a question mark not addressed by the study. Then the achievements are: (1) Each qualified calculation should imply the demonstration that the concerned phenomena are properly modeled: i.e., a parameter time histories can be calculated not necessarily based upon phenomena part of the used models. (2) Ranking of phenomena as proposed in some approaches to accident analysis is not an outcome from the study: rather, all identified phenomena must be modeled according to their best knowledge.

One may also state that the phenomena-scenarios cross-link process is one part of activity needed to ensure the qualification of a calculation.

15.5.2.1 Application of system thermal-hydraulics codes A few notes are given below in relation to the application of system thermal-hydraulic codes to the accident analysis of NPP considering the content of other chapters of the book. First, the connection between modeling and phenomena in system thermalhydraulics is not part of the present chapter: this is discussed in Chapters 5–10.

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Second, the systematic relationship between phenomena (identified in Chapter 6 and partly characterized in the present chapter) and model features constitutes the topic of code manuals which also include the demonstration of code predictive capabilities. Third, the listed phenomena constitute the basis for code validation, as discussed in Chapter 13. Fourth, the study summarized in this chapter allows the confirmation that system thermal-hydraulics codes constitute the repository for the expertise associated with considered phenomena and the best (unique) tool to calculate accident scenarios. Finally, the list of technology significant achievements from the application of system thermal-hydraulics codes to the accident analysis in water cooled reactor (not given here) might complement the present chapter as well as Chapters 11 and 13. This may also constitute a suitable prerequisite for the BEPU framework discussed in Chapter 14.

15.6

Conclusions

Two main objectives have been pursued in this chapter: (a) the description of accident scenarios; and (b) the association of thermal-hydraulics phenomena to accident scenario. Reference NPP units at the basis of the analyses documented in this chapter are PWR-UTSG, PWR-OTSG, PWR-HOSG (VVER-440 and VVER-1000), BWR, CANDU, PHWR, RBMK, AP1000, APR1400, EPR, SMR, and ABWR. The accident scenarios are part of the nuclear reactor safety technology restricted, within the present framework, to the DBA or to the situation “before loss of core integrity.” The thermal-hydraulics phenomena derive from OECD/NEA/CSNI and IAEA documents: the phenomena already described in Chapter 6 have been considered. In some cases more than one phenomenon deals with similar transient evolutions of two-phase flows, or in different terms, duplication in the definition of a few phenomena may have occurred. This might be connected with modeling features needed for establishing predictive capabilities and/or different validation framework. Notes are added to point out similarities, rather than reducing the number of phenomena. The accident scenarios are depicted based on results of system thermal-hydraulics code calculations: related time histories are associated with phenomena. No emphasis is given in the process to the quality of the calculations as well as to the needed evaluation of uncertainty. However, due importance is given to experimental programs: among the others, data or information from BC, PKL, LSTF, ATLAS, PSB, and LOBI experiments, in addition to data measured in NPP (e.g., Doel-2 SGTR) have been considered. The main outcomes can be summarized as follows: l

l

A global vision of nuclear reactor thermal-hydraulics is provided from the side of phenomena and accident scenarios: 12 reactor types are considered for the characterization of 47 accident scenarios cross-linked with 113 phenomena: the phenomena are largely affected by geometry and boundary conditions and need suitable modeling capabilities (not discussed in the present chapter). The knowledge and the understanding of phenomena is a prerequisite for performing meaningful accident analysis. However, the process of associating time histories and phenomena

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does not include the demonstration that the phenomenon is part of the modeling or, in case it is part of the modeling, no quality proof is ensured.

The expected impact of the document is a contribution to the knowledge and the knowledge management in nuclear system thermal-hydraulics: the complexity of the cross-link process may give an idea of the complexity of subject (i.e., highly system geometry and boundary conditions dependent). However, the gathered information can be used as a part of the qualification process for system code calculations. All of this may constitute a guidance to formulate and to address the following issues or questions in relation to each NPP accident scenario calculation: a. What are phenomena expected to be relevant in the scenario? (The list in Table 15.2 to be considered: in general terms all phenomena should be considered.) b. Are any of the phenomena expected to be relevant in the scenario part of the adopted calculation model? For instance, in case of SBLOCA with two-phase conditions occurring in the hot leg of a PWR, importance should be given to the phenomenon “liquid and vapor mixing with condensation in SG mixing chamber.” Namely what equations (or equation parameters) are developed to account for those phenomena? c. What are the qualification bases for the phenomena expected to be relevant? Namely, what experiments may be used to demonstrate a suitable knowledge for the phenomenon including addressing the scaling issue?

The distinction between a code calculation and a qualified code calculation should involve the answer (or the capability to answer) the questions/issues a. to c.

Acknowledgments The authors wish to acknowledge the invaluable effort made by A. Capperi from University of Pisa for having drafted most of the figures in the text. He had to become aware of the connection between accident scenarios and phenomena.

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