MELCOR sensitivity studies for a low-pressure, short-term station blackout at the Peach Bottom plant

Nuclear Engineen.'ng Nuclear Engineering and Design 152 (1994) 287-317

ELSEVIER

andDes,gn

MELCOR sensitivity studies for a low-pressure, short-term station blackout at the Peach Bottom plant J u a n J. C a r b a j o Engineering Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-8057, USA

Abstract

This paper summarizes the results of analyses performed to assess the effect of a variety of design parameters and

operational procedures on a station blackout severe accident at the Peach Bottom Atomic Power Station. The severe-accident MELCOR code, version 1.8.1 was used in these analyses. The following sensitivity studies were completed: effect of the automatic depressurization system actuation timing on the accident progression; effect of fuel and cladding porosities on vessel failure and containment failure times; effect of several parameters on the amount of in-vessel steel ejected into the cavity after vessel failure; effect of different parameters on vessel penetration failure time; vessel failure timing; and lower plenum shroud and core shroud temperatures. These sensitivity studies provided valuable insights into the MELCORcode behavior and into the progression of this severe accident. The most significant results are: (a) the optimum steam cooling of the core is accomplished when the automatic depressurization system is actuated when the core water level is at one-third of the active core height, delaying vessel failure by minutes and containment failure by hours, (b) vessel failure is significantly delayed (by 2 11) when lower-plenum debris quenching is included in the model, and (c) the core shroud melts during this transient.

1. Introduction Thermal-hydraulic sensitivity studies have been performed for a low-pressure, short-term, station blackout at the Peach Bottom Atomic Power Station, Unit 2, with the severe accident code MELCOR (Sandia, 1991), version 1.8.1. Six sensitivity studies were conducted to access the impact of several input parameters on the t h e r m a l hydraulic accident results. These six sensitivity studies are: (a) effect o f the automatic depressurization system (ADS) actuation timing on accident progression (given in Section 4); (b) effect of fuel and cladding porosities on vessel failure and con-

tainment failure times (Section 5); (c) effect of several parameters on the amount of in-vessel steel ejected into cavity after vessel failure (Section 6); (d) effect of several variables on penetration failure timing (Section 7); (e) reactor vessel failure timing (Section 8); and (f) lower-plenum and core shroud temperatures (Section 9). The severe accident analyzed is a low-pressure, short-term, station blackout. It is a low-pressure transient because the ADS is used to depressurize the reactor vessel. It is a short-term, station blackout transient because it is assumed that all the a.c. and d.c. power is lost (except for the d.c. power needed to actuate the ADS and the safety

0029-5493/94/$07.00 © 1994 Elsevier Science S.A. All rights reserved SSD1 0029-5493(94)000795-Z

288

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

relief valves) and no emergency cooling systems (ECCS) can inject water into the reactor vessel. Therefore, without water injection, all the core water is rapidly vaporized, and core meltdown and vessel failure occur in a short time. The MELCOR severe-accident code (Sandia, 1991) versions 1.8.1 HN and KH were used to perform these sensitivity studies. MELCOR is a fully integrated computer code that models the progression of severe accidents in nuclear power plants using light water reactors (LWRs). The MELCOR code is being developed and enhanced by Sandia National Laboratories (SNL) for the US Nuclear Regulatory Commission (NRC). MELCOR is the successor of the Source Term Code Package (STCP) (Gieseke, 1986). The entire spectrum of severe-accident phenomena, including vessel and thermal-hydraulic response, core heatup, core degradation and relocation, and radionuclide release and transport is treated in MELCOR. Thermal-hydraulic behavior is modeled in MELCOR using control volumes and flowpaths. The version 1.8.1 of MELCOR used in these studies considers neither core debris quenching by the water in the lower plenum after core meltdown nor debris heat transfer to other structures in the lower plenum. Consequently, vessel failure is predicted almost immediately after core-plate failure and after core debris enters the lower plenum. An enhanced model for debris interactions in the lower plenum has been developed by Oak Ridge National Laboratory (ORNL) and has been interfaced with MELCOR to form the MELCOR/BH (Bottom Head) package (Hodge, 1993; Hyman, 1993a,b). Calculations using the MELCOR/BH package are also presented. This package models core-debris quenching in the lower plenum, and is believed to provide more realistic estimates for the timing of vessel failure. The an package has been incorporated into version 1.8.2 of MELCOR after this analysis was completed. Results from other calculations completed in the past (Hyman, 1990) for the same accident sequence at the same plant (Peach Bottom) using the BWRSAR (Ott, 1989; Hodge, 1990b) and CONTAIN (Washington, 1991) codes are also presented. The BWRSAR code, developed at ORNL, is a predecessor of the aH package and used models for debris interaction in

the lower plenum similar to the ones used in the BH package. A description of the Peach Bottom plant is presented in Section 2. A description of the MELfOR model used is presented in Section 3. The six sensitivity studies are presented in Sections 4-9. Final conclusions are presented in Section 10.

2. Description of Peach Bottom Atomic Power Station

2.1. General plant description The Peach Bottom Atomic Power Station, Unit 2, located in Delta, PA, is a boiling water reactor (BWR) with a Mark-I containment referred to as an inverted light-bulb (drywell) and torus (wetwell), as illustrated in Fig. 1. The drywell and wetwell or torus are made of steel, with an internal design pressure of 0.485 MPa (56 psig). The estimated failure pressure is 1.2 MPa (159 psig), significantly higher than the design pressure. Containment failure criteria and containment failure modes are discussed in Section 2.3. The reactor coolant system design is the BWR-4. The thermal power is 3293 MW(t) and the net electrical output is 1065 MW(e), with a power density of 50.7 kW 1-~. The core contains 764 fuel bundles and 185 control rods. The free volumes in the drywell and wetwell are 4777 m 3 (169 500 ft 3) and 3170 m 3 ( 112 000 ft3), respectively. The total volume of the wetwell (torus) is 7132 m 3 (252000 ft3). About half of this volume is occupied by water. The primary containment structure in the Peach Bottom plant is enclosed by a reactor building. Leakage from the primary containment may be filtered by the standby gas treatment system (SGTS) in the reactor building before being released to the environment. In the station blackout studied in this paper, it is assumed that the SGTS is not operational. The secondary containment (reactor building) has five different floors. The lowest floor is the building basement where the torus room is located. The highest floor is the refueling bay. The safety relief valves (SRVs) and the ADS play an important role in the sequence of events

289

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

LOCATION OF CONTAINMENT FAILURE MODE 2 AT THE HEAD FLANGE SEALS

BUILDING

LOCATION OF CONTAINMENT FAILURE MODE 3 AT THE LINER

- - D R Y WELL

LOC~TiON OF CONTAINMENT FAILURE MODE 1 AT THE WETWELL

MU!UION POOL

Fig. 1. Peach Bottomplant, with containment failuremodes. of the station blackout. The SRVs vent steam from the reactor vessel into the wetwell. Each SRV is located on a steam header attached to the main steam lines leaving the reactor vessel. The Peach Bottom plant has a total of 11 SRVs and two spring safety valves. Data for the SRVs are shown in Table 1. The 11 SRVs have different opening and closing pressures; they open automatically when the opening pressure is reached. They also close automatically when the closing pressure in the vessel is reached. SRV No. 1 opens at the lowest pressure (7.687 MPa or 1115 psia); SRV No. 11 opens at the highest pressure (7.867 MPa or 1141 psia). The eleven SRVs are distributed into three banks of four, four, and three SRVs each, respectively. The two remaining spring safety valves have an automatic opening pressure of 8.555 MPa (1245 psia). Consequently,

the spring safety valves will only open at high pressures after all the SRVs are already open. The spring safety valves close at a low pressure of 7.260 MPa (1053 psia). The SRVs can also be opened manually at a pressure below the automatic set-point. Manual operation of the SRVs is mandated in the Peach Bottom plant-specific operating procedures to avoid automatic operation of the same SRV (the one opening at the lowest pressure) releasing steam to the same point of the wetwell (which results in unequal heating). These operating procedures keep the vessel pressure between 6.49 and 7.18 MPa (945 and 1045 psia) by opening a different SRV each time. By doing this, a more even wetwell water temperature distribution is obtained. In these procedures, the SRV is manually opened at a pressure of 6.49 MPa (945 psia),

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

290

Table 1 The Peach Bottom safety relief valve characteristics Valve No.

I 2 3 4 5 6 7 8 9 10 11

Bank No.

1 1 1 1 2 2 2 2 3 3 3

Opening pressure

Closing pressure

Rated flow

Rated pressure

Rated denfity

[MPa (psia)]

[MPa (psia)]

[kg s -I (lb h-I)]

[MPa (psia)]

[kg m -3 (lb ft-3)]

7.687 7.708 7.722 7.756 7.763 7.791 7.797 7.825 7.846 7.860 7.867

7.253 7.101 7.184 6.991 7.053 7.322 7.184 7.246 7.391 7.115 7.308

109 109 109 109 110 110 110 110 Ili 111 111

7.722 (1120) 7.722 (1120) 7.722 (1120) 7.722 (1120) 7.791 (1130) 7.791 (1130) 7.791 (1130) 7.791 (1130) 7.860 (1140) 7.860 (1140) 7.860 (1140)

40.8 40.8 40.8 40.8 41.2 41.2 41.2 41.2 41.6 41.6 41.6

(1115) (1118) (1120) (1125) (1126) (1130) (1131) (1135) (1138) (1140) (1141)

(1052) (1030) (1042) (1014) (1023) (1062) (1042) (1051) (1072) (1032) (1060)

a value well below the automatic opening point of the first SRV, 7.687 M P a (1115 psia). Therefore, manual operation of the SRVs will prevent pressure-initiated automatic operation of the SRVs. SRV actuation releases the steam produced in the vessel by the decay heat of the fuel. Since water is being vaporized in this process, the water level in the vessel drops. Eventually, the water level in the vessel will drop below the top of the core. Once the core is uncovered, core heatup and core damage will occur. 2.2. The Automatic Depressurization System (ADS) The ADS opens simultaneously five SRVs (Nos. 1, 2, 3, 5, and 6 of Table 1) that discharge symmetrically into and around the torus below the suppression pool water level. Steam released through the five SRVs of the ADS depressurizes the reactor. Vessel depressurization by ADS actuation can be used to delay fuel heatup by "steam-cooling" uncovered portions of the core by the steam produced in the depressurization. The timing of the initiation of this depressurization for steam cooling is very important in the accident progression. I f the depressurization is initiated when the core is still covered with water, or when the uncovered portion of the core is not very hot, vaporization

(865000) (865000) (865000) (865000) (873000) (873000) (873000) (873000) (880000) (880000) (880000)

ADS

(2.55) (2.55) (2.55) (2.55) (2.57) (2.57) (2.57) (2.57) (2.60) (2.60) (2.60)

yes yes yes no yes yes no no no no no

of the coolant will not provide effective cooling. In this case, some of the water inventory will be lost during the depressurization without benefit. On the other hand, if depressurization is initiated later in the transient when the core is uncovered and very hot, the flow of steam from depressurization m a y increase Zircaloy oxidation (with the "run-away" Z i r c a l o y - s t e a m reaction) increasing the fuel and cladding temperature instead of providing cooling. During a short-term station blackout, all coolant injection systems and all a.c. power are lost. Emergency procedure guidelines (EPGs) outline the sequence of operations to be performed during an emergency situation. The last revision of these EPGs, Rev. 4 (General Electric, 1987), specifies steam cooling of the core to be initiated via ADS actuation when the core water level is about two-thirds of the active core height (ACH). A previous revision of the EPGs, Rev. 3 (General Electric, 1982), indicated that during the same emergency situation, steam cooling of the core should be initiated when the core water level is about one-third of the ACH. Revision 3 of the EPGs has a two-step maneuver for ADS actuation. First, when the reactor vessel water level has fallen to one-third of the A C H , one SRV is manually opened. As a result of this open SRV, the vessel pressure decreases. When the vessel pressure reaches 4.9 M P a (715 psia), the ADS is

J.J. Carbajo /Nuclear Engineering and Design 152 (1994) 287-317

actuated and the reactor vessel is depressurized, resulting in steam cooling. The time difference between these two steps (one SRV open and ADS actuation) is very small, typically only a few minutes. It was concluded in a previous study (Greene, 1991; Hodge, 1992) using the aWRSAR code (Ott, 1989; Hodge, 1990b) that ADS actuation at the time indicated by Rev. 3 of the EPGs delays the accident progression relative to the earlier ADS actuation indicated by Rev. 4 of the EPGs. The benefits of the later ADS actuation were identified as: (a) later initiation of core relocation, (b) less in-vessel hydrogen production, and (c) longer time to vessel failure. On February 14, 1990, ORNL presented these results to the N R C and the BWR Owners Group (BWROG). The BWROG is currently considering incorporation of these findings in future EPGs revisions (Thadani, 1990). Revision 5 of the EPGs is anticipated to be similar to the previous Rev. 3 of the EPGs, but with only one step for ADS actuation (the initial SRV actuation will be eliminated) when the core water level is at one-third of the ACH. This one-step ADS actuation is used in the MELCOR model in this study in anticipation of the future release of Rev. 5 of the EPGs. Furthermore, this single-step ADS actuation is also a good approximation of Rev. 3 of the EPGs. The effect of the ADS actuation timing on the accident progression has been studied with the MELCOR code and the result of this study is presented in Section 4. 2.3. Containment failure modes There are basically three different failure modes of the Peach Bottom Mark-I containment. These failure modes are: (a) by high-pressure-induced primary containment liner failure, (b) by hightemperature degradation of the drywell head flange seals, and (c) by liner melt-through (or creep rupture) when hot debris in the cavity contacts the drywell liner. The first failure mode (high pressure inside the primary containment) has historically been considered to be the dominant primary containment failure mode. There have, however, been different opinions on the pressure and location of this

291

failure. The Reactor Safety Study (US Nuclear Regulatory Commission, 1975) indicated that the most likely location of primary containment overpressurization failure would be in the torus air space (wetwell) region of the containment at a pressure of 1.2 MPa (160 psig). After containment failure, a path from the wetwell to the torus room (Fig. 1) in the basement of the reactor building would be established. The torus room would then leak into the reactor building and from there into the environment. A more recent structural analysis by the Chicago Bridge and Iron Company (1987) performed for the specific geometry of the Peach Bottom Mark-I containment, indicated that the primary containment will fail by high pressure in the upper region of the wetwell at a high pressure of 1.2 MPa (159 psig). This recent result agrees well with the previous result of the Reactor Safety Study (US Nuclear Regulatory Commission, 1975). This failure pressure and location have been used for the "containment failure mode 1" of this study. The resulting breach opening area has been assumed to be 0.1 m 2 (1 ft2). The second mode of primary containment failure in Mark-I containments is by failure of the drywell head flange seals, as confirmed by analyses performed by Chicago Bridge and Iron Company (1987) for the Peach Bottom Mark-I containment. The containment fails in this mode by a combination of high temperature and moderate pressure. The high temperature degrades the seals first, then the moderate pressure lifts the head from the drywell and leakage can occur. The Idaho National Laboratory and Sandia National Laboratories have conducted thermal performance tests of different seals used in BWRs (Bridges, 1987; Brinson, 1988). The experiments indicated that these seals sustain a permanent loss of their elasticity and structural integrity when subjected to temperatures above 371°C or 644 K (700°F). According to Chicago Bridge and Iron (1987), once the seals lose their structural integrity, leakage will start when the pressure differential across the seals reaches 0.565 MPa (82 psid). These analyses were performed for the actual geometry of Peach Bottom. The actual flow area available for leakage through this path is a

J.J. Carbajo / Nuclear Engineering and Design 152 (I 994) 287- 317

292

function of the drywell pressure. It is assumed that the open flow area increases linearly from 0 m 2 (0 ft 2) at 0.565 MPa (82 psid) to 0.04 m 2 (0.43 ft2) at 1.378 MPa (200 psid). As already explained, once the seals are degraded by high temperature, leakage will occur if the drywell pressure is above 0.565 MPa (82 psid), even if the temperature of the seals drops below the degrading temperature of 644 K (700°F). Finally, the third failure mode of the primary containment is by drywell liner failure due to contact with the molten debris in the cavity. The liner may fail by either melt-through or creeprupture. The melting temperature of the carbon steel liner is about 1810 K (2800°F). A lower temperature can produce liner creep-rupture. There are many studies on Mark-I containment liner melt-through. A number of these studies have been reviewed (Speis, 1985; Greene, 1985; Weingardt, 1988; Kress, 1988; Sienicki, 1988; Theofanous, 1991). Theofanous et al. (1991) state that when the cavity is dry, the probability of liner melt-through after vessel failure is between 0.63 and 1.0. The same reference states that if the cavity is flooded after vessel failure, the probability of liner melt-through is less than 10 - 4 . Nevertheless, there are many variables with high uncertainties that influence the outcome of this event. The occurrence of liner melt-through results from very complex phenomena that are difficult to quantify. This containment failure mode has not been implemented in the MELCOR model of this study because of all these uncertainties and because containment failure was not one of the main subjects studied in this paper.

There is, however, another study on Peach Bottom (Carbajo, 1993a) using a detailed, multi-node containment model with the three containment failure modes implemented.

3. MELCOR model of the Peach Bottom plant

The MELCORmodel employed in this study uses a total of 19 control volumes, six of which correspond to the reactor vessel, 12 correspond to the primary and secondary containment, and one volume models the outside environment. The six volumes of the reactor vessel are described in Table 2. Fig. 2 shows the volumes and the flowpaths connecting the volumes in the reactor vessel. Fig. 2 shows three different flow paths connecting the steam dome of the reactor vessel (Vol. 360) with the wetwell (Vol. 200). These three flowpaths correspond to different modes of operation of the safety relief valves (SRVs). Flowpath 362 models the automatic operation of the SRVs. Flowpath 363 models the ADS, that opens five SRVs simultaneously when the water level in the core reaches a predetermined level. Finally, flowpath 364 models the manual operation of one SRV to avoid cycling of the same SRV during automatic operation. Manual operation of the SRVs is started 200 s after transient initiation. This time lapse of 200 s between manual operation of the SRVs and blackout initiation is arbitrary, but it is a reasonable time since the operators will not start the SRV actuation immediately after the blackout initiation. The 13 volumes of the primary and secondary containment and the environment are described in

Table 2 Reactor vessel volumes and initial conditions Volume No.

310 320 330 340 350 360

Volume name

Downcomer (annulus) Lower plenum By-pass Channel Shroud dome Steam dome

Volume

Water level

Temperature

Pressure

[m 3 (ft3)]

[m

[K (°F)]

[MPa (psia)]

183.8 103.5 25.8 33.3 44.9 218.6

14.125 (556.1) Full Full Full 14.125 (556.1) Empty

560 560 560 560 560 560

7.134 7.134 7.134 7.134 7.134 7.134

(6491) (3655) (911) (1176) (1586) (7720)

(in)]

(549) (549) (549) (549) (549) (549)

(1034.7) (1034.7) (1034.7) (1034.7) (1034.7) (1034.7)

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

/ [

3611

aRn STEAM'DOME

o~,, SRVsAUTO

~

~ 363 AI)S

--

293

TO VOLUME200 (WETWELL)

I 350 SHROUDDOME

TO VOLUMElOO (DRYWELL)

370 LEAKAGE

312

I

320

]

TO VOLUME100 (IN-PEDESTALDRYWELL)

Fig. 2. Vo|umes and flowpaths in the reactor vcsse|.

Table 3 and shown in Fig. 3. The primary containment is modeled using three volumes: one for the drywell (with the cavity at the bottom), one for the wetwell and one for the vent legs connecting the drywell and the wetwell. Flowpaths 398 and 400 model two containment failure modes. Flowpath 398 models the head flange seals failure (containment failure mode 2) and connects (after containment failure) Vol. 100 with Vol. 408, the refueling bay. Flowpath 400 models containment failure mode 1 by high pressure at the wetwell (torus air space) and connects (after containment failure) Vol. 200 with Vol. 401, the torus room. All the calculations presented in this paper used a dry cavity (without water) and all resulted in containment failure mode 2, at the head flange seals.

The secondary containment is represented by nine volumes (Fig. 3) at five different floors. A total of 35 flowpaths are used; ten of them have variable flow area (valves) modified by control functions. A total of 34 control functions, 30 tabular functions and 68 heat slabs are used in this model. The core and lower plenum are nodalized with three radial rings and 11 axial levels (with a total of 33 nodes) as shown in Fig. 4. Five axial levels are in the lower plenum, one is the core plate region and the remaining five levels are the active core region (with a total of 15 nodes). High burnup fuel (end of life) after the reactor has been operating at full power for a long time is considered in these calculations. Total masses used are: fuel, 168 480 kg (371 435 lb); Zircaloy, 61 708 kg

294

d.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

Table 3

Volumes used in the primary and secondary containment Volume No.

Volume name

Volume [m3

100

Drywell

150 200 401 402 403 404 405 406 407 408 409 410

Drywell legs Wetwell Torus room South 135 level North 135 level South 165 level Remain 165 level South 195 level Remain 195 level Refuelingbay Turbine building Environment

(ft3)]

4235 (149557)

565 (19953) 7133 (251880) 5426 (191612) 5154 (182011) 5154 (182011) 2356 (83201) 7066 (249532) 2452 (86591) 4866 ( 171840) 31048 (1096444) 148825(5255677) 10t° (3.5 x 1011)

Volume gas

Volume liquid

Temperature Pressure

[m 3

[m3 (ft3)]

[K (°F)]

[MPa (psia)]

1.47 (52) 21.8 (770) 3961 (139881) 0 0 0 0 0 0 0 0 0 0

336 (145) 336 (145) 306 (90) 306 (90) 301 (82) 301 (82) 300(80) 300(80) 300(80) 300 (80) 300 (80) 300 (80) 300 (80)

0.108 0.108 0.101 0.101 0.101 0.101 0.101 0.101 0.101 0.101 0.101 0.101 0.101

(ft3)]

4233.5 (149505)

543 ( 19183) 3172 (112018) Full Full Full Full Full Full Full Full Full Full

(136 043 lb); steel (which can melt and relocate), 121 190 kg (267 202 lb); and poison (boron carbide), 1785 kg (3935 lb). The total mass of Zircaloy includes the cladding (35 425 kg) and the canisters (26 283 kg). The total mass of steel includes the mass of the penetrations (2112 kg), the mass of the structures and control rod guide tubes in the lower plenum (70 243 kg), the core plate mass (15 000 kg), the steel mass of the control rods (9680 kg), and the mass of the top guide (24 145 kg). Three penetrations through the lower head are used. The three penetrations divide the lower head into three radial rings as the core, core plate, and lower plenum regions. Each penetration will fail by high temperature when hot debris from each radial ring contacts the penetration. Other steel present in the vessel that does not melt and relocate (like the vessel itself, the shroud, the steam separators, and the dryers) is input as heat slabs. The shroud that separates the downcomer from the core and lower plenum is modeled with eleven heat slabs, one at each axial node of the core, core plate, and lower plenum (Fig. 4). Additional details of this model can be found in Carbajo (1993b). Additional details about MELCOR and its models can be found in Sandia National Laboratories (1991).

(15.7) (15.7) (14.7) (14.7) (14.7) (14.7) (14.7) (14.7) (14.7) (14.7) (14.7) (14.7) (14.7)

4. ADS actuation timing As described in Section 2.2, the timing of the ADS actuation to depressurize the vessel and to provide steam cooling to the hot core is very important in the station blackout accident progression. During the first 200 s of the MELCOR calculation, one SRV opens and closes automatically. After 200 s, manual operation of the SRVs is started (as described in Section 3). Once the water level in the core reaches the predetermined level, the ADS is manually actuated and the vessel is depressurized. Revisions 3 and 4 of the E P G s (General Electric, 1982, 1987) mandate ADS actuation at different core water levels. This study includes a base calculation with no depressurization, Case 1, and four calculations with ADS actuated at four different times. The actuation times for each of these four calculations are: Case 2, when the collapsed water level is 0.3 m ( 12 in) below the b o t t o m of the active fuel (BAF) (the core is completely uncovered and the water level is in the lower plenum); Case 3, when the core is just uncovered (collapsed water level at exactly BAF); Case 4, when the collapsed water level is 1.3 m (50 in) above the BAF or one-third of the active

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

295

410 ENVIRONMENT 416 413 411

VESSEL

362 SRVsAUTO

195 LEVEL

¢'
363 AUS

407

364 SRV MANUAL

LEAKAGE FLOWS

165 LEVEL

402

'VESSEL ,BREACH

100 DRYWELL

ISo CAVITY

TORUSROOM

Fig. 3. Volumes and flowpaths in the primary and secondary containments..

core height (ACH) (in accordance with EPGs, Rev. 3); and Case 5, when the collapsed water level in the core is 2.5 m (100 in) above the BAF, equivalent to two-thirds of the ACH (as in EPGs Rev. 4). The results are presented in Table 4. These results are evaluated on: (a) time when core oxidation starts, (b) amount of hydrogen produced in-vessel, (c) time when core relocation starts, and (d) vessel and containment failure times. Emphasis is placed on vessel phenomena rather than on containment phenomena. The results of Table 4 indicate that Case 4, with depressurization when the collapsed water level in the core is 1.3 m (50 in) above BAF (about one-third of the ACH), is the MELCOR case that

yields the longest time to initiation of core oxidation and relocation, and to vessel and containment failure. The amounts of in-vessel hydrogen produced at vessel failure and at containment failure are also the smallest for Case 4. This case provides the optimum steam cooling of the core of all the cases analyzed here. All these cases are for cores with high burnup fuel. The times calculated by MELCOR in Case 4 for vessel failure and containment failure are shorter than the ones calculated in Hyman (1990) for the same short-term station blackout at the same plant using the BWRSAR (Ott, 1989; Hodge, 1990b) and CONTAIN (Washington, 1991) codes. The results of Hyman (1990) are on the last line of Table 4. MELCOR 1.8.1 does not consider debris

,

~

..~ P~ ~

,

-a

ro

I'O

o

o

¢.O

¢.0

4~

108 102 67 53 80

-0.3 0 1.3 2.5 1.3

--

(50) (100) (50)

(--12)

(m) (in)

Core level BAF a

0 1/3 2/3 1/3

--

Fraction ACH b

116 156 165 151 255

114

Vessel failure (min)

a BAF, bottom active fuel. Level zero is the BAF. b ACH, active core height. c Results from Hyman (1990).

CONTAIN c

BWRSAR/

--

1

(min)

At time

ADS actuated

2 3 4 5

Case No.

8 54 98 98 175

--

(min)

Time between ADS actuation and vessel failure

41 41 41 41 41 40

Core starts uncovering (min)

O1

O~

-q

00

SLABS

~D

O

t~ \ \ \ N K \ \ \ \

O

102 102 102 84 74 81

77 77 77 89 79 95

103 103 104 117 109 131

751 750 580 100 275 504

[kg

(1656) (1653) (1279) (220) (606) (1111)

(lb)]

(lb)]

1421 (3133) 1420 (3130) 950 (2094) 325 (717) 325 (717) 737 (1637)

[kg

Complete Core Core In-vessel H 2 produced at core oxidation relocation uncovering starts starts Vessel Containment (rain) (min) (rain) failure failure

HEAT

~ \ \ \ \ x ¢q ~ \ \ \ \ " k \ \ \ \ ~ \ \ \ x

SHROUD

I N \ \ "'f ~ \ \ \ N 1 X \ \ \ N X \ \ \ \ N

~

Table 4 MELCOR results for different ADS actuation times on short-term station blackout sequences

~

~.~ ~

4:51 4:56 6:06 8:18 6:02 9:53

Containment failure (h:min)

1

,.4 I

¢

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

297

Table 5 Comparison of BWRSARand MELCOR results for core uncovering with different ADS actuation times during a station blackout BWRSAR

Water level at

Time (min)

Time ADS depress required to as EPG uncover (% ACH) each third of the core (min)

Top core (100%) 71%

40 48.2

10

2/3 core 1/3 core

(66%) (33%)

50 70

20

30% Bottom core 0%

80 102

TOTAL core

32

Rev. 4 (71%)

results

Complete core uncovering (min)

50.4

Rev. 3 (30%) 81 No 102

Time diff. Fuel Core plate Vessel H 2 between ADS relocation failure failure generated actuation and (min) (min) (min) at vessel complete core failure uncovering (kg) (min) 2.2

102

106.4

209

531

1.0 --

131 87.4

132 116.7

255 187.4

504 859

62 MELCOR 1.8.1 results

Water level at

Time (min)

Top core

41

2/3 core

53

1/3 core Bottom core TOTAL Core

67 102

Time ADS depress required to as EPG uncover (% ACH) each third of the core (min)

Complete core uncovering (min)

Time diff. Fuel Core plate Vessel H2 between ADS relocation failure failure generated actuation and (min) (min) (min) at vessel complete core failure uncovering (kg) (min)

12 14 35

Rev. 4 (66%)

74

21

109

150

151

275

Rev. 3 (33%)

84

17

117

163

165

100

102

--

103

110

114

751

No 61

s o m e m i n o r differences in the A D S a c t u a t i o n times b e t w e e n the BWRSAR a n d the MELCOR c a l c u l a t i o n s for each E P G s revision. I n the BWRSAR c a l c u l a t i o n for Rev. 4 o f the E P G s , the A D S was a c t u a t e d at 71% o f the A C H , while in the MELCOR c a l c u l a t i o n , the A D S was a c t u a t e d at exactly t w o - t h i r d s (67%) o f the A C H . A l s o , in the BWRSAR c a l c u l a t i o n for Rev. 3 o f the E P G s , the A D S was a c t u a t e d at 30% o f the A C H , while in the MELCOR c a l c u l a t i o n the A D S was a c t u a t e d at e x a c t l y o n e - t h i r d (33%) o f the A C H . T h e results o f T a b l e 5 indicate t h a t the BWRSAR c a l c u l a t i o n using Rev. 3 o f the E P G s yields also the longest time (like MELCOR) to vessel failure (255 min) o f the three BWRSAR c a l c u l a t i o n s p r e s e n t e d in T a b l e 5.

BWRSAR c a l c u l a t e d times to vessel failure are always l o n g e r t h a n MELCOR c a l c u l a t e d times to vessel failure for the same transient. This is because the MELCOR c o d e calculates very s h o r t time differences b e t w e e n c o r e plate failure a n d vessel failure, as the h o t d e b r i s does n o t q u e n c h in the l o w e r p l e n u m water. T h e BWRSAR l o w e r - p l e n u m d e b r i s m o d e l allows d e b r i s q u e n c h i n g a n d yields l o n g e r times to vessel failure ( a f t e r c o r e - p l a t e failure) t h a n d o e s MELCOR version 1.8.1. T i m e s f r o m A D S a c t u a t i o n i n i t i a t i o n to c o m plete core u n c o v e r i n g are m u c h s h o r t e r in BWRSAR ( b e t w e e n 1 a n d 2.2 min) t h a n in MELCOR ( b e t w e e n 17 a n d 21 min). This is because in MELCOR, after A D S a c t u a t i o n , the d e p r e s s u r i z a -

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

298

tion raises the water level in the core as water from the lower plenum flows into the core. Also, after depressurization, the lower-plenum liquid water starts boiling and its level decreases while there is still water in the core volumes above. The lower-plenum volume is no longer full of water while the core volume above (Fig. 2) still contains water. This is a consequence of the MELCOR control volume modeling approach. If water in the core volume is transferred to instantly fill voids of the lower plenum volume, complete core uncovering will occur sooner in MELCOR, and MELCOR and BWRSAR calculated times for complete core uncovering will be more similar. Column 2 of Table 5 shows that the times to begin core uncovering are very similar for both codes (40 and 41 min for BWRSAR and MELCOR respectively). The times to full core uncovering without ADS actuation are the same for both codes (102 min). Therefore, the times from top to

I.

~

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7.5

i

i

bottom of the core uncovering are very similar for both codes (62 and 61 min). The times to uncover the bottom one-third of the core are the longest and the times to uncover the top one-third of the core are the shortest of the three thirds of the core for both codes. These results indicate that vessel depressurization during a short-term station blackout in a BWR-4 should be initiated when the water level in the core is about one-third of the active core height in order to optimize steam cooling of the core. This depressurization timing (as per EPGs Rev. 3) yields the longest times to core oxidation initiation, core relocation and vessel and containment failure. The time to vessel failure is longer by 14 min than the vessel failure time calculated using Rev. 4 of the EPGs. Under severe-accident conditions, any time gained before vessel failure is very important. The containment failure time was calculated for the case using EPGs Rev. 3 (Case 4), longer by over

J

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MANUAL SRV OPERATION

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5

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AUTOMATIC OPERATION

.8

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J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

2 h than for the case using EPGs Rev. 4 (Case 5). Also, Case 4 resulted in the least amount of in-vessel hydrogen produced (both at vessel and containment failure times) of all the cases analyzed in Table 4 (with up to 1100 kg less than Case 1, the case without depressurization). Some graphical results of these calculations are presented in Figs. 5-12. Fig. 5 shows the pressure inside the reactor vessel (specifically in Vol. 360, the reactor dome) for Case 4 (the case with the optimum cooling of the core). The initial vessel pressure of 7.134 MPa (1034.7 psig) at time 0 increases until it reaches the automatic opening set-point of the first SRV at about 7.7 MPa (1115 psia). The SRV closes when the pressure decreases to 7.25 MPa (1052 psia). After five complete SRV cycles (open and close) of automatic operation, one SRV is opened manually at 200 s. Manual operation of the SRVs keeps the pressure in the reactor vessel between 6.43 and 7.13 MPa (930

8

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299

and 1030 psig). There are 28 cycles (open and close) of this manual operation of the SRVs before the ADS is actuated. At 3993 s (67 min of calculations), the level in the core reaches onethird of the active core height and the ADS is actuated, resulting in a pressure reduction. Fig. 6 depicts the pressure inside the vessel during the complete calculation of Case 4. Fig. 7 shows the collapsed water levels for three volumes inside the reactor vessel: downcomer (Vol. 310), core channel (Vol. 340), and lower plenum (Vol. 320). After the ADS is actuated, the water level in the core first increases owing to water that flows in from the lower plenum after depressurization. Afterwards, the core water level decreases to the bottom of the core volume owing to coolant vaporization. Vessel failure occurs at 9871 s (165 min); the water level in the lower plenum drops to zero after vessel failure as the lower-plenum contents are emptied. The water level in each vessel

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TIME (103s) Fig. 6. Reactor vessel pressure for Case 4 of Table 4.

t,o

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

300

16

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600

DOWNCOMER CORE

14

LOWER

PLENUM

500

12 400

DOWNCOMER WATER LEVEL

"-'10

=

.==

> 8

[-

.,300

1/3 CORE ACTIVE HEIGHT

ADS / 7 ACTUATED

6

CORE WATER LEVEL

BOTTOM ACTIVE FUEL _200

O

4

VESSEL FAILURE

_100 LOWER PLENUM WATER LEVEl

2

I

0 0

5

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l

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10

l

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15

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0

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20

25

TIME ( 1 0 3 s )

Fig. 7. Collapsed water levels inside the vessel for Case 4 of Table 4.

volume varies between the top and the bottom of the volume. When the core volume is empty, its water level is at the bottom of the active fuel, which is also the top of the lower plenum. The downcomer volume has some water trapped around the jet pumps even after the vessel fails and the lower plenum is empty. This water slowly vaporizes and the downcomer water level slowly decreases with time after vessel failure. For comparison, Fig. 8 shows the same water levels in the core for Case 5 (with depressurization when the water level is at two-thirds of the active core height). Vessel failure was calculated to occur at 9030 s (151 min), earlier (by 14 min) than in Case 4. Fig. 9 shows vessel water levels for Case 3 of Table 4 (depressurization when the core water level is at the bottom of the active fuel, at about 102 min). Vessel failure occurs at 9350 s (156 min).

Fig. 10 shows the core water levels for Case 1 of Table 4 (no depressurization). The vessel fails at 6820 s (114 min), shortly after the core is completely uncovered (which occurs at about 6250 s). This vessel failure time is the shortest of all the times calculated for vessel failure in Table 4. Fig. 11 shows the core-plate temperatures for Case 4. The core plate is divided into three rings (as explained in Section 3). Temperature increases after 6000 s are due to hot debris from the core (either molten steel or fuel) relocated to the core plate. Ring 1 fails first, followed by ring 2 and ring 3. The first two rings also melt after vessel failure. Ring 3 does not melt. The temperature decrease after 3993 s is due to ADS actuation. Fig. 12 shows the temperatures of the three vessel penetrations for Case 4. The effect of the ADS depressurization can also be seen at 3993 s. After the core plate fails (Fig. 11), debris from each ring

~ [-¢

301

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317 16

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-600

DOWNCOMER CORE LOWER

PLENUM

.

DOWNCOMER WATER LEVEL

-500

12 "~'10

.~j.~ TOP .......2/3COREACTIVE HEIGHT .....

~"ADS

~

2

-

0

-300

i

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O------ COREWATERL :~t _EVEL

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-I00

LOWER PLENUM WATER LEVEL

I

0

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10

5

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15

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TIME (103s) Fig. 8. Collapsedwater levels inside the vessel for Case 5 of Table 4.

falls to the corresponding penetration. Once the debris reaches the penetration, the penetration heats up very rapidly and fails.

large value of the porosity(p = 0.99) produces a large packed volume of a component, 100 times the original volume Iio: Vpack = Vo/(1 -- 0.99) = 100Vo

5. Fuel and cladding porosities MELCOR has a user-defined input called "porosity" to define the "effective" or "packed" volume of a component (Vp~k). The porosity, p, is employed to calculate the packed volume from the component "original" volume, Vo, by the following formula:

Vpa~k= Vo/(1 --p)

(1)

The porosity can vary between 0 and 0.99. A zero value of the porosity does not modify the original volume of a component: Vp~¢k= 1Io. A

(2)

The porosity can be used with intact components (fuel and cladding) and with debris. When a large value of the porosity is used with the intact components of the core (fuel and cladding) the resulting large packed volume will occupy all the "free" volume of the core and no relocation of debris from above is possible. Therefore, the core will support debris regardless of the intact component mass. The large packed volume of the intact components only affects relocation of debris from above; the original free volume between components is still open for flow calculations. Similarly, when a large porosity is used with the debris, a

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

302

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~-~

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600

0

DOWNCOMER

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5OO 12 10

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r"t

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i

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4

FAILURE

ACTUATED

100 2 0

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2

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4

I

6

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t

8

10

TIME ( l O $ s ) Fig. 9. Collapsed water levels inside the vessel for Case 3 of Table 4.

large effective packed volume of debris results that will occupy or spread out over large free volumes in the core or lower plenum. Recommended values of the porosities are 0.99 for the intact components (fuel and cladding) in the core. The porosities of the intact components do not apply t o the lower plenum since there is neither fuel nor cladding present in the lower plenum. A porosity o f 0.3 for the particulate debris in the core and lower plenum is commonly used. The use of a porosity value of 0.99 for the core intact components is adequate to prevent debris relocation but is not realistic. Based on the original volumes of the intact components and on the free volumes between components of the Peach Bottom core it was calculated that a realistic value for the core intact component porosity is 0:53. This porosity value allows free space for reloca-

tion of debris into the space between fuel assemblies, but not into the region between fuel pins, because the grid spacers are assumed to block the debris. If relocation inside the fuel pins is desired (assuming that the grid spacers have failed) a intact component porosity around 0.40 should be used. Table 6 shows the results of two runs performed for the station blackout with MELCOR 1,8.1 using two different core intact component porosities: 0.99 and 0.53. Both runs used ADS actuation at one-third of the ACH. The differences between the two runs are not significant. Run 2, with the lowest core porosity (0.53) yielded vessel failure 1 min earlier and containment failure 16 min later. Therefore, these two calculations indicate that the core intact component porosities do not significantly affect the vessel and containment failure times.

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317 16

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DOWNCOMER CORE L O W E R PLENUM STEAM SEPARATORS

0

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(103s)

Fig. 10. Collapsedwater levels insidethe vessel for Case 1 of Table 4. The core intact component porosity of 0.53 was chosen for use in the remaining calculations of this paper because it is a more realistic value than 0.99. The effect of the debris porosity is covered in the next section.

flow areas through the lower plenum, and lowerplenum surface areas. Finally, calculations using lower-plenum debris bed models with the MELCOR/BH package (Hodge, 1993; Hyman, 1993a,b) and the BWRSAR(Ott, 1989; Hodge, 1990b) codes are also presented and compared to the MELCOR 1.8.1 results.

6. Parameters that affect the amount of in-vessel steel ejected into the cavity after vessel failure

6. I. Lower-plenum steel mass

This section describes the results of several studies conducted to investigate the effect of certain parameters on the amount of in-vessel steel ejected after vessel failure. The amounts of invessel steel and fuel ejected influence containment failure time and source terms. The parameters investigated are: mass of steel in the lower plenum, debris porosity, support flags in the lower plenum, lower-plenum structural failure temperature, open flow area through the core plate, open

Two different amounts of in-vessel steel were used: a low inventory of 69 824 kg (153 962 lb) and the actual inventory of 121 180 kg (267 202 lb). The results of this study are summarized in Table 7. All these calculations were performed for the station blackout with ADS actuated at onethird of the ACH and with core intact component (fuel and cladding) porosities of 0.53. The first ten runs of this table were performed with MELCOR 1.8.1. Run 11 used the MELCOR/BH package

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

304

1

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(Hodge, 1993; Hyman, 1993a,b) and the last run (Run 12), is the combined result from the BWRSAR (Ott, 1989; Hodge, 1990b) and CONTArN (Washington, 1991) codes taken from Hyman (1990) and Hodge et al. (1990a). Run 1, with the low inventory of steel inside the vessel, predicted that 62% of the initial steel inventory (43 149 kg) was ejected to the cavity. Run 2, with the large inventory of steel in the lower plenum, predicted only 9.9% (11 980 kg) of the initial steel inventory ejected to the cavity. These results, unexpected at first, are because the additional steel in the lower plenum of the run with the large steel inventory (Run 2) acts as a heat sink to the relocated molten steel from the core above. With more steel present in the lower plenum, more molten steel from the core refreezes on the lower plenum structures and, therefore, less steel leaves the vessel as debris. The MELCOR code models the core heatup, melting and relocation process in a separate way

for the uranium dioxide fuel and for the steel inside the vessel. After the core dries out, heat is transferred from the fuel rods to the steel of the control rods primarily by radiation. The steel of the control rods is the first material that melts inside the core. When this steel melts, it candles down into the steel nodes below. This process continues until the molten steel reaches the core plate, where the steel piles up on the core plate itself. The molten steel does not refreeze in the core region above the core plate when the fuel heat source is present. After the core plate fails or melts, the molten steel candles down into the steel node below (part of the control rod guide tubes). Depending on the temperatures and masses of the molten and solid steel, some of the solid steel in the node below the core plate may be melted by the molten steel coming from the core region, or, some of the molten steel may refreeze on this steel node. The molten steel will candle down into the steel node

305

J.J. Carbajo /Nuclear Engineering and Design 152 (1994) 2 8 7 - 3 1 7

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MELTING TEMPERATURE 12600°For 1700K)

2.5 6

4

2.0 F FAILURE TEMPERATURE (1832°F or 1 2 7 3 / ~ i.

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VESSEL FAILURE

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t.J-I

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a.

PENETRATION

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PENETRATION

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TIME (10 ~s) Fig. 12. Penetration temperatures for Case 4 of Table 4.

below. This candling process continues until all the steel refreezes, or until some molten steel reaches the bottom of the lower plenum, where it can be ejected as debris through failed penetrations. MELCOR treats the fuel separately from the steel. After core-plate failure or meltdown, fuel debris on top of the core plate falls down directly to the bottom of the vessel. At the bottom of the vessel, the fuel debris heats up the penetrations, which fail in a very short time after they contact the fuel debris. No energy is transferred from the fuel debris to the lower-plenum structures or molten steel when there is water present in the lower plenum, because heat can only be transferred by radiation in MELCOR 1.8.1. The molten steel from the core region can only reach the bottom of the vessel by candling down. This is not realistic. If the initial amount of steel in the lower plenum is not very large (compared to the molten

steel from the core region), the lower-plenum steel may be melted by the molten steel coming from above, and both the core steel and the lowerplenum steel will be melted and available to be ejected from the vessel. On the other hand, if the inventory of steel in the lower plenum is large, the molten steel from above may be insufficient to melt it. In fact, some of this molten steel may refreeze on the steel nodes in the lower plenum. Consequently, if a large inventory of steel is present in the lower plenum, the lower-plenum steel may not be melted by the molten steel from above, and less molten steel may reach the bottom of the vessel where it can be released as debris. This situation may be different if the additional steel is added to the core region, above the core plate. This additional steel will be melted by the fuel decay heat and, in this case, more molten steel will be available for melting the steel in the lower plenum and for ejection after vessel failure.

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

306

Table 6 Core intact component porosity results for the station blackout sequence Case No.

1 2

Core porosities

0.99 0.53

ADS actuated (min)

66 66

Vessel Time between failure ADS actuation and (min) vesselfailure (min)

161 160

95 94

Core complete uncovering (min)

86 86

For the Peach Bottom sequence analyzed here, a larger inventory of steel in the lower plenum resulted in less molten steel being ejected from the vessel. These results demonstrate the MELCOR 1.8.1 code's inability to transfer decay heat from the fuel debris to the structures in the lower plenum when the lower plenum contains water (as is the case in these calculations). The MELCOR core package only allows energy to be transferred from the debris to the structures in the lower plenum by radiation when there is no water present. In a more realistic scenario, some decay heat should be transferred from the fuel debris to the structures in the lower plenum, even if water is present. Several attempts to model this more realistic scenario were pursued by varying some MELCOR input parameters and are presented in Sections 6.2-6.5. Basically, all of these runs focused on an attempt to force the fuel debris and the lower-plenum structures into contact so that energy would be transferred into these structures. It is precisely this issue that has led to the development of the MELCOR/BH version of MELCOR (now incorporated into version 1.8.2 of MELCOg).

6.2. Debris porosity The porosity of the debris was increased from 0.3 to 0.9 in the MELCOR input. This large debris porosity results in a very large packed volume of fuel debris (as explained in Section 5). Fuel debris will occupy all the open space at all the levels of structures in the lower plenum instead of falling directly to the bottom head or lower levels. En-

Core oxidation starts (min)

88 88

Core relocation starts (min)

117 117

In-vessel H2 produced at Vessel failure

Containment failure

kg (lb)

kg (lb)

105 (231) 405 (893) 100 (220) 425 (937)

Containment failure (min)

431 447

ergy will be transferred by radiation from the fuel debris to the lower-plenum structures not covered with water, and more steel should melt and reach the bottom of the vessel for release. This approach affected the results significantly, as shown in Run 3 of Table 7. At the end of the calculations (500 min), 54% of the initial steel inventory in the vessel (core and lower plenum) was released into the cavity vs. only 9.9% for Run 2. Larger debris porosities resulted in significantly larger amounts of steel ejected into the cavity.

6.3. Lower-plenum support flags and structural failure temperature A new approach assumed that each level of the lower plenum supports debris (as the core plate does) until the failure temperature of 1273 K (1832°F, the default value) is reached. The melting temperature of stainless steel is 1700 K (2600°F). This new approach was accomplished by setting the input for the logical support flag at all levels of the lower plenum to 11. Normally, the support flag is set to 11 only at the support core plate. The support flags at the other levels of the core and lower plenum are set to 00, ~ t h the exception of the second axial node of the lower plenum, which is set to 01. The second axial node of the lower plenum needs the support flag at 01 for the structures of this level to stay in place without any structures in the node below. The results of Run 4, with each level of the lower plenum supporting debris, did not show any increase in the amount of steel melted and ejected from the vessel when compared to Run 2. In fact,

(lb)]

69824 (153962) 121190 (267202) 121190 (267202) 121190(267202) 121190(267202) 121190(267202) 119068 (262545) 119068 (262545) 119068 (262545) 119068 (262545) 119068 (262545) 121190 (267202)

[kg

Steel inside the vessel

0.3 0.3 0.9 0.3 0.3 0.3 0.9 0.9 0.9 0.9 0.9 --

Debris porosity

01,00 01,00 01,00 11 11 00 00 00 00 00 00 --

Support flags in lower plenum (°F)]

1273 (1832) 1273 (1832) 1273 (1832) 1273 (1832) 1700(2600) 1273 (1832) 1273 (1832) 1273 (1832) 1273 (1832) 1273 (1832) 1273 (1832) --

[K

165 163 163 193 329 153 163 191 163 163 285 255

Failure Vessel temperature failure for supports (rain)

498 415 366 414 656 412 374 394 361 376 598 593

Containment failure (rain)

333 252 203 221 327 259 211 203 198 213 313 338

(rain)

Time difference, containmerit and vessel failure

325 (716) 450 (992) 708 (1561) 629 (1387) 679 (1497) 1132 (2496) 545 (1202) 527 (1162) 568 (1252) 624 (1376) 600(1323) 737 (1632)

[kg (lb)]

H2 produced at containment failure (lb)]

163072 (359512) 152974 (337308) 144739 (319149) 142513 (314241) 144229 (318025) 168300(371101) 144738 (319093) 144608 (318860) 165369 (364638) 149033 (328618) 115668 (255048) 157550 (547400)

[kg

UO2 ejected

97 91 86 85 86 99.9 86 86 98 88 67 98

(% initial inventory)

(lb)]

43149 (95127) 11980 (24416) 63859 (140809) 6600 (14553) 105361 (232321) 120500 (265702) 61679 (135979) 25057 (55251) 24758 (54591) 63212 (139382) 120000(264600) 159935 (352657)

[kg

Stainless steel ejected

62 9.9 54 5.5 87 99.9 52 21 21 53 100 132

(% initial inventory)

(lb)]

261920 (577435) 199810 (440581) 251230 (553962) 156530 (343149) 291500(642757) 361340(796755) 249100(549171) 218590(481991) 228930 (504791) 255410 (563179) 279520 (616342) 384871 (848642)

[kg

Total debris ejected

a Run 8 has reduced core-plate open area and reduced lower-plenum surface areas. b Ran 9 has only reduced lower-plenum surface areas (most realistic values). This run has the most realistic MELCOR input. c Run 10 has reduced lower-plenum surface areas for tings 1 and 3 only (larger values for ring 2). d Ran 11 used the MELCOR/BH package with the same input values of Ran 9. This is the most credible run. Results from Hyman (1990) and Hodge et al. (1990a) using the codes BWRSAR and CONTAIN. This run has a larger amount of steel ejected (132%) than the MELCOR runs because it also included part of the vessel bottom head that had melted.

1 2 3 4 5 6 7 8a 9b 10 ¢ 11 d 12e

Run No.

Table 7 Steel inside the vessel a n d ejected in the s t a t i o n b l a c k o u t sequence

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J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

Run 4 released less steel than Run 2. This was because the structures in the lower plenum failed at 1273 K (1832°F) - - long before any melting occurred at 1700 K (2600°F). Once the structure at a level fails, the fuel debris is released and no further heating of that structure takes place. However, Run 4 calculated a longer time to vessel failure since the debris took a longer time to fail the five levels of structures in the lower plenum and to reach the penetrations and the bottom of the lower plenum. This run was repeated with the failure temperature increased to the melting temperature of steel (1700 K) in Run 5. The results of Run 5 predict that as much as 87% of the initial inventory of steel is released to the cavity. Also, this run predicted vessel failure at 329 min (19 762 s), much later than any other previous run. Longer times to vessel failure also resulted in longer times to containment failure. The fuel debris required a much longer time to fall through the five different levels of structures in the lower plenum and reach the lower head penetrations. In addition to melting and releasing significant amounts of steel, at the time of vessel failure (329 min), all the water in the lower plenum was boiled off by the decay heat of the fuel. Previous runs predicted vessel failure much earlier, when there is still a significant mass of water in the lower plenum (about 2.7 m (8.85 ft) above the bottom of the vessel). Another approach was used in Run 6. In this Run 6, the steel structure in the lower plenum was collapsed onto the vessel bottom before the fuel debris reached the lower head. (This approach forced the hot fuel debris to heat the steel in the lower plenum at the same time that the penetrations were heated.) This was accomplished by changing the support flag in the second axial node of the lower plenum from 01 to 00. (The structures need a support flag set to 01 to stay in place without structures below.) This run (Run 6) predicted that, at the end of the calculation, almost all the contents of the vessel including steel, control rods and fuel were released to the cavity. However, vessel failure was predicted to occur earlier than in any other run. Core-plate failure and melting occurred earlier than in previous runs because there was no structure under the core plate

to act as a heat sink for the core debris. After core-plate failure, the debris heated both the penetrations and the steel in the lower-plenum bottom simultaneously. Therefore, MELCOR'Slower-plenum support flags and structural failure temperatures have a very significant effect on the amount of steel ejected and on vessel failure and containment failure times. Run 7 has a slightly smaller amount o f steel inside the vessel than previous Runs 2-6. The total amount of steel inside the core and lower plenum, 121 190 kg (267 202 lbs), was split into 2112 kg (4657 lbs) for the vessel penetrations and the remaining 119 068 kg (262 545 lbs) are input as structural steel inside the vessel available for melting and debris ejection. This was done because the mass of the penetrations is not available for expulsion in MELCOR from the reactor vessel after vessel failure. The only way to eject the penetrations' mass is by adding it to the first axial node of the lower plenum. This results in a duplication of the penetration mass. Since this is undesirable, the mass used for the penetrations should be a small mass and it should be subtracted from the mass of steel in the axial node 1 of the lower plenum. In Run 7, the lower-plenum steel mass was distributed among all the five nodes of the lower plenum, including the first axial node. The support flag for all the axial nodes of the lower plenum was 00, as there are intact structures below each axial level to support the structures above. This Run 7 is very similar to Run 3. The differences between the two runs are in the total amount of steel in the lower plenum and the way this steel is distributed. As expected, the results from Run 3 and 7 are very similar. Thus, the mass distribution of steel in the lower plenum has a minor effect on the results. 6.4. Core-plate open area and lower-plenum surface areas

The first seven runs of Table 7 used an open area at the core-plate axial node of 10.38 m 2 (111.73 ft2). Run 8 is similar to Run 7 but used reduced values for the open area in the core plate (2 m 2 or 22 ft 2) and reduced values for the surface areas of the structures in the lower plenum. The

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317 results of this run indicate that vessel and containment failure times occur later, and a smaller amount of steel is ejected to the cavity (only 21% in Run 8 vs. 54% in Run 7). The smaller open area at the core plate reduced the amount of debris that could relocate onto the core plate, and its failure occurred later. This open area is multiplied in MELCOR by the axial node height to calculate the volume available for relocated material. The core-plate axial node of the MELCOR input is 0.277 m (0.9 ft) in height, and it includes the core plate and some structures on top of the core plate (below the active fuel). Consequently, the open flow area through the core plate (which is actually 2 m 2) is not the appropriate value to be used for this input, since this axial node extends from the bottom of the core plate to the bottom of the active fuel region. The actual "open flow area" through the core plate is used in the input of the flowpaths connecting the lower plenum with the core volumes (Fig. 2). The "open area" for the core-plate axial node is actually the area that, multiplied by the height of the axial node, will result in the "open free volume" for this axial node. The value that should be used for this open area is 10.38 m 2 (111.73 ft2). Run 9 used smaller lower-plenum structure surface areas, and larger open area at the core plate axial node than Run 8 (the same open areas as Runs 1-7). Vessel and containment failure times for Run 9 were very similar to the ones calculated in Run 7. However, the amount of steel ejected was still similar to the amount calculated in Run 8, much lower than the value of Run 7. The reduced surface areas in the lower plenum were affecting these results. The smaller surface areas provide less surface for heat transfer from the debris or from molten steel coming from the core region. Consequently, less steel in the lower plenum was melted and less molten steel was ejected. In particular, both Runs 8 and 9 predicted that almost no steel is melted and ejected from radial ring number 2. On the contrary, Run 7 predicted that all the steel in the lower plenum for this radial ring 2 is completely molten and ejected. Run 9 was repeated, with the surface areas of the structures in the lower plenum of ring number 2 increased to the values of Run 7. This

309

new run, Run 10, predicted 53% of the steel inventory in the core is ejected because all the steel in ring number 2 of the lower plenum is melted and ejected. This result is comparable to the value calculated in Run 7. Therefore, smaller core-plate open areas result in less debris relocated into the core plate, and longer times to core-plate failure and vessel failure. Smaller lower-plenum structures' surface areas result in less heat transferred to the lower plenum structures and less steel melted and ejected from the vessel. It is believed that Run 9, with reduced surface areas in the lower plenum, has the most realistic input of all MELCOR runs of Table 7. It should also be noted that, although Run 9 ejected little steel into the cavity, it ejected a lot of fuel debris into the cavity (about 98% of the total initial inventory). Larger ejected fuel masses generally result in larger source terms. 6.5. Lower-plenum debris bed models

Run 11 used the MELCOR/BH(Bottom Head) package (Hodge, 1993; Hyman, 1993a,b) and the MELCOR input of Run 9. The results of Run 11 predicted instrument tube failure at 285 min (17 092 s) and lower-head creep-rupture at 467 min (28 040 s). Of the total 279 520 kg (616 342 lb) of corium released into containment in this run, about 100 000 kg (220 000 lb) were released between the times of instrument tube failure and lower-head creep rupture. The remaining mass was released at 467 rain, when the lower head ruptured. Fig. 13 shows the debris mass ejected into the cavity for this run. This is the most credible run of Table 7 because it considers debris quenching in the lower plenum. Finally, Run 12 shows the results obtained using the codes aWRSAR (Ott, 1989; Hodge, 1990b) and CONTAIN (Washington, 1991). These results are taken from Hyman (1990) and Hodge et al. (1990a). These results indicate that about 98% of the total core inventory is released together with a portion of the vessel bottom head. The results from Runs l l and 12 using the lower-plenum debris bed model are in good agreement for vessel and containment failure times and for the amount

310

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

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of steel ejected. Both vessel and containment failure times from these two runs are the longest times of Table 7 with the exception of MELCOR Run 5. The BH and BWRSAR lower-plenum debris bed models result in longer (and probably more realistic) vessel failure times than MELCOR 1.8.1.

6.6. Summary of Table 7 results The MELCOR results presented in Table 7 demonstrate a wide range of behavior. In particular, the amount of steel ejected varies between 5.5% for Run 4 and almost 100% for Runs 6 and 11. The amount of fuel ejected varies between 67% (Run 11) and 99.9% (Run 6). Vessel failure times vary between 153 and 329 min. Containment failure times vary between 361 and 598 min.

These vastly different results were obtained by varying MELCOR input parameters over reasonable values within their ranges of uncertainty.

7. Penetration failure timing

There are several input parameters that control the timing of the failure of the penetrations in the MELCOR code. These parameters are: (a) penetration failure temperature; (b) heat transfer coefficient between the debris and the penetration; and (c) penetration mass. There are uncertainties in the values of these parameters. The effects on reactor vessel failure timing of varying these parameters are presented in Table 8. All these calculations were performed

311

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

Table 8 Vessel failure times calculated by MELCORby varying penetration input parameters in the station blackout sequence Run

1 a

2 3 4 5

Penetration failure temperature

Penetration h

Penetration mass

Vessel failure

Difference

[K

(°F)]

[W m-2 K (Btu ft-2 h -~ °F)]

[kg

(lb)]

[min (s)]

[min (s)]

1273 1700 1273 1273 1273

(1832) (2600) (1832) (1832) (1832)

500 500 50 5 500

2112 2112 2112 2112 12672

(4657) (4657) (4657) (4657) (27942)

163 164 167 169 165

Base 0.5 (26) 4 (242) 6 (369) 2 (103)

(88) (88) (8.8) (0.8) (88)

(9793) (9819) (10035) (10162) (9986)

a This run is Run 3 of Table 7. for the station blackout with ADS depressurization at one-third ACH. The base case in these calculations used a penetration mass of 2112 kg (4657 lb), a penetration heat transfer coefficient of 5 0 0 W m - E K -~ (88 Btu f t - : h -~ ° F - l ) , and a penetration failure temperature of 1273 K (1832°F). This base case is Run 1 of Table 8. In Run 2 of Table 8 the value used for the penetration failure temperature was increased to 1700 K (2600°F), the melting temperature of steel. This is the maximum reasonable value that could be used for this parameter. MELCOR estimated that vessel failure would occur only 26 s later for this case (Run 2 of Table 8) than the base case (Run 1). The base heat-transfer coefficient between debris and penetration used in Run 1 is 500 W m - 2 K (88 Btu f t - : h -~ ° F - I ) . In order to delay vessel failure, MELCOR runs were performed with values 1/10 and 1/100 of this value of the heat transfer coefficient of Run 1. The run with 1/10 the original value (Run 3), predicted vessel failure about 4 min (242 s) later. The run with 1/100 the original value predicted vessel failure 6 min (369 s) later. Both time differences are not very significant. Finally, the mass of the penetrations was increased to six times the original value. In order to avoid duplication of mass, this additional mass of the penetrations was removed from the structural steel mass of axial node 1 in the lower plenum. The results of this Run 5 predicted vessel failure about 2 min (103 s) later than the base case. The timing of vessel failure varied between 163 (Run 1) and 169 min (Run 4). Even if a combina-

tion of these parameters is used (large penetration mass together with high failure temperature and low heat transfer coefficient) the delay in vessel failure timing would probably be only a few minutes. The results of this study (Table 8) indicate that MELCOR'S estimated time of vessel failure is not very sensitive to the input values of the penetration parameters.

8. Timing of vessel failure

The results from Section 7 indicated that vessel failure time predicted by MELCOR for the station blackout is not very sensitive to the input values of the penetration parameters. Once the core plate fails (or melts), the hot debris falls to the bottom of the lower plenum and contacts the penetrations, which fail shortly thereafter. The only way to delay vessel failure in MELCOR version 1.8.1 is by delaying the time at which debris reaches the penetrations. This delay was accomplished in Runs 4 and 5 of Table 7, where debris was supported at each level of the lower plenum until that level structure failed or melted. Run 5 predicted the longest time to vessel failure, 329 min, more than twice the time calculated by Run 6 (153 min). Some results from Run 5 are similar to the results from Run 11 using the O R N L Lower Plenum Debris Bed Package (BH package) (Hodge, 1993; Hyman, 1993a,b) added to MELCOR. The MELCOR 1.8.1/BH package

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J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

starts calculations when the lower plenum is dry (all the water in the lower plenum has been boiled off). Heat transfer between the debris and the water in the lower plenum has been increased in the MELCOR 1.8.1/BH package to represent boiling of water during the simulated debris quenching. Also, the penetrations of the lower head were prevented from failing in MELCOR prior to the activation of the BH package. This was done by artificially increasing the penetration failure temperature to a very high value (i.e. 10 000 K or 17 500°F) and by eliminating heat transfer between the debris and the penetrations. Once the lower plenum is dry, the BH package is activated. Comparison of the vessel failure time of Run 11 with the other times of Table 7 shows the following: (a) the time predicted by the MELCOR 1.8.1/BH package (Run 11) is in good agreement with the time predicted by the codes BWRSAR/CONTAIN (Run 12). (b) the time calculated by the MELCOR 1.8.1/BH package for vessel failure (285 min) is in-between the shortest time calculated by MELCOR (153 min for Run 6) and the longest time (329 min for Run 5) and is closer to the longest time. The most credible run of Table 7 is Run 11. The MELCOR 1.8.1/BH package has an improved lowerplenum molten debris thermal-hydraulic model. The time calculated by the MELCOR 1.8.1/BH package in Run 11 (Table 7) is 285 min (17 092 s). This is the best-estimate time for vessel failure for the low pressure, short-term, station blackout analyzed at the Peach Bottom plant.

9. Lower-plenum and core shroud temperatures The shroud is a cylindrical stainless-steel assembly that separates the downcomer from the core and lower plenum. The lower-plenum shroud is located below the core plate and separates the lower plenum from the downcomer. The core shroud is above the core plate and surrounds the core. The shrouds are modeled in MELCORas heat slabs. Heat slabs cannot melt and relocate in MELCOR. However, the results presented here indicate that melting of the shroud occurred, and a

new flowpath between the core and the downcomer would be established. Each axial level of the core (Section 3) is modeled with a heat slab. Consequently, there are 11 shroud heat slabs, five in the lower plenum, one at the core plate level, and five more in the core region (Fig. 4). Each heat slab has five radial nodes. Node 5 is the outside node, next to the downcomer. Node 1 is the inside node, next to the core or lower-plenum structures. Fig. 14 shows the temperature of the inside radial node (node 1) of the shroud heat slabs in the first six axial levels (includes the five levels of the lower plenum plus the core plate level), for Run 9 of Table 7. The temperatures at all six levels are well below the melting temperature of steel ( 1700 K or 2600°F). Therefore, the shroud at the core plate level and in the lower plenum does not melt. However, creep rupture cannot be ruled out with shroud temperatures in the vicinity of 1300 K (1900°F). The temperature decrease at 3993 s is due to the ADS depressurization. Temperature increases at about 10 000 s and afterwards are due to the fuel debris relocated to the lower plenum. Fig. 15 shows the temperature of the inside radial node (node 1) of the shroud in the five axial levels of the core (levels 7 through 11) for the same Run 9 of Table 7. Fig. 16 shows the temperature of the outside radial node (node 5) of the same five axial levels of Fig. 15. These figures show that the shroud temperature at these five axial levels goes above the steel melting temperature for a significant amount of time (about 1800 s or 30 min). Consequently, according to these results, melting and relocation of these structures would occur. It is anticipated, however, that the molten steel will probably refreeze on colder shroud structures below (lower plenum) and will not be available for ejection into the cavity. Nevertheless, it is important to know that the shroud structure located at the core levels will probably melt during this transient. A new flowpath between the core and the downcomer would be established if the shroud were to melt. The temperature decreases after 12 000 s in Figs. 15 and 16 are because the fuel relocated to

J.J. Carbajo / N u c l e a r

Engineering and Design 152 (1994) 287-317

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the lower plenum, and there is less heating at the core levels of the shroud. The opposite occurs in the lower-plenum shroud (Fig. 14). Shroud melting occurs in Run 9 after vessel failure. However, in the MELCOR 1.8.1/BH analysis (Run 11 of Table 7) melting of the shroud starts at about the same time (10 000 s or 167 min) but vessel failure occurred 7100 s later (at 285 min or 17092 s). This might possibly provide a more direct flow path via the ADS for the fission products to enter the wetwell, because some o f the flow through the debris would probably bypass the steam dryers. Consequently, higher releases of fission products into the containment may result in this scenario owing to the possible bypass of the steam dryers. After vessel failure, this new flowpath is not important.

p.--

10. Conclusions

These sensitivity studies have provided valuable insights into the behavior of the MELCOR code (version 1.8.1) with changes in some input parameters and into the progression of the lowpressure, short-term station blackout severe accident at the Peach Bottom plant. Newer versions of MELCOR have incorporated enhancements that remove some of the limitations found in this study. The results of these sensitivity studies are summarized in this section. A D S actuation time. MELCOR calculations confirmed previous studies that the optimum steam cooling of the core is obtained when the ADS is actuated (or the vessel is depressurized) when the core water level is at one-third of the active core

314

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

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height (ACH), as per EPGs Rev. 3. This result is for high burnup cores. This depressurization timing yields the longest times to core oxidation initiation, core relocation, and vessel and containment failure, because the core is cooled by steam when it is very hot, but before Zircaloy oxidation has started. If the vessel is depressurized too early, steam cooling is wasted since the core is not very hot. If the vessel is depressurized too late, the core has already started to oxidize, and the steam from depressurization will exacerbate Zircaloy oxidation.

Fuel and cladding (intact components)porosities. These parameters have an insignificant effect on the accident progression. Debris porosities. Larger debris porosities resulted in larger volumes of debris occupying the free space of the lower plenum and, consequently,

a..

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in more heat transferred between debris and structures in the lower plenum and more molten steel ejected from the vessel. Steel inventory in the lower plenum. Larger steel inventories in the lower plenum resulted in less molten steel ejected from the vessel because the lower-plenum steel acts as a heat sink for the relocated molten steel from the core.

Lower plenum supporting debris at different axial levels (support flags Of MELCOR). Supporting debris at the different axial levels o f the lower plenum delayed vessel and containment failure times. It also resulted in larger amounts of molten steel ejected from the vessel when the failure temperature of each supporting level was increased to 1700 K (2600°F), the melting temperature of steel. Core-plate open areas. Smaller core-plate open areas resulted in less debris relocated into the core

J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

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TIME ( 1 0 3 s ) Fig. 16. Outside temperatures of the five axial levels of the core shroud for Run 9 of Table 7.

plate and longer times to core plate failure and vessel failure. Surface area of the lower-plenum structures. Smaller lower-plenum surface areas resulted in less heat transferred from debris to these structures and less steel melted and ejected from the vessel. Lower-plenum debris quenching models (SH or B WRSAR models). This improved lower-plenum debris bed model delays vessel failure significantly when compared to MELCOR 1.8.1 results. In the BH model, the debris in the lower plenum is quenched and boils the lower plenum water before the penetrations fail. This is a more realistic model than the MELCOR 1.8.1 model without quenching. The BH Package has been incorporated into version 1.8.2 of MELCOR. Penetration failure parameters. No significant effect. Penetration failure time can be delayed

slightly by increasing penetration failure temperature, by reducing debris/penetration heat transfer coefficients, or by increasing the penetration masses. Lower-plenum and core-plate shroud temperatures. During this transient, the lower-plenum and core-plate shroud temperatures do not reach the melting point but may reach creep-rupture conditions. Core shroud temperatures. MELCOR calculations indicated that the core shroud will melt and relocate during this transient, providing a new flowpath inside the vessel for the fission products that may bypass the steam dryers. Consequently, higher release of fission products into the containment may result. Overall, it appears that the large number of parameters and options in MELCOR that can be

w I---

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J.J. Carbajo / Nuclear Engineering and Design 152 (1994) 287-317

changed without changing the general input of the problem can yield widely different results. Different values of one parameter can be justified (using different criteria) producing different results.

Acknowledgments These sensitivity studies are part of a comprehensive source term study for the Peach Bottom plant (Carbajo, 1993b) performed with the MELCOR code and sponsored by the US Nuclear Regulatory Commission.

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ORNL/TM-11548, Oak Ridge National Laboratory, Oak Ridge, TN, May 1991. S.A. Hodge, C.R. Hyman and L.J. Ott, Primary containment response calculations for unmitigated short-term station blackout at Peach Bottom with consideration of (I) separated metal and oxide pour, (2) melting of reactor vessel bottom head, and (3) degassing of drywell concrete, Letter report, Oak Ridge National Laboratory, Oak Ridge, TN, November 28, 1988. Also published as Appendix A to D.L. Kelly, K.R. Jones, R.J. Dallman and K.C. Wagner, An assessment of BWR Mark-II containment challenges, failure mode, and potential improvements in performance, Report NUREG/CR-5528, EGG-2593, Idaho National Laboratory, July 1990a. S.A. Hodge and L.J. Ott, BWRSAR calculations of reactor vessel debris pours for Peach Bottom short-term station blackout, Nucl. Eng. Des., 121 (1990b) 327-339. S.A. Hodge, J.C. Cleveland, T.S. Kress and M. Petek, Identification and assessment of BWR (in-vessel) severe accident mitigation strategies, Report NUREG/CR-5869, ORNL/ TM-12080, Oak Ridge National Laboratory, Oak Ridge, TN, 1992. S.A. Hodge, C.R. Hyman and R.L. Sanders, BWR Lower Plenum Debris Bed Package Reference Manual, Ver. 1.0.4, Oak Ridge National Laboratory, Oak Ridge, TN, September 1993. C.R. Hyman, CONTAINCalculations of debris conditions adjacent to the BWR Mark-I drywell shell during the later phases of a severe accident, Nucl. Eng. Des., 121 (1990) 379-393. C.R. Hyman and R.L. Sanders, BH Package Programmer's Guide, Ver. 1.0.4, Oak Ridge National Laboratory, Oak Ridge, TN, September 1993a. C.R. Hyman and R.L. Sanders, BWR Lower Plenum Debris Bed Package User's Guide, Version 1.0.4, Oak Ridge National Laboratory, Oak Ridge, TN, September 1993b. T.S. Kress and J.C. Cleveland, Analysis of the thermal response of a BWR Mark-I containment shell to direct contact by molten core materials, Proc. 16th Water Reactor Safety Info. Meeting, Gathersburg, MD, October 24-27, 1988, NUREG/CP-0097, Vol. 5, pp. 429-451. L.J. Ott, Advanced severe accident response models for BWR application, Nucl. Eng. Des., 115 (1989) 289-303. Sandia National Laboratories, MELCOR 1.8.1 Computer Code Manual, Vol. 1: Primer and Package User's Guide; Vol. 2: Reference Manual and Programmer's Guide, prepared for the USNRC by Sandia National Laboratories, June 1991. J.J. Sienicki, M.T. Farmer and B.W. Spencer, Spreading of molten corium in Mark-I geometry following vessel meltthrough, Proc. 4th Nuclear Thermal Hydraulics Conf., Washington, DC, October 30-November 4, 1988. T.P. Speis et al., Estimates of early containment loads from core melt accidents, Report NUREG-1079, July 1985. A.C. Thadani, Research results related to Emergency Procedure Guidelines, Rev. 4 contingency # 3, regarding reactor

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