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SCDAPSIM
Nuclear Engineering and Design 307 (2016) 299–308
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AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM A.K. Trivedi a, C. Allison b, A. Khanna a,⇑, P. Munshi a a b
Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016, India Innovative Systems Software Idaho Falls, ID 83406, USA
h i g h l i g h t s A representative RELAP5/SCDAPSIM model of AP1000 has been developed. Core is modeled using SCDAP. A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). Significant differences in the damage progression have been observed for the two cases. Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario.
a r t i c l e
i n f o
Article history: Received 28 December 2015 Received in revised form 17 July 2016 Accepted 18 July 2016 Available online 5 August 2016 JEL classification: L. Safety and Risk Analysis
a b s t r a c t Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The maximum core surface temperature is 3096 K and 3236 K for the high pressure and low pressure case; leading to maximum hydrogen generation rate of 0.87 and 9.22 kg/s with total hydrogen generated is 747 and 425 kg respectively. Significant differences in the damage progression have been observed for the two cases. Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. Ó 2016 Elsevier B.V. All rights reserved.
1. Introduction Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most
⇑ Corresponding author. Fax: +91 512 2597408. E-mail address: [email protected] (A. Khanna). http://dx.doi.org/10.1016/j.nucengdes.2016.07.019 0029-5493/Ó 2016 Elsevier B.V. All rights reserved.
of the countries. For such improbable accidents, an SBO for the AP1000 (Westinghouse, 2011a) using RELAP5/SCDAPSIM (Allison and Hohorst, 2010) has been simulated. Many improvements have been made in fuel damage progression models in SCDAP portion of the code (Allison et al., 2012) after the Fukushima accident which has now been applied to an advanced reactor design. AP1000 is a 2loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies.
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The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP (Siefken et al., 2001) fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. This model will further be used for core re-flooding study at high pressure (using core makeup tanks) and at low pressure (using In-containment refueling water storage tank) and results will be qualitatively compared with the Quench-11, a PWR fuel (electrically heated) re-flooding experiment performed at KIT, Germany. There is a broad classification of severe accidents depending on pressure. A low pressure event like LOCA or SBO with pressurizer safety relief valve (SRV) stuck open leads to failure of cladding (EPRI, 2013) by ballooning and rupture becomes possible for 1000 K to 1200 K core temperatures. A high pressure event like SBO accompanied by continuous opening and closing of SRV tends to collapse the cladding onto the fuel at lower temperature leading to release of some fission products but there exists a high probability of arresting the progression of accident if cooling water is reintroduced.
Fig. 1 (Schulz, 2006) shows a schematic of the primary heat transport system (PHTS) and the passive safety systems for this design. The AP1000 is a 2-loop pressurized water reactor (PWR) with power rating of 3415 MWt (3400 MWt from core and 15 MWt from the pumps). Its core design is very similar to a 3loop PWR containing 157 fuel assemblies and a 14 foot active core length. Two hundred sixty-four (264) fuel rods in a 17 17 square array constitute a fuel assembly. Remaining 25 positions contain the guide tubes and a central instrument tube. This design has two Delta 125 type steam generators (SG) (Condrac et al., 2004) each rated at 1707.5 MWt. The ‘‘Delta” and ‘‘125” refer to the triangular pitch and the approximate heat transfer surface area of 125,000 square feet based on outside diameter – the actual area is 123,540 square feet (11,477 m2). Each SG has 10,025 tubes made of Inconel Alloy 690 with the steam line having one power operated relief valve (PORV) and six SRV’s. There are four cannedmotor pumps (Baumgarten et al., 2010) integrated into the two SG channel heads. Each pump has (a) a height of 6.73 m and power input of 6.6 MW; (b) the rated speed and flow are 188.5 rad/s and 4.97 m3/s with a pump head of 111 m; (c) minimum required moment of inertia is 695 kg-m2 and a torque of 35,014 N.m. The pressurizer (Westinghouse, 2006) in the AP1000 is larger than in earlier PWRs with vessel (with hemispherical heads) volume of 59.46 m3 (47.6% water) and length of 12.77 m. There are two spray lines with a diameter 0.1016 m connected from the cold legs and a heater with total input power of 1600 kW. A surge line with a diameter 0.4572 m connects the pressurizer bottom to the hot leg in loop 1. There are two SRV’s connected from pressurizer dome while PORV has been removed from this design. 2.1. Passive safety systems (Westinghouse, 2011c)
2. Overview of the design Few advantages of AP1000 design (Westinghouse, 2011b) are summarized here. (1) Cold legs (core inlet) are at a higher elevation than hot leg (core outlet) which allow mid-loop operation during loss of main coolant pump without discharge of the core. Midloop operation is a method to cool the core during refueling outage. It necessitates maintaining the water level in the core up to the hot leg because passive residual heat removal (PRHR) system takes suction from the hot leg. (2) High inertia, high reliability and hermetically sealed canned motor pumps are used in AP1000. They are directly integrated to SG channel head and removing the cross-over leg (U-shaped) between SG cold plenum and the pump. It has reduced the loop pressure drop and also reduced the chances of core uncovery by eliminating the requirement of clearing the loop seal during a small break loss of coolant accident (SBLOCA). (3) All the safety systems are based on natural circulation which means that they can perform their required action without human or power intervention. These systems are reported in detail in a later section on passive (no pumps) safety systems. (4) Automatic depressurization system (ADS) valves are provided to reduce the system pressure during any abnormal event like SBLOCA. This depressurization has a critical role in actuating long term cooling system which is a low pressure tank. (5) In-vessel retention capability of the debris by cooling of the external surface of the reactor vessel in case of any improbable core melting.
Passive safety systems in AP1000 have been summarized here although they are intentionally considered unavailable to analyze an SBO at high pressure and depressurized (low pressure) scenario with no water addition. 2.1.1. Core makeup tanks There are two CMTs in this design in place of high pressure injection system (HPIS) in older PWRs. Each CMT is a vertical cylindrical tank (volume 70.79 m3) with hemispherical heads located inside the containment. The CMTs are designed to inject coolant (@ 322 K) into the reactor vessel at a high pressure during any abnormal event. The bottom of CMT is connected to the direct vessel injection (DVI) line while the top is connected to cold leg through the pressure balance line (PBL) to maintain it at RCS pressure. 2.1.2. Accumulators There are two spherical accumulators (volume of each equal to 56.63 m3 @ 322 K) for injection at medium pressure, 4.83 MPa, through the DVI line. The coolant in the accumulator is pressurized with nitrogen. Volume of nitrogen is 8.5 m3 and the water volume is 48.14 m3 (Tower et al., 1998). These accumulators (isolated from RCS during normal operation) are designed to provide a high flow of water into the vessel for quick cooling of the core during an LBLOCA. The tank water level, gas pressure, and discharge line resistance have been kept similar to the AP600 (Schulz et al., 2001) 2.1.3. In-containment refueling water storage tank (IRWST) IRWST is a large tank with a volume 2510 m3 located underneath the operating deck inside the containment. The bottom of IRWST is above the RCS and it can drain the coolant due to gravity into the RCS. It is designed to inject coolant near atmospheric
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Fig. 1. PHTS and passive safety systems of AP1000 (Schulz, 2006).
pressure for long term cooling capability using two separate lines connecting IRWST to DVI through check valves. This tank also acts as a heat sink for the passive residual heat removal (PRHR) system. 2.1.4. Passive residual heat removal (PRHR) system This heat exchanger consists of C-shaped tubes inside the IRWST tank. These tubes take suction from the hot leg (pressurizer side) and discharge into cold plenum of SG. It is designed to remove decay heat from the core by setting up a natural circulation flow path between hot and cold legs and transferring heat to IRWST. 2.1.5. Automatic depressurization system (ADS) The ADS goes through four stages to depressurize the RCS pressure to a low enough pressure to enable the IRWST to inject coolant at high flow rate. Three valves are connected from top of the pressurizer and discharge into IRWST while the fourth (connected from hot leg) discharges into containment. The first stage actuates when the CMT water level sinks to 67.5% (Westinghouse, 2011d). 3. RELAP5/SCDAPSIM/MOD3.5 This code (Allison and Hohorst, 2010) is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behavior under normal operating conditions or under design basis or severe accident conditions. The overall RCS thermal hydraulic response, control system behavior, reactor kinetics and behavior of special reactor system components such as valves and pumps are being calculated by RELAP5 models. The behavior of core and vessel structures under normal and accident conditions are being calculated by SCDAP. This portion of the code
includes user-selectable reactor component models for LWR fuel rods, Ag–In–Cd and B4C control rods, BWR control blade/channel boxes, electrically heated fuel rod simulators, general core and vessel structures. It also includes models to treat the later stages of a severe accident including debris and molten pool formation, debris/vessel interactions and the structural failure (creep rupture) of vessel structures. The later models are automatically invoked by the code as the damage in the core and vessel progresses. This is the first release of the code with QUENCH and PARAMETER (Allison and Hohorst, 2010) driven SCDAP modeling improvements (Allison et al., 2012): (a) modeling of gap conductance, (b) radiation heat transfer across radial gap in shroud (c) oxidation of zircaloy in presence of air. They are applicable to both fuel bundle experiments and nuclear power plant analysis. 4. Plant model and null transient The details of the thermal hydraulic nodalization (Fig. 2) has already been reported in Trivedi et. al (2015) which involves a large break LOCA in AP1000 using RELAP5/SCDAPSIM and its verification against publically available TRACE and WCOBRA-TRAC results. Core is now modeled with SCDAP fuel and control components. It is divided into 6 radial rings and 20 axial nodes. The equivalent flow area in the six group of assemblies is modeled as RELAP5 pipe components 108, 110, 112, 114, 116 and 118. Each volume of these channels is connected to the respective outer channel using cross flow junctions as shown in Fig. 2. Channel 108 is modeled with maximum radial power peaking representing the single assembly at the center. Channels 110, 112, 114, 116, 118 represent a group of 24, 30, 24, 42, 36 assemblies in radially outward direction. Each group contains a fuel component and a control compo-
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Fig. 2. RELAP5/SCDAPSIM nodalization of PHTS of AP1000.
Table 1 Steady state as obtained from null transient. Parameter
RELAP5 value
Unit design
Error, %
Core thermal power (MWt) Core inlet flow rate (kg/s) Core outlet flow rate (kg/s) (loop1/2) Pump velocity (rad/s) Hot leg pressure (MPa) Cold leg pressure (MPa) Core inlet temperature (K) Core outlet temperature (K) SG steam outlet pressure (MPa) SG secondary inlet temperature (K) SG secondary outlet temperature (K) SG secondary side water level (m) SG secondary steam flow rate (kg/s) Pressurizer steam/water volume (m3)
3468 (102%) 15,177 7588.8/7588.8 188.7 15.62 15.93 553.6 594.8 5.6617 499.7 544.7 13.46 957/957 31.24/28.22
3415 15,170 7585 188.5 15.49 15.92 553.8 594.2 5.6609 499.7 545.9 13.46 943 31.15/28.32
0.04 0.05 0.1 0.8 0.06 0.03 0.001 0.014 0.0 0.2 0.0 1.4 0.28/0.35
nent with corresponding number of fuel and control rods in each group. Hence the core consist of the total 13 SCDAP components: 1, 3, 5, 7, 9, 11 represents the fuel rod and 2, 4, 6, 8, 10, 12 represents the control components while 13th component is a shroud component corresponding to core shroud surrounding the core. Radiation enclosure is modeled corresponding to each channel to simulate bundle to bundle radiation heat transfer for which the view factors are set to be calculated by the code. A matrix of 17 17 fuel bundle in the core is the input with elements similar to an assembly. Decay heat has been calculated based on 1979 ANS standard data, ANS79-3 which specifies the three isotopes, 235U, 238 U, and 239Np.
The steady state conditions after running a null transient for 2000 s are reported in Table 1. They closely match with the design data (Westinghouse, 2011d) and the relative error is less than the acceptable limits (OECD, 2004). 5. Results and discussion Transients have been run for 5 h (18,000 s) for both the low pressure (due to pressurizer safety valve stuck open) and high pressure (accompanied by continuous opening and closing of SRV) case. SBO starts with a total loss of feed water due to loss of offsite power which then leads to closer of main steamline iso-
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5.1. Decay and oxidation heat
lation valve (MSIV) and tripping of reactor coolant pumps. Reactor scrams due to high pressure (16.96 MPa) on primary side. Decay heat continues to provide heat to the RCS. All the safety systems including automatic depressurization system (ADS) are assumed unavailable. Steam line safety valves do not get actuated during these transients. This section compares the fuel damage progression for these two cases. High pressure and low pressure cases are referred as the cases of no depressurization and depressurization. These calculations have been limited to in-vessel phase of accident progression which is heating and melting of the fuel. It has been reported by Sehgal (2012) that the time span available for severe accident mitigating actions like core re-flooding is of the order of 2 h while it is 4–5 h to restrict the core melt progression.
Transient starts at 2010 s and reactor trips with in 5 s of the transient which results into decay heat of about 200 MWt (6% of power). It decreases sharply during initial 1800 s (30 min) and reaches about 65 MWt (1% of power) at this time. The behavior of decay and oxidation heat has been depicted in Fig. 3. RELAP5 calculated value of decay heat has also been compared with that reported by Westinghouse for SBO (Westinghouse, 2011e). They are in good agreement except for the first 1800 s: the fall of decay heat is steeper than the reported. Here, it is important to note that oxidation heat is quite large in magnitude in comparison to decay heat for some portion of the transient. Maximum oxidation heat is 1357 and 128 MWt in depressurized and high pressure progression respectively.
Decay heat (MWt)
Oxidation heat (MWt)
1360 1352 300 250 200 150 100 50 0 225
No depressurization Depressurization Reported (Westinghouse)
200 175 150 125 100 75 50 0
2000
4000
6000
8000
10000
15000 18000
Time (s)
Hydrogen generation (kg/s) Total hydrogen (kg)
Fig. 3. Decay and oxidation heat.
800
747 kg
600
425 kg 400 200 0 10 8
9.22 kg/s @ 560 s No depressurization Depressurization
2.5 2.0 1.5 1.0
0.87 kg/s @ 7030 s
0.5 0.0 0
2000
4000
6000
8000
10000
Time (s) Fig. 4. Total hydrogen and rate of generation.
15000 18000
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5.2. Hydrogen generation
5.3. Temperatures
Fig. 4 shows the hydrogen generation rate and total hydrogen produced during the transient. The maximum rate for high pressure and depressurized case is 0.87 (@ 7030 s) and 9.22 (@ 560 s) kg/s while total hydrogen is 747 and 425 kg respectively. Time at which oxidation of steam and zircaloy starts is quite different for the two cases which is mainly due to the RCS pressure. For the high pressure case, progression is delayed as water does not boil instantaneously while in the depressurized case, steam is quickly available for oxidation. Total hydrogen for the high pressure case is more than 75% of the depressurized case (although the rate is quite high in this case). Figure also shows that almost no hydrogen getting produced after 8000 s for the depressurized case and after 11,000 s for the high pressure case.
The maximum core surface temperature in SCDAP is a variable (bgmct) which represents maximum cladding temperature up to the melting of cladding (2200 K) and a hot spot in the core afterwards. It has been reported in Fig. 5: 3096 K @ 15,975 s for the high pressure and 3236 K @ 7380 s for the low pressure case. For the high pressure case, temperature starts increasing (due to decay heat) reaching 1033 K @ 4310 s when melting of Ag–In–Cd alloy (Hofmann, 1999) starts; a temporary decrease in temperature is noticed around this time. As temperature reaches 1500 K @ 6265 s, rapid Zircaloy steam reaction starts leading to uncontrolled temperature escalation and hydrogen generation rate peaking to 0.87 kg/s @ 7030 s. The liquefied control materials (including Inconel spacer grids) relocate (Allison and Hohorst, 2008) into
18
3236 K @ 7380 s
3126 K
3250
16
Pressurizer pressure (MPa)
2790 K @ 7140 s
3000 2750
2831 K @ 855 s
2500
3096 K @ 15975 s
2250
No depressurization Depressurization
2000 1750 1500
1500 K @ 6265 s
1033 K @ 4310 s
1250 1000 750
No depressurization Depressurization
16.5 MPa@1740 s
14
18
12
16
12.58 MPa @105 s
10
14 12
Expanded view upto 2000 s
10
8
8 6
6
4 2
4 0
200
400
600
800
1000 1200 1400 1600 1800 2000
2
0.56 MPa@ 105 s
500 0
2000
4000
6000
8000 10000 12000 14000 16000 18000
0
2000
4000
6000
8000 10000 12000 14000 16000 18000
Time (s)
Time (s) Fig. 5. Core maximum surface temperature.
Fig. 7. Pressurizer pressure.
1800
1255 K@ 10325 s
1600
Vessel lower plenum
1400 1200
No depressurization Depressurization
1000
Wall temperatures (K)
Core maximum surface temperature (K)
3500
800 600 400 2400
1733 K @ 10325 s Pressurizer surge line
2000 1600 1200 800 400 0
2000
4000
6000
8000
10000
12000
14000
Time (s) Fig. 6. Lower plenum and surge line wall temperatures.
16000
18000
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8.4 8.0
Secondary side pressure (MPa)
7.6
SG-1 Pressurizer side
7.2 6.8
No depressurization Depressurization
6.4 6.0 5.6 5.2 8.4 8.0 7.6 7.2 6.8
SG-2
6.4 6.0 5.6 5.2 0
2000
4000
6000
8000
10000
12000
14000
16000
18000
Time (s)
Pressurizer water level (m) Core water level (m)
Fig. 8. SG secondary side pressure.
6
6.24 m @ 2575 s
5
TAF @ 6.2484 m BAF @ 1.9812 m
4 3
2.1 m @ 60 s
2
1.99 m @ 4160 s
1 0
12
12.67 m @ 2510 s 11.82 @ 25 s
10
No depressurization Depressurization
8 6
6.18 m @ 1280 s 1 m @ 85 s 0.08 m @ 1270 s
4 2
0.08 m @ 4740 s
0 0
2000
4000
6000
Time (s) Fig. 9. Core and pressurizer water level.
16000 18000
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1.51 m
1.4
0.6
No depressurization Depressurization
1.2
0.4 0.2
0.0=No, 1.0=Yes
1.0
0.0
0.8
-0.2
0.6
-0.4
0.4 -0.6 0.2
0.13 m@ 7775 s
0.13 m@ 855 s
-0.8
Crust supporting molten pool fails
0.8
1.85 m@ 6105 s
1.6
Molten pool radius (m)
lower (colder) part of the assembly. This material freezes and forms limited blockages in the lower portion of each assembly: leading to diversion of the flow from hotter channels into colder channels. This diversion increase due to remaining Zircaloy melts (dip in temperature @ 7140 s) relocating into lower portion of assembly. When, temperature reaches near 2800 K, fuel and oxidized cladding, (U–Zr)–O2 melts and start to form ceramic molten pool (Allison and Hohorst, 2008) at 7775 s. The other case of depressurization leads to flashing of water due to sharp decrease in pressure. Oxidation quickly comes into picture and temperature starts rising steeply reaching to 2831 K @ 855 s with a maximum (3236 K @ 7380 s) and settling to 3126 K at the end of transient. Lower plenum and surge line outer wall temperatures are presented in Fig. 6. For the high pressure case, the surge line temperature reaches 1733 K @ 10,325 s while lower plenum wall temperature is much lower, 1255 K, reinforcing the fact that surge line fails before vessel failure as also noticed by Park et al. (2006) for APR1400. These temperatures are quite lower in the depressurized case hence this kind of scenarios does not pose a threat to the integrity of the vessel (for this 5 h transient) even
1.0
1.8
0.0 0
2000
4000
6000
-1.0 8000 10000 12000 14000 16000 18000
Time (s) Fig. 10. Molten pool radius.
Table 2 Core degradation map after 5 h (18,000 s) of transient.
P = Porous debris
I = Intact fuel
V = RELAP fluid volume now void of fuel
xxMxx = Molten or frozen ceramic pool & blockage occurs in the volume Fuel rod components 1
3
Fuel rod components
5
7
9
11
Axial node
1
No depressurization
3
5
7
9
11
Depressurization
I
I
I
I
I
I
20
V
I
I
I
I
I
I
I
I
I
I
I
19
V
V
V
V
V
I
V
V
I
I
I
I
18
V
V
V
V
V
I
V
V
V
I
I
I
17
V
V
V
V
V
I
V
V
V
I
I
I
16
V
V
V
V
V
I
V
V
V
V
I
I
15
V
V
V
V
V
I
14
V
V
V
V
V
V
I
13
V
V
V
V
V
V
xxMxx
xxMxx xxMxx
xxMxx
xxMxx
xxMxx
I
V
V
V
I
V
V
V
V
I
V
I
I
12
V
V
V
V
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
11
V
V
V
V
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
10
V
V
V
V
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
9
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
8
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
7
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx V
I I
I
6
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
P
I
5
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
P
P
P
P
I
I
4
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
I
I
I
I
I
I
3
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
I
I
I
I
I
I
2
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
I
I
I
I
I
I
1
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
Table 3 Summary of the results after 5 h (18,000 s) of transient. Case
Scenario
Maximum core surface temperature (K)
Maximum hydrogen generation rate (kg/s)
Total hydrogen (kg)
Molten pool radius (m)
1 2
No depressurization Depressurization
3096 @ 15,975 s 3236 @ 7380 s
0.87 @ 7030 s 9.22 @ 560 s
747 425
1.51 1.85
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though it leads to much earlier severe core damage. The initial surge line wall temperature increase for the depressurized case is probably due to the choking of the flow in small diameter pipe (surge line). This two phase conditions has been created by sudden depressurization.
is intact for the high pressure case. In the depressurized case, almost all the core node except outer channel has been melted. The outer channel assemblies have survived because of the shroud surrounding the core which is modeled to radiatively receive the heat from this channel.
5.4. Primary and secondary side pressure
6. Conclusions
Fig. 7 represents the primary side pressurizer pressure. For the high pressure case, pressurizer SRV has been modeled as motor valve to simulate cyclic behavior (Vierow et al., 2004) caused by opening and closing of SRV. It opens when the RCS pressure increases to 17.75 MPa closes when the pressure decreases to 17.13 MPa with a valve change rate of 5 s 1 (0.2 s to change the valve position) which is the rate of change of the normalized valve area as the valve opens or closes. RCS pressure initial decrease (12.58 MPa @ 105 s) is primarily due to reactor SCRAM which in turn takes about 2000 s to boil the RCS water inventory. This SRV has been simply modeled as trip valve (to simulate stuck open behavior) which fully opens (instantaneously) when pressure reaches 17.75 MPa for the case of depressurization (low pressure). This SRV stuck open scenario has been taken from the TMI-2 accident scenario. RCS very quickly depressurizes (0.56 MPa @ 105 s) due to high pressure coolant expulsion from the SRV. Fig. 8 represents the pressure on the secondary side: for high pressure case, SG PORV gets actuated @ 7.92 MPa instantaneously and both the SGs show an expanded cycling time. For the low pressure case, pressure profiles for two SG’s are different because of the large coolant discharge (from pressurizer safety valve) on SG-1 side which provides additional cooling and results in little more depressurization on the secondary side.
SBO in AP1000 with and without depressurization has been studied and the summary of the transient results are reported in Table 3. Decay heat based on RELAP5 calculation has also been compared with that of Westinghouse for SBO. They are in good agreement except during the first 1800 s: the RELAP5 decay heat is steeper than the reported. The calculated value of maximum core surface temperature is 3096 K and 3236 K for the high and low pressure cases respectively. These two cases give maximum hydrogen generation rate of 0.87 and 9.22 kg/s with total H2 generated being 747 and 425 kg. Significant differences in the damage progression have been observed for the two cases. The depressurized SBO looks quite superior as less hydrogen is produced and almost no threat to integrity of the vessel is posed. The drawbacks are the early start of fuel damage and almost all the core except outer channel has melted after this 5 h transient. The high pressure SBO has the advantage of delaying the core damage progression which can be arrested using severe accident management guidelines like core flooding otherwise this poses a threat to vessel integrity; this simulation also reinforces the fact that surge line failure occurs before the vessel. The progression is observed to be very sensitive to steady state pressurizer water volume specially with respect to hydrogen hence it should be modeled as accurately as possible.
5.5. Core and pressurizer water level Fig. 9 shows the water level in the core and pressurizer. Water level is measured from bottom of the lower plenum to top of the active fuel (TAF @ 6.2484 m) with bottom of the active fuel (BAF) at 1.9812 m while that of pressurizer is from bottom to top of cylinder. Pressurizer water level decrease is due to initial pressure dip for high pressure case and later @ 1800 s, the vapor is vented through SRV and simultaneously liquid is forced (Vierow et al., 2004) to occupy whole pressurizer volume. As the core starts to uncover, large amount of steam is generated accompanied by decrease in water level. The core water follows the pressurizer water level sinking after 3000 s reaching the BAF at 4160 s (1.99 m) and lower plenum completely vaporizes at 6000 s. In the depressurized case, pressurizer water level continually decreases and bottom is completely voided in 1270 s; core water level quickly decrease to BAF (2 m @ 60 s) due to early core uncovery and leveling off in the lower plenum @ 0.2 m at the end of transient. 5.6. Molten pool radius and core degradation map The molten pool radius has been presented in Fig. 10 for both the transients. For depressurized case molten pool starts getting formed very early in the transient with a value of molten pool radius of 0.13 m @ 855 s which achieves a value of 1.85 m @ 6105 s, confirming the early uncovery of the core. For high pressure case, it reaches to 1.51 m at the end of the transient and showing that core uncovery is delayed; molten pool (radius = 0.13 m) started to form at 7775 s. Figure on the right axis shows the parameter for behavior of crust supporting the molten pool. Its value is 0 for both the cases which means that crust does not fail. Core degradation map after 18,000 s is depicted in Table 2 which clearly shows that one third of the core including outer channel
7. Disclaimer This article presents the work performed at Indian Institute of Technology Kanpur, India as an academic exercise based on the publically available data; hence Westinghouse Electric Company and the US NRC are not responsible for the details reported here. References Allison, Chris M., Hohorst, Judith K., 2008. An assessment of effectiveness of core exit temperatures with respect to PWR core damage state using RELAP/ SCDAPSIM/MOD3.4. Nucl. Eng. Des. 238 (2008), 1547–1560. Allison, C.M., Hohorst, J.K., 2010. Role of RELAP/SCDAPSIM in nuclear safety. Sci. Technol. Nucl. Install. 2010 (2010), 1–17. Allison, Chris, Hohorst, Judith K., Allison, Brian, Konjarek, Damir, Bajs, Tomislav, Pericas, R., Reventos, Francesc, Lopez, Ramon, 2012. Preliminary assessment of the possible BWR core/vessel damage states for Fukushima Daiichi station blackout scenarios using RELAP/SCDAPSIM. Sci. Technol. Nucl. Install. 2012, 1– 25. Baumgarten, Sven, Brecht, Bernhard, Bruhns, Uwe, Fehring, Pete, 2010. Reactor coolant pump type RUV for Westinghouse Reactor AP1000. Proceedings of ICAPP ’10, San Diego, CA, USA. Condrac, Robert F., Waszink, Remco P., Pankiewics-Nohr, Christopher G., 2004. The delta 125 steam generator design for the AP1000. In: Proceedings of ICAPP ’04, Pittsburgh, PA USA. EPRI, 2013. Fukushima technical evaluation: phase 1-MAAP5 analysis. Palo Alto, CA. Hofmann, P., 1999. Current knowledge on core degradation phenomena, a review. J. Nucl. Mater. 270 (1999), 194–211. OECD, 2004. Neutronic/thermal-hydraulic coupling in LWR technology: state-ofthe-art report. Nuclear Energy Agency, NEA NO. 5436. Park, R.J., Kim, S.B., Suh, K.Y., Rempe, J.L., Cheung, F.B., 2006. Detailed analysis of late-phase core-melt progression for the evaluation of in-vessel corium retention. Nucl. Technol. 156 (2006), 270–281. Schulz, T.L., 2006. Westinghouse AP1000 advanced passive plant. Nucl. Eng. Des. 236 (2006), 1547–1557. Schulz, T.L., Kemper, R.M., Gagnon, A.F., 2001. AP1000 passive safety system design and analysis. In: 9th International Conference on Nuclear Engineering, Nice, France. Sehgal, Bal Raj, 2012. Nuclear Safety in Light Water Reactors - Severe Accident Phenomenology. SARNET (Severe Accident Network). Academic Press, Elsevier (2012).
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Siefken, L.J., Coryell, E.W., Harvego, E.A., Hohorst, J.K., 2001. SCDAP/RELAP5/MOD3.3 Code Manual, Modeling of Reactor Core and Vessel Behavior During Severe Accidents, NUREG/CR-6150, vol. 2, Rev. 2, INEL-96/0422, Idaho National Engineering and Environmental Laboratory. Tower, S.N., Schulz, T.L., Vijuk, R.P., 1998. Passive and simplified system features for the advanced Westinghouse 600 MWe PWR. Nucl. Eng. Des. 109 (1998), 147– 154. Trivedi, A.K., Allison, C., Khanna, A., Munshi, P., 2015. RELAP5/SCDAPSIM model development for AP1000 and verification for large break LOCA. Nucl. Eng. Des. 305 (2016), 222–229. Vierow, K., Liao, Y., Johnson, J., Kenton, M., Gauntt, R., 2004. Severe accident analysis of a PWR station blackout with MELCOR, MAAP4 and SCDAP/RELAP5 codes. Nucl. Eng. Des. 234 (2004), 129–145.
Westinghouse, 2006. AP1000 Pressurizer Design. APP-GW-GLR-0 16, Rev. 0 (ML061600440). Westinghouse, 2011a. Westinghouse AP1000 design control document Rev. 19. Chapter 4- Nuclear Design (ML11171A445). Westinghouse, 2011b. Westinghouse AP1000 design control document Rev. 19. Chapter 5- Reactor Coolant System and Connected Systems (ML11171A454). Westinghouse, 2011c. Westinghouse AP1000 design control document Rev. 19. Chapter 6- Passive Core Cooling Systems (ML11171A459). Westinghouse, 2011d. Westinghouse AP1000 design control document Rev. 19. Chapter 15- Accident Analysis (ML11171A373). Westinghouse, 2011e. Westinghouse AP1000 Nuclear Power Plant-coping with station blackout. NPP_NPP_000065 1-9.
Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
AP1000 station blackout study with and without depressurization using RELAP5/SCDAPSIM A.K. Trivedi a, C. Allison b, A. Khanna a,⇑, P. Munshi a a b
Nuclear Engineering and Technology Program, Indian Institute of Technology, Kanpur 208016, India Innovative Systems Software Idaho Falls, ID 83406, USA
h i g h l i g h t s A representative RELAP5/SCDAPSIM model of AP1000 has been developed. Core is modeled using SCDAP. A SBO for the AP1000 has been simulated for high pressure (no depressurization) and low pressure (depressurization). Significant differences in the damage progression have been observed for the two cases. Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario.
a r t i c l e
i n f o
Article history: Received 28 December 2015 Received in revised form 17 July 2016 Accepted 18 July 2016 Available online 5 August 2016 JEL classification: L. Safety and Risk Analysis
a b s t r a c t Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most of the countries. For such improbable accidents, a SBO for the AP1000 using RELAP5/SCDAPSIM has been simulated. Many improvements have been made in fuel damage progression models of SCDAP after the Fukushima accident which are now being tested for the new reactor designs. AP1000 is a 2-loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies. The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. The maximum core surface temperature is 3096 K and 3236 K for the high pressure and low pressure case; leading to maximum hydrogen generation rate of 0.87 and 9.22 kg/s with total hydrogen generated is 747 and 425 kg respectively. Significant differences in the damage progression have been observed for the two cases. Results also reinforced the fact that surge line fails before vessel failure in case of high pressure scenario. Ó 2016 Elsevier B.V. All rights reserved.
1. Introduction Severe accidents like TMI-2, Chernobyl, Fukushima made it inevitable to analyze station blackout (SBO) for all the old as well as new designs although it is not a regulatory requirement in most
⇑ Corresponding author. Fax: +91 512 2597408. E-mail address: [email protected] (A. Khanna). http://dx.doi.org/10.1016/j.nucengdes.2016.07.019 0029-5493/Ó 2016 Elsevier B.V. All rights reserved.
of the countries. For such improbable accidents, an SBO for the AP1000 (Westinghouse, 2011a) using RELAP5/SCDAPSIM (Allison and Hohorst, 2010) has been simulated. Many improvements have been made in fuel damage progression models in SCDAP portion of the code (Allison et al., 2012) after the Fukushima accident which has now been applied to an advanced reactor design. AP1000 is a 2loop pressurized water reactor (PWR) with all the emergency core cooling systems based on natural circulation. Its core design is very similar to 3-loop PWR with 157 fuel assemblies.
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The primary circuit pumps, pressurizer and steam generators (with necessary secondary side) are modeled using RELAP5. The core has been divided into 20 axial nodes and 6 radial rings; the corresponding six groups of assemblies have been modeled as six pipe components with proportionate flow area. Fuel assemblies are modeled using SCDAP (Siefken et al., 2001) fuel and control components. SCDAP has 2d-heat conduction and radiative heat transfer, oxidation and complete severe fuel damage progression models. The final input deck achieved all the steady state thermal hydraulic conditions comparable to the design control document of AP1000. To quantify the core behavior, under unavailability of all safety systems, various time profiles for SBO simulations @ high pressure and low pressure have been compared. This analysis has been performed for 102% (3468 MWt) of the rated core power. This model will further be used for core re-flooding study at high pressure (using core makeup tanks) and at low pressure (using In-containment refueling water storage tank) and results will be qualitatively compared with the Quench-11, a PWR fuel (electrically heated) re-flooding experiment performed at KIT, Germany. There is a broad classification of severe accidents depending on pressure. A low pressure event like LOCA or SBO with pressurizer safety relief valve (SRV) stuck open leads to failure of cladding (EPRI, 2013) by ballooning and rupture becomes possible for 1000 K to 1200 K core temperatures. A high pressure event like SBO accompanied by continuous opening and closing of SRV tends to collapse the cladding onto the fuel at lower temperature leading to release of some fission products but there exists a high probability of arresting the progression of accident if cooling water is reintroduced.
Fig. 1 (Schulz, 2006) shows a schematic of the primary heat transport system (PHTS) and the passive safety systems for this design. The AP1000 is a 2-loop pressurized water reactor (PWR) with power rating of 3415 MWt (3400 MWt from core and 15 MWt from the pumps). Its core design is very similar to a 3loop PWR containing 157 fuel assemblies and a 14 foot active core length. Two hundred sixty-four (264) fuel rods in a 17 17 square array constitute a fuel assembly. Remaining 25 positions contain the guide tubes and a central instrument tube. This design has two Delta 125 type steam generators (SG) (Condrac et al., 2004) each rated at 1707.5 MWt. The ‘‘Delta” and ‘‘125” refer to the triangular pitch and the approximate heat transfer surface area of 125,000 square feet based on outside diameter – the actual area is 123,540 square feet (11,477 m2). Each SG has 10,025 tubes made of Inconel Alloy 690 with the steam line having one power operated relief valve (PORV) and six SRV’s. There are four cannedmotor pumps (Baumgarten et al., 2010) integrated into the two SG channel heads. Each pump has (a) a height of 6.73 m and power input of 6.6 MW; (b) the rated speed and flow are 188.5 rad/s and 4.97 m3/s with a pump head of 111 m; (c) minimum required moment of inertia is 695 kg-m2 and a torque of 35,014 N.m. The pressurizer (Westinghouse, 2006) in the AP1000 is larger than in earlier PWRs with vessel (with hemispherical heads) volume of 59.46 m3 (47.6% water) and length of 12.77 m. There are two spray lines with a diameter 0.1016 m connected from the cold legs and a heater with total input power of 1600 kW. A surge line with a diameter 0.4572 m connects the pressurizer bottom to the hot leg in loop 1. There are two SRV’s connected from pressurizer dome while PORV has been removed from this design. 2.1. Passive safety systems (Westinghouse, 2011c)
2. Overview of the design Few advantages of AP1000 design (Westinghouse, 2011b) are summarized here. (1) Cold legs (core inlet) are at a higher elevation than hot leg (core outlet) which allow mid-loop operation during loss of main coolant pump without discharge of the core. Midloop operation is a method to cool the core during refueling outage. It necessitates maintaining the water level in the core up to the hot leg because passive residual heat removal (PRHR) system takes suction from the hot leg. (2) High inertia, high reliability and hermetically sealed canned motor pumps are used in AP1000. They are directly integrated to SG channel head and removing the cross-over leg (U-shaped) between SG cold plenum and the pump. It has reduced the loop pressure drop and also reduced the chances of core uncovery by eliminating the requirement of clearing the loop seal during a small break loss of coolant accident (SBLOCA). (3) All the safety systems are based on natural circulation which means that they can perform their required action without human or power intervention. These systems are reported in detail in a later section on passive (no pumps) safety systems. (4) Automatic depressurization system (ADS) valves are provided to reduce the system pressure during any abnormal event like SBLOCA. This depressurization has a critical role in actuating long term cooling system which is a low pressure tank. (5) In-vessel retention capability of the debris by cooling of the external surface of the reactor vessel in case of any improbable core melting.
Passive safety systems in AP1000 have been summarized here although they are intentionally considered unavailable to analyze an SBO at high pressure and depressurized (low pressure) scenario with no water addition. 2.1.1. Core makeup tanks There are two CMTs in this design in place of high pressure injection system (HPIS) in older PWRs. Each CMT is a vertical cylindrical tank (volume 70.79 m3) with hemispherical heads located inside the containment. The CMTs are designed to inject coolant (@ 322 K) into the reactor vessel at a high pressure during any abnormal event. The bottom of CMT is connected to the direct vessel injection (DVI) line while the top is connected to cold leg through the pressure balance line (PBL) to maintain it at RCS pressure. 2.1.2. Accumulators There are two spherical accumulators (volume of each equal to 56.63 m3 @ 322 K) for injection at medium pressure, 4.83 MPa, through the DVI line. The coolant in the accumulator is pressurized with nitrogen. Volume of nitrogen is 8.5 m3 and the water volume is 48.14 m3 (Tower et al., 1998). These accumulators (isolated from RCS during normal operation) are designed to provide a high flow of water into the vessel for quick cooling of the core during an LBLOCA. The tank water level, gas pressure, and discharge line resistance have been kept similar to the AP600 (Schulz et al., 2001) 2.1.3. In-containment refueling water storage tank (IRWST) IRWST is a large tank with a volume 2510 m3 located underneath the operating deck inside the containment. The bottom of IRWST is above the RCS and it can drain the coolant due to gravity into the RCS. It is designed to inject coolant near atmospheric
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Fig. 1. PHTS and passive safety systems of AP1000 (Schulz, 2006).
pressure for long term cooling capability using two separate lines connecting IRWST to DVI through check valves. This tank also acts as a heat sink for the passive residual heat removal (PRHR) system. 2.1.4. Passive residual heat removal (PRHR) system This heat exchanger consists of C-shaped tubes inside the IRWST tank. These tubes take suction from the hot leg (pressurizer side) and discharge into cold plenum of SG. It is designed to remove decay heat from the core by setting up a natural circulation flow path between hot and cold legs and transferring heat to IRWST. 2.1.5. Automatic depressurization system (ADS) The ADS goes through four stages to depressurize the RCS pressure to a low enough pressure to enable the IRWST to inject coolant at high flow rate. Three valves are connected from top of the pressurizer and discharge into IRWST while the fourth (connected from hot leg) discharges into containment. The first stage actuates when the CMT water level sinks to 67.5% (Westinghouse, 2011d). 3. RELAP5/SCDAPSIM/MOD3.5 This code (Allison and Hohorst, 2010) is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behavior under normal operating conditions or under design basis or severe accident conditions. The overall RCS thermal hydraulic response, control system behavior, reactor kinetics and behavior of special reactor system components such as valves and pumps are being calculated by RELAP5 models. The behavior of core and vessel structures under normal and accident conditions are being calculated by SCDAP. This portion of the code
includes user-selectable reactor component models for LWR fuel rods, Ag–In–Cd and B4C control rods, BWR control blade/channel boxes, electrically heated fuel rod simulators, general core and vessel structures. It also includes models to treat the later stages of a severe accident including debris and molten pool formation, debris/vessel interactions and the structural failure (creep rupture) of vessel structures. The later models are automatically invoked by the code as the damage in the core and vessel progresses. This is the first release of the code with QUENCH and PARAMETER (Allison and Hohorst, 2010) driven SCDAP modeling improvements (Allison et al., 2012): (a) modeling of gap conductance, (b) radiation heat transfer across radial gap in shroud (c) oxidation of zircaloy in presence of air. They are applicable to both fuel bundle experiments and nuclear power plant analysis. 4. Plant model and null transient The details of the thermal hydraulic nodalization (Fig. 2) has already been reported in Trivedi et. al (2015) which involves a large break LOCA in AP1000 using RELAP5/SCDAPSIM and its verification against publically available TRACE and WCOBRA-TRAC results. Core is now modeled with SCDAP fuel and control components. It is divided into 6 radial rings and 20 axial nodes. The equivalent flow area in the six group of assemblies is modeled as RELAP5 pipe components 108, 110, 112, 114, 116 and 118. Each volume of these channels is connected to the respective outer channel using cross flow junctions as shown in Fig. 2. Channel 108 is modeled with maximum radial power peaking representing the single assembly at the center. Channels 110, 112, 114, 116, 118 represent a group of 24, 30, 24, 42, 36 assemblies in radially outward direction. Each group contains a fuel component and a control compo-
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Fig. 2. RELAP5/SCDAPSIM nodalization of PHTS of AP1000.
Table 1 Steady state as obtained from null transient. Parameter
RELAP5 value
Unit design
Error, %
Core thermal power (MWt) Core inlet flow rate (kg/s) Core outlet flow rate (kg/s) (loop1/2) Pump velocity (rad/s) Hot leg pressure (MPa) Cold leg pressure (MPa) Core inlet temperature (K) Core outlet temperature (K) SG steam outlet pressure (MPa) SG secondary inlet temperature (K) SG secondary outlet temperature (K) SG secondary side water level (m) SG secondary steam flow rate (kg/s) Pressurizer steam/water volume (m3)
3468 (102%) 15,177 7588.8/7588.8 188.7 15.62 15.93 553.6 594.8 5.6617 499.7 544.7 13.46 957/957 31.24/28.22
3415 15,170 7585 188.5 15.49 15.92 553.8 594.2 5.6609 499.7 545.9 13.46 943 31.15/28.32
0.04 0.05 0.1 0.8 0.06 0.03 0.001 0.014 0.0 0.2 0.0 1.4 0.28/0.35
nent with corresponding number of fuel and control rods in each group. Hence the core consist of the total 13 SCDAP components: 1, 3, 5, 7, 9, 11 represents the fuel rod and 2, 4, 6, 8, 10, 12 represents the control components while 13th component is a shroud component corresponding to core shroud surrounding the core. Radiation enclosure is modeled corresponding to each channel to simulate bundle to bundle radiation heat transfer for which the view factors are set to be calculated by the code. A matrix of 17 17 fuel bundle in the core is the input with elements similar to an assembly. Decay heat has been calculated based on 1979 ANS standard data, ANS79-3 which specifies the three isotopes, 235U, 238 U, and 239Np.
The steady state conditions after running a null transient for 2000 s are reported in Table 1. They closely match with the design data (Westinghouse, 2011d) and the relative error is less than the acceptable limits (OECD, 2004). 5. Results and discussion Transients have been run for 5 h (18,000 s) for both the low pressure (due to pressurizer safety valve stuck open) and high pressure (accompanied by continuous opening and closing of SRV) case. SBO starts with a total loss of feed water due to loss of offsite power which then leads to closer of main steamline iso-
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5.1. Decay and oxidation heat
lation valve (MSIV) and tripping of reactor coolant pumps. Reactor scrams due to high pressure (16.96 MPa) on primary side. Decay heat continues to provide heat to the RCS. All the safety systems including automatic depressurization system (ADS) are assumed unavailable. Steam line safety valves do not get actuated during these transients. This section compares the fuel damage progression for these two cases. High pressure and low pressure cases are referred as the cases of no depressurization and depressurization. These calculations have been limited to in-vessel phase of accident progression which is heating and melting of the fuel. It has been reported by Sehgal (2012) that the time span available for severe accident mitigating actions like core re-flooding is of the order of 2 h while it is 4–5 h to restrict the core melt progression.
Transient starts at 2010 s and reactor trips with in 5 s of the transient which results into decay heat of about 200 MWt (6% of power). It decreases sharply during initial 1800 s (30 min) and reaches about 65 MWt (1% of power) at this time. The behavior of decay and oxidation heat has been depicted in Fig. 3. RELAP5 calculated value of decay heat has also been compared with that reported by Westinghouse for SBO (Westinghouse, 2011e). They are in good agreement except for the first 1800 s: the fall of decay heat is steeper than the reported. Here, it is important to note that oxidation heat is quite large in magnitude in comparison to decay heat for some portion of the transient. Maximum oxidation heat is 1357 and 128 MWt in depressurized and high pressure progression respectively.
Decay heat (MWt)
Oxidation heat (MWt)
1360 1352 300 250 200 150 100 50 0 225
No depressurization Depressurization Reported (Westinghouse)
200 175 150 125 100 75 50 0
2000
4000
6000
8000
10000
15000 18000
Time (s)
Hydrogen generation (kg/s) Total hydrogen (kg)
Fig. 3. Decay and oxidation heat.
800
747 kg
600
425 kg 400 200 0 10 8
9.22 kg/s @ 560 s No depressurization Depressurization
2.5 2.0 1.5 1.0
0.87 kg/s @ 7030 s
0.5 0.0 0
2000
4000
6000
8000
10000
Time (s) Fig. 4. Total hydrogen and rate of generation.
15000 18000
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5.2. Hydrogen generation
5.3. Temperatures
Fig. 4 shows the hydrogen generation rate and total hydrogen produced during the transient. The maximum rate for high pressure and depressurized case is 0.87 (@ 7030 s) and 9.22 (@ 560 s) kg/s while total hydrogen is 747 and 425 kg respectively. Time at which oxidation of steam and zircaloy starts is quite different for the two cases which is mainly due to the RCS pressure. For the high pressure case, progression is delayed as water does not boil instantaneously while in the depressurized case, steam is quickly available for oxidation. Total hydrogen for the high pressure case is more than 75% of the depressurized case (although the rate is quite high in this case). Figure also shows that almost no hydrogen getting produced after 8000 s for the depressurized case and after 11,000 s for the high pressure case.
The maximum core surface temperature in SCDAP is a variable (bgmct) which represents maximum cladding temperature up to the melting of cladding (2200 K) and a hot spot in the core afterwards. It has been reported in Fig. 5: 3096 K @ 15,975 s for the high pressure and 3236 K @ 7380 s for the low pressure case. For the high pressure case, temperature starts increasing (due to decay heat) reaching 1033 K @ 4310 s when melting of Ag–In–Cd alloy (Hofmann, 1999) starts; a temporary decrease in temperature is noticed around this time. As temperature reaches 1500 K @ 6265 s, rapid Zircaloy steam reaction starts leading to uncontrolled temperature escalation and hydrogen generation rate peaking to 0.87 kg/s @ 7030 s. The liquefied control materials (including Inconel spacer grids) relocate (Allison and Hohorst, 2008) into
18
3236 K @ 7380 s
3126 K
3250
16
Pressurizer pressure (MPa)
2790 K @ 7140 s
3000 2750
2831 K @ 855 s
2500
3096 K @ 15975 s
2250
No depressurization Depressurization
2000 1750 1500
1500 K @ 6265 s
1033 K @ 4310 s
1250 1000 750
No depressurization Depressurization
16.5 MPa@1740 s
14
18
12
16
12.58 MPa @105 s
10
14 12
Expanded view upto 2000 s
10
8
8 6
6
4 2
4 0
200
400
600
800
1000 1200 1400 1600 1800 2000
2
0.56 MPa@ 105 s
500 0
2000
4000
6000
8000 10000 12000 14000 16000 18000
0
2000
4000
6000
8000 10000 12000 14000 16000 18000
Time (s)
Time (s) Fig. 5. Core maximum surface temperature.
Fig. 7. Pressurizer pressure.
1800
1255 K@ 10325 s
1600
Vessel lower plenum
1400 1200
No depressurization Depressurization
1000
Wall temperatures (K)
Core maximum surface temperature (K)
3500
800 600 400 2400
1733 K @ 10325 s Pressurizer surge line
2000 1600 1200 800 400 0
2000
4000
6000
8000
10000
12000
14000
Time (s) Fig. 6. Lower plenum and surge line wall temperatures.
16000
18000
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8.4 8.0
Secondary side pressure (MPa)
7.6
SG-1 Pressurizer side
7.2 6.8
No depressurization Depressurization
6.4 6.0 5.6 5.2 8.4 8.0 7.6 7.2 6.8
SG-2
6.4 6.0 5.6 5.2 0
2000
4000
6000
8000
10000
12000
14000
16000
18000
Time (s)
Pressurizer water level (m) Core water level (m)
Fig. 8. SG secondary side pressure.
6
6.24 m @ 2575 s
5
TAF @ 6.2484 m BAF @ 1.9812 m
4 3
2.1 m @ 60 s
2
1.99 m @ 4160 s
1 0
12
12.67 m @ 2510 s 11.82 @ 25 s
10
No depressurization Depressurization
8 6
6.18 m @ 1280 s 1 m @ 85 s 0.08 m @ 1270 s
4 2
0.08 m @ 4740 s
0 0
2000
4000
6000
Time (s) Fig. 9. Core and pressurizer water level.
16000 18000
306
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1.51 m
1.4
0.6
No depressurization Depressurization
1.2
0.4 0.2
0.0=No, 1.0=Yes
1.0
0.0
0.8
-0.2
0.6
-0.4
0.4 -0.6 0.2
0.13 m@ 7775 s
0.13 m@ 855 s
-0.8
Crust supporting molten pool fails
0.8
1.85 m@ 6105 s
1.6
Molten pool radius (m)
lower (colder) part of the assembly. This material freezes and forms limited blockages in the lower portion of each assembly: leading to diversion of the flow from hotter channels into colder channels. This diversion increase due to remaining Zircaloy melts (dip in temperature @ 7140 s) relocating into lower portion of assembly. When, temperature reaches near 2800 K, fuel and oxidized cladding, (U–Zr)–O2 melts and start to form ceramic molten pool (Allison and Hohorst, 2008) at 7775 s. The other case of depressurization leads to flashing of water due to sharp decrease in pressure. Oxidation quickly comes into picture and temperature starts rising steeply reaching to 2831 K @ 855 s with a maximum (3236 K @ 7380 s) and settling to 3126 K at the end of transient. Lower plenum and surge line outer wall temperatures are presented in Fig. 6. For the high pressure case, the surge line temperature reaches 1733 K @ 10,325 s while lower plenum wall temperature is much lower, 1255 K, reinforcing the fact that surge line fails before vessel failure as also noticed by Park et al. (2006) for APR1400. These temperatures are quite lower in the depressurized case hence this kind of scenarios does not pose a threat to the integrity of the vessel (for this 5 h transient) even
1.0
1.8
0.0 0
2000
4000
6000
-1.0 8000 10000 12000 14000 16000 18000
Time (s) Fig. 10. Molten pool radius.
Table 2 Core degradation map after 5 h (18,000 s) of transient.
P = Porous debris
I = Intact fuel
V = RELAP fluid volume now void of fuel
xxMxx = Molten or frozen ceramic pool & blockage occurs in the volume Fuel rod components 1
3
Fuel rod components
5
7
9
11
Axial node
1
No depressurization
3
5
7
9
11
Depressurization
I
I
I
I
I
I
20
V
I
I
I
I
I
I
I
I
I
I
I
19
V
V
V
V
V
I
V
V
I
I
I
I
18
V
V
V
V
V
I
V
V
V
I
I
I
17
V
V
V
V
V
I
V
V
V
I
I
I
16
V
V
V
V
V
I
V
V
V
V
I
I
15
V
V
V
V
V
I
14
V
V
V
V
V
V
I
13
V
V
V
V
V
V
xxMxx
xxMxx xxMxx
xxMxx
xxMxx
xxMxx
I
V
V
V
I
V
V
V
V
I
V
I
I
12
V
V
V
V
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
11
V
V
V
V
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
10
V
V
V
V
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
9
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
8
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
7
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx V
I I
I
6
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
P
I
5
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
P
P
P
P
I
I
4
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
I
I
I
I
I
I
3
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
I
I
I
I
I
I
2
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
I
I
I
I
I
I
1
xxMxx
xxMxx
xxMxx
xxMxx
xxMxx
I
Table 3 Summary of the results after 5 h (18,000 s) of transient. Case
Scenario
Maximum core surface temperature (K)
Maximum hydrogen generation rate (kg/s)
Total hydrogen (kg)
Molten pool radius (m)
1 2
No depressurization Depressurization
3096 @ 15,975 s 3236 @ 7380 s
0.87 @ 7030 s 9.22 @ 560 s
747 425
1.51 1.85
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though it leads to much earlier severe core damage. The initial surge line wall temperature increase for the depressurized case is probably due to the choking of the flow in small diameter pipe (surge line). This two phase conditions has been created by sudden depressurization.
is intact for the high pressure case. In the depressurized case, almost all the core node except outer channel has been melted. The outer channel assemblies have survived because of the shroud surrounding the core which is modeled to radiatively receive the heat from this channel.
5.4. Primary and secondary side pressure
6. Conclusions
Fig. 7 represents the primary side pressurizer pressure. For the high pressure case, pressurizer SRV has been modeled as motor valve to simulate cyclic behavior (Vierow et al., 2004) caused by opening and closing of SRV. It opens when the RCS pressure increases to 17.75 MPa closes when the pressure decreases to 17.13 MPa with a valve change rate of 5 s 1 (0.2 s to change the valve position) which is the rate of change of the normalized valve area as the valve opens or closes. RCS pressure initial decrease (12.58 MPa @ 105 s) is primarily due to reactor SCRAM which in turn takes about 2000 s to boil the RCS water inventory. This SRV has been simply modeled as trip valve (to simulate stuck open behavior) which fully opens (instantaneously) when pressure reaches 17.75 MPa for the case of depressurization (low pressure). This SRV stuck open scenario has been taken from the TMI-2 accident scenario. RCS very quickly depressurizes (0.56 MPa @ 105 s) due to high pressure coolant expulsion from the SRV. Fig. 8 represents the pressure on the secondary side: for high pressure case, SG PORV gets actuated @ 7.92 MPa instantaneously and both the SGs show an expanded cycling time. For the low pressure case, pressure profiles for two SG’s are different because of the large coolant discharge (from pressurizer safety valve) on SG-1 side which provides additional cooling and results in little more depressurization on the secondary side.
SBO in AP1000 with and without depressurization has been studied and the summary of the transient results are reported in Table 3. Decay heat based on RELAP5 calculation has also been compared with that of Westinghouse for SBO. They are in good agreement except during the first 1800 s: the RELAP5 decay heat is steeper than the reported. The calculated value of maximum core surface temperature is 3096 K and 3236 K for the high and low pressure cases respectively. These two cases give maximum hydrogen generation rate of 0.87 and 9.22 kg/s with total H2 generated being 747 and 425 kg. Significant differences in the damage progression have been observed for the two cases. The depressurized SBO looks quite superior as less hydrogen is produced and almost no threat to integrity of the vessel is posed. The drawbacks are the early start of fuel damage and almost all the core except outer channel has melted after this 5 h transient. The high pressure SBO has the advantage of delaying the core damage progression which can be arrested using severe accident management guidelines like core flooding otherwise this poses a threat to vessel integrity; this simulation also reinforces the fact that surge line failure occurs before the vessel. The progression is observed to be very sensitive to steady state pressurizer water volume specially with respect to hydrogen hence it should be modeled as accurately as possible.
5.5. Core and pressurizer water level Fig. 9 shows the water level in the core and pressurizer. Water level is measured from bottom of the lower plenum to top of the active fuel (TAF @ 6.2484 m) with bottom of the active fuel (BAF) at 1.9812 m while that of pressurizer is from bottom to top of cylinder. Pressurizer water level decrease is due to initial pressure dip for high pressure case and later @ 1800 s, the vapor is vented through SRV and simultaneously liquid is forced (Vierow et al., 2004) to occupy whole pressurizer volume. As the core starts to uncover, large amount of steam is generated accompanied by decrease in water level. The core water follows the pressurizer water level sinking after 3000 s reaching the BAF at 4160 s (1.99 m) and lower plenum completely vaporizes at 6000 s. In the depressurized case, pressurizer water level continually decreases and bottom is completely voided in 1270 s; core water level quickly decrease to BAF (2 m @ 60 s) due to early core uncovery and leveling off in the lower plenum @ 0.2 m at the end of transient. 5.6. Molten pool radius and core degradation map The molten pool radius has been presented in Fig. 10 for both the transients. For depressurized case molten pool starts getting formed very early in the transient with a value of molten pool radius of 0.13 m @ 855 s which achieves a value of 1.85 m @ 6105 s, confirming the early uncovery of the core. For high pressure case, it reaches to 1.51 m at the end of the transient and showing that core uncovery is delayed; molten pool (radius = 0.13 m) started to form at 7775 s. Figure on the right axis shows the parameter for behavior of crust supporting the molten pool. Its value is 0 for both the cases which means that crust does not fail. Core degradation map after 18,000 s is depicted in Table 2 which clearly shows that one third of the core including outer channel
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