ML20024H523
| ML20024H523 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 05/31/1991 |
| From: | Martin R EG&G IDAHO, INC. |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-A-1050 EGG-2633, NUREG-CR-5663, NUDOCS 9106040385 | |
| Download: ML20024H523 (76) | |
Text
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l Idaho National Engineering Laboratory EG&G Idaho. Inc.
Prepared l'or U.S. Nuclear Regulatory Coinmission l
l 9106040385 910531 PDR NUREO CR-5bb3 R PDR
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NURiiG/ Cit-5663 11GG-2633 11 4 RELAP5 Thermal-Hydraulic Analysis of the WNP1 Pressurized Water Reactor hianuscript Completed: April 1991 Date Published: hiay 1991 Prepared by R. P. htartin Idaho National Engineering laboratory hianaged by the I).S. Departrnent of linergy EG&G Idaho, Inc.
Idaha Falls,ID 83415 l
Prepared for Division of Systems Research Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN A1050 Under DOE Contract No. DE-AC07-761D01570
)
ABSTRACT Thermal-hydraulic analyses of five hypothetical accident scenarios were performed with the RELAP5/M003 computer code for the Babcock & Wilcox Company Washington Nuclear Project Unit 1 (WNP1) pressurized water reactor.
This work was sponsored by the U.S. Nuclear Regulatory Commission (NRC) and is being performed in conjunction with future analysir work at the NRC Technical Training Center in Chattanooga, Tennessee.
The accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Technical Training Center WNPl simulator to model abnormal transient conditions.
iii
CONTENTS ABSTRACT iii ACKNOWLEDGMENTS..........................,..
ix ACRONYMS xi EXECUTIVE
SUMMARY
I 1.
INTRODUCTION.........................
3 2.
MODEL DESCRIPTION 5
2,1 Thermal-Hydraul i c Model......,,............
5 2.1.1 Primary System 5
2.1.2 Secondary System 13 2.2 Control System Model 15 2.3 Steady State Conditions....................
16 3.
SCENARIO 1:
LOSS OF AC POWER 18 i
i 3.1 Scenario Description 18 3.2 Calculation Assumptions 18 3.3 Calculation Results...................,..
19 I
i 4.
SCENARIO 2:
SMALL BREAK LOCA WITH LOSS OF AC POWER 26 l
l 4.1 Scenario Description 76 4.2 Calculation Assumptions. <..................
26 4.3 Calculation Results...............,,,....
27 1l-l 5.
SCENARIO 3:
STUCK-0 PEN PRESSURIZER SAFETY VALVE,,........
36
-5.1 Scenario Description 36 5.2 Calculation Assumptions....,
36 5.3 Calculation Results......................
37 6.
SCENARIO 4:
MAIN STEAMLINE BREAK WITH STEAM GENERATOR l
TUBE RUPTURE,........,
44 6.1 Scenario Description 44 6.2 Calculation Assumptions....................
44 6.3 Calculation Results......................
46 7.
SCENARIO 5:
LOSS OF MAIN FEE 0 WATER WITH DELAYED SCRAM.......
54 7.1 Scenario Description 54 7.2 Calculation Assiimptions..
54 7.3 Calculation Results......
55 y
8.
CONCLUSIONS AND RECOMMENDATIONS 62-9.
REFERENCES.
63 FIGURES 1.
Nodalization of WNP1 primary coolant Loop A.....
9 2.
Nodalization of WNP1 primary coolant Loop B.............
10 3.
RELAP5 model of the WNP1 reactor vessel.
11 4.
Nodalization of the WNP1 secondary system, 5.
Scenario 1: primary and secondary pressure..........,..
21 6.
Scenario 1: primary hot and cold leg temperature......,...
21 7.
Scenario 1: pressurizer level.
22 8.
Scenario 1:
steam generator narrow range levels.
22 9.
Scenario 1:
auxiliary feedwater flow................
24
- 10. Scenario 1: primary hot leg mass flow rates.
24 g
- 11. Scenario 1:
system heat transfer..............,.
25
- 12. Scenario 2: primary and secondary pressure 29 13.
Scenario 2: primary hot and cold leg temperature 29 14.
Scenario 2:
pressurizer level...................
30
- 15. Scenario 2:
steam generator narrow range levels..........
30 16.
Scenario 2:
auxiliary feedwater flow 31 17.
Scenario 2: primary hot leg mass flow rates............
31 18.
Scenario 2:
reactor vessel upper head void fraction.
33 19.
Scenario 2:
hot leg high point void fraction............
33 20.
Scenario 2:
break mass and volumetric flow rates..........
-34
- 21. -Scenario 2:
representative steam generator void fractions.
34 22.
Scenario 3:
primary and secondary pressure 39 23.
Scenario 3:
pressurizer l evel..
39 I
vi l'
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24.
Scenario 3:
safety valve and total ECCS mass flow rates......
40 25.
Scenario 3:- primary hot and cold leg temperature 40 26.
Scenario 3:
steam generator narrow range levels..........
41 27 Scenario 3:
auxiliary feedwater flow rate.............
41 28.
Scenario 3:
primary hot leg mass flow rates............
43 29.
Secondary renodalization for steamline break............
45 30.
Scenario 4:
steam generator pressure 48 31.
Scenario 4:. Loop B hot and cold leg temperature..........
48 1
32.
Scenario 4:
Loop A hot and cold leg temperature.......,..
50
- 33. Scenario 4:
steam generator narrow range levels..........
50 34.
Scenario 4:
pressurizer pressure 51 35.
Scenario 4:
ECCS mass flow rates 51 36 Scenario 4:
steam generator tube rupture mass flow rates 52 37.
Scenario 4:
steamline break mass flow rate 52 38.
Scenarlo 5:
primary and secondary pressure 57 39.
Scenario 5:
pressurizer safety and relief valve mass flow rates,.
57 40.
Scenario 5:
primary hot leg mass flow rates,...........
58 41.
Scenario 5:
steam generator narrow range levels....
58 42.
Scenario 5:
system heat transfer at steam generator........
59 43.
Scenario 5:
reactor power.....................
59
- 44. Scenario 5:
pressurizer level...................
61 45.
Scenario 5:
primary hot and cold leg _ temperatures.........
61 TABLES 1.
_ Summary of scenarios analyzed 4
2.
. Correspondence between the physical and mathematical components in the primary and secondary loops for the RELAp5 model of WNP1 6
I vil
1.
Correspondence between the shysical and mathematical components in the reactor vessel for tie RELAP5 model of WNP1.........
8 4.
Comparison of the RELAP5 and simulator initial conditions 17 5.
Transient sequence information for Scenario 1 20 6.
Transient sequence information for Scenario 2 28 7.
Transient sequence information for Scenario 3 38 8.
Transient sequence information for Scenario 4 47
-9.
Transient sequence information for Scenario 5 56 viii i
ACKNOWLEDGMENTS The author gratefully acknowledges the assistance of Steve Showe and Larry Bell at the NRC Technical Training Center for their advice and direction
,in formulating the accident scenarios in this document.
The assistance and critical review of this document by John Burtt, Nels Jensen. Don Fletcher, Craig Kullberg, Steve Polkinghorne, Ron Beelman, and Cliff Davis were also appreciated.
i l
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ACRONYMS B&W Babcock and Wilcox Company ECCS emergency core cooling system ECI essential controls and instrumentation ESFAS engineered safety features actuation system F0GG.
feed-only-good generator FSAR final safety analysis report HPI.
high pressure injection ICS integrated control system INEL Idaho National. Engineering Laboiatory LOCA loss-of-coolant accident MSIV main steam isolation valve NRC U.S. Nuclear Regulatory Commission OSO operation system description OTSG once-through steam generator PORV power-operated relief valve PWR pressurized water reactor RCS reactor coolant system RPS reactor protection system
.SGTR steam generator tube rupture SRV safety relief valve TTC Technical Training Center TVA Tennessee Valley Authority WNP1 Washington Nuclear Project Unit 1 xi
RELAP5-THERMAL-HYDRAULIC ANALYSIS OF THE WNP1 PRESSURIZED WATER REACTOR EXECUTIVE
SUMMARY
-In 1979, the U.S. Nuclear Regulatory Commission (NRC) adopted recommendations from the Kemeny Commission requiring that all nuclear plants have a plant-specific simulator for operator training with the capability to model plant operation and transients in an environment closely resembling the plant control room.
Today's simulators have evolved to become a reliable souice for simulation of normal plant operational transients.
However, the current simulators have produced incorrect responses to or have been unable to model many other transients and accidents.
Keeping to the policy established in 1979, the NRC has initiated a project that examines the capabilities of the current generation of simulators. The focus of this initiative is the
-evaluation of simulator capabilities under uniquely challenging transient scenarios measured against the predictions of advanced thermal-hydraulic system codes, such as RELAP5 and TRAC-BWR.
The Babcock and Wilcox Company Washington Nuclear Project Unit 1 (WNPI) simulator, located at the NRC Technical Training Center, was modeled using RELAP5/M003. The model, a two-loop pressurized water reactor (PWR), contained detailed thermal-hydraulic representations of the pertinent PWR primary and secondary systems, it iing the emergency core cooling system and-steamlines.
Detailed models of the key plant control systems were included.
The RELAP5 model was used to analyze five separate transients, selected to cover a wide. range of possible thermal-hydraulic conditions that could occur in a reactor accident. The transients were (a) loss of ac power, (b) small break loss-of-coolant-accident with loss of ac power, (c) failed-open pressurizer safety valve, (d) main steamline break with steam generator tube rupture, and (e) loss of feedwater without scram.
In general, the calculated RELAP5 trends were reasonable for the scenarios studied in the analysis, and will provide a good basis for 1
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Some uncertainties concerning boundary conditions and modeling options have not been resolved and could affect simulator /RELAPS comparisons, s
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1.
INTRODUCTION One of the lessons learned from the accident at Three Mile Island was the recognition of the need for effective reactor operator training.
The President's Commission on the Accident at Three Mile Island (the Kemeny Commission)' recommended that all plants have access to a plant-specific simulator for operator training.
Such a simulator should be capable of modeling plant operation and transients in an environment that closely resembles the plant control room.
Additionally, the Kemeny Commission recommended that research and development be carried out for improving simulation and simulation systems to establish and sustain a higher level of realism in operator training and to improve the diagnostics and general knowledge of nuclear power plant systems.
The U.S. Nuclear Regulatory Commission (NRC) adopted these recommendations following release of the Kemeny Commission's report.
Today's simulators have evolved to become a reliable source for simulation of normal plant operational transients. However, the current simulators have produced incorrect responses to or have been unable to model many other transients and accidents; thus, defeating the designed intent of the simulator.
Keeping to the policy established in 1979, the NRC has initiated a project that examines the capabilities of the current generation 0T simulators. The focus of this initiative is the evaluation of simulator capabilities under uniquely challenging transient scenarios measured against the predictions of advanced thermal-hydraulic system codes, such as RELAPS and TRAC-BWR. The simulators to be evaluated reside at the NRC Technical Training Center (TTC) in Chattanooga, Tennessee.
The TTC uses three resident simulators, representing specific Westinghouse, Babcock and Wilcox (B&W), and General Electric plants; in addition, the TTC has use of a Combustion Engineering simulator in Windsor, Connecticut. The scope of this project involves developing advanced system code models of the plants, performing a series of transient calculations with the models, and comparing the code results with simulator results, both before and af ter scheduled simulator upgrade.
l 3
This report documents the RELAP5 transient analyses of the B&W Washington Nuclear Project Unit 1 (WNP1) simulator.
Table 1 describes the five scenarios analyzed.
This report will discuss only the code results; comparison with simulator data will follow at a later date.
Section 2 of this report contains a detailed description of the RELAPS/M003 WNP1 model used in the analyses.
Sections 3 through 7 document the model changes, the assumptions specific to each scenario, and the calculated results for each of the five scenarios.
Section 8 contains a discussion of the conclusions drawn from the analyses, and a reference list is provided in Section 9.
Table 1.
Summary of scenarios analyzed Scenario initiatina Event 1
Loss of ac power (loss of offsite power with diesel generator failure) 2 Small break loss-of-coolant accident (1000 gpm initially) with loss of ac power 3
Failed-open pressurizer safety relief valve j
4 Double-ended main steamline break with a steam generator tube rupture 5
Loss of feedwater pumps with a delayed scram 4
2.
MODEL DESCRIPTION A description of the input model used to represent the WNPl plant in the RELAP5 calculations follows.
This discussion focuses on the modeled thermal-hydraalic components, the control system, and the steady-state 2
initialization. The RELAP5/M003 computer code was used to perform these calculations.
A RELAP5 model of the Tennessee Valley Authority (TVA)
Bellefonte plant,3 which was developed at the Idaho National Engineering Laboratory (INEL), represented the foundation of the WNP1 model and nodalization scheme. The WNP1 and the Bellefonte plant designs are based on similar B&W specifications.
The model incorporates available WNPl-specific information gathered from the TTC.
I 2.1 THERMAL-HYDRAULIC MODEL l
l The RELAP5 WNP1 model includes all of the major system components.
Specific features modeled include all major primary system coolant loop flow paths, secondary system main feedwater paths downstream of the main feedwater I-valves, and secondary main steam paths upstream of the turbine stop valves, including the main steam isolation valves.
Modeling also included the emergency core cooling and the auxiliary feedwater systems on the primary and secondary sides, respectively.
The WNPI RELAP5 steady-state model used 190 control volumes, 197 junctions, and 195 heat structures to simulate the nuclear steam supply system.
Tables 2 and 3 summarize the correspondence between the reactor system and the model components.
Figures 1 through 4 l
illustrate the RELAP5 model nodalization scheme.
2.1.1 Primary System The WNP1 plant has two primary loops, designated as Loops A and 8, and both are represented in the RELAPS model.
As shown in Figures 1 and 2, each modeled loop includes a hot leg, a steam generator, two pump suction legs, two I
reactor coolant pumps, and two cold legs.
The pressurizer and pressurizer spray lines are attached to Loop A.
The safety relief valves (SRVs) and the power operated relief valve (PORV) connected to the pressurizer steam dome 5
9 Table 2.
Correspondence between the physical and mathematical components in the primary and secondary loops for the REl.AP5 model of WilPI Physical Component RElAPS Component (si l
L299 A Hot leg 100, 108, 110 Steam generator inlet plenum 115 Steam generator tubes 120 Steam generator outlet plenum 125 A 1 pump suction leg 160 A-2 pump suction leg 130 A-1 reactor coolant pump 165 A 2 reactor coolant pump 135 A 1 cold leg 170, 175, 180, 181 A 2 cold leg 140, 145, 150, 151 Loop B Hot leg 200, 208, 210 Steam generator inlet plenum 215 Steam generator tubes 220 Steam generator outlet ;)1enum 225 B-1 pump suction leg 260 B-2 pump suction leg 230 B-1 reactor coolant pump 265 B-2.eactor coolant pump 235 B 1 cold-leg 270, 275, 280, 281 B 2 cold leg 240, 245, 250, 251 Pressurizer r
Surge line 600, 601, 605 Pressurizer 610 Pressurizer dome 615 Spray line 620 Spray valve 616 Power operated relief valve 804 Safety relief valve 802 Containment backpressure for 803, 805 relief valves l
6
Table 2.
(continued)
Phvsical COrpanint
_ REl APJ CompArLtn1[5.L n
Steam Generater A feedwater line Main feedwater valve 700, 827 Downcomer 705 Tube bundle 300 Steam downtomer 310 Main steamline 3)c 320
- 30, 340, 350, 370 Turbine stop valve 365 Main steam isolation valve Safety relief valves 335 811 Modulating atmospheric dump valves 809 Containment backpressure for relief and dump valves 810, 812 Steam Generator B Feedwater line Main feedwater valve 750, 927 Downtomer 755 Tube bundle 400 Steam downcomer 410 Main steamline 415, 420 425, 430, 440, 450, 460, 470 Turbine stop valve 465 Main steam isolation valve Safety relief valves 435 911 Modulating atmospheric dump valves 909 Containment backpressure for relief and dump valves 910, 912 N
Correspondence between the physical and mathematical components in Table 3.
the reactor vessel for the RELAPS model of WNPI RELAPS Component (s)
Physical Component 555, 557, 560, 562, 565, 567 Inlet annulus 570 Downtomer Lower plenum 505, 575 515 Core 510 Core bypass 520, 525, 530, 535, 538, 540, 545 Upper plenum 550 Vpper head Vent valve 536 8
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Nodalization of WNP) primary coolant Loop A.
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RELAP5 model of the WNP1 reactor vessel.
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II 360 Steam Header Feedwater 952 300 311 Boundary 3
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12 V Turbine Stop Condition Zh Valve 370 Steam Boundary Condition Figure 4. _Nodalization of the WNPl-secondary system.
12
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i provide for primary system pressure relief when necessary, lhe steam generators are the B&W once-through design.
The primary and secondary sides of the boiler region thermally communicate through twelve axial heat structures representing the metal mass of the steam generator tube barrier.
The four reactor coolant pumps are Bingham Willamette pumps.
pump rated conditions, pump friction data, and the single and two-phase head and torque data were taken directly from the TVA Bellefonte input model.
A high pressure injection (HPI) port, which injects one-fourth of the total HPI, is attached to each cold leg.
The makeup and letdown system, which is equivalent to the chemical and volume control system in other pressurized water reactors, is l
also attached to the Loop A cold leg. A single junction and a control system represent the combined functions of the makeup and letdown system, lhe model includes heat structures connected to each volume in the primary loop to represent the metal mass of the piping, Heat structures also are included to model the pressurizer proportional and backip heaters.
2 figure 3 shows the nodalization scheme used for the reactor vessel, Modeling of the upper head includes a flowthrough volume with a junction representing the flow from inside the plenum cylinder to the upper head via guide weldment penetrations. A second junction represents the drain holes that allow flow from the upper head to the region between the plenum cylinder f
and the core support cylinder, As is common to B&W design vessels, eight reactor vent valves, which are fully open at a differential pressure of 0,125 psid (862 Pa), provide pressure relief from the upper plenum to the downcomer.
Heat structures attached to the reactor vessel model the reactor vessel walls, 4
core barrel, core baffles and 'nrmer plates, plenum cylinder, internal supports and guide structures, and the nuclear fuel.
A power table can respond to a scram signal to initiate a control rod insertion simulation based on the 1979 American National Standard for decay power,'
2.1.2 Secondary System figure 4 shows the RELAPS WNPl modeling of the secondary system, lhe RELAPS once-through steam generator (OTSG) model, shown in figure 4, contains the major flow paths in the OTSG and includes the downcomer, boiler riser region, and baffle.
Both SRVs and modulating atmospheric dump valves provide i
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overpressure control for the steam generator secondary side.
Turbine stop valves and main feedwater and steam isolation valves provide secondary side isolation.
Heat structures modeled in the secondary side represent only the l
steam generator attal mass and a small section of the steamline adjacent to J
the steam generator.
Table 2 presents the relationship between the steam j
generator secondary side physical components and the corresponding mathematical components of the RELAP5 model.
figure 4 also shows the major hardware components of the steamline out to a
the turbine stop valves.
The main steam isolation valve, the SRV, and the modulating atmospheric dump valves are contained within the steamlines_from i
the steam generator.
The steamline model combines the two steamlines in each loop into one.
The appropriate control logic to simulate the operation of l
individual valves is also included in the model.
The feedwater system is modeled as a boundary condition (or the secondary side. A control variable in a time-dependent junction governs the feedwater flow from the feedwater pumps to the steam generators.
For closure of the main feedwater isolation valves, a control variable linked to the nominal valve area varies from one to zero within one second.
4 The auxiliary feedwater system includes a specific model for the purpose of providing steam generator level control in response to an essential controls and instrumentation (ECI) signal. The model includes control variables to determine secondary level, time-dependent volumes and junctions to simulate auxiliary feedwater injection, and trips to initiate the system with best-estimate delays for pump turbine powerup (assumed to be 15 seconds)-
and instrumentation response times.
A time-dependent junction provides'a mass flow boundary condition equivalent to the total mass flow provided by the two i
motor-operated and the one turbine operated auxiliary feedwater pumps.
The time-dependent junction relies on a control variable to calculate the appropriate auxiliary feedwater flow rate, for lack of specific information, I
the model incorporated the assumption that the maximum auxiliary feedwater flow is equivalent to 1200 gpm at_the inlet of the steam generator.
This is 1
equivalent to 900 gpm at the outlet of the pumps.
The 1200 gpm value is 14
documented in the WNP1 operation system description (050)' for the capacity of the two motor operated or the one turbine-operated auxiliary feedwater pumps.
2.2 CONTROL SYSTEM MODEL Because of the proprietary nature of most of these data, this discussion does not include detailed information regarding relevant setpoints and time constants.
Setpoints from data provided by the WNP1050 and the Bellefonte final safety analysis report (FSAR)6 were included in the trip logic.
Because of the scope of this project, certain WNPl control systems were not modeled.
Specifically, an explicit model of the integrated control system (ICS) does not exist.
For the transients documented in this report, these control systems were assumed inoperative or not challenged and therefore, not required in the model.
The main control systems used to model the WNP) transients were the reactor protection system (RPS), the engineered safety features actuation system (ESFAS), and the EC1 system.
Other control systems modeled were the pressurizer pressure, the pressurizer level, and the feed only-good generator (FOGG) control system.
Physical process instrumentation delay times that exist in an actual plant were not available for the WNP1 model and were not modeled.
The RPS is responsible for identifying abnormal system conditions that are typically precursors to a more serious system problem that could threaten the reactor fuel and its cladding. Upon identification of these abnormal conditions, the RPS will signal for an automatic reactor shutdown. Relevant RPS parameters monitored include reactor power, reactor coolant pressure, hot leg temperature, and pressurizer level.
Other RPS control logic not challenged in the transients documented was not included in the model.
The ESFAS handles the control of safety systems responsible for reducing the consequences of accidents.
Features of the ESFAS included in the model are the emergency core cooling system (ECCS), the steamline isolation, and the steamline auxiliary feedwater system.
Relevant parameters modeled include 15 i
i
OTSG pressure and level. The ESFAS model does not include ESTAS features unchallenged for the transients perfonned in this study.
The ECl system provides for the indication, control, and interlock features required to place the plant in a safe shutdown. The only active feature of the EC1 system modeled here was the manipulation of auxiliary feedwater flow (valves) to control the OlSG level.
The ECl system also ensures power for many of the instruments us' ' 'v the other control systems.
Additional control systers maintain the pressurizer pressure and pressurizer level.
Control of the pressurizer pressure is by banks of pressurizer heaters, spray valves, and the relief valves responding to low and high pressure setpoints. Control of the p*essurizer level is by the makeup and charging system from a comparison of actual level to desired level.
N.o system responds to an ESFAS signal to reduce the consequences of a<u break upstream of the main steam isolation valves.
Since a steam ' ~ oreak can only be terminated by allowing the affected OTSG to boil dry, the FOGG system identifies an affected generator and closes the auxiliary feedwater supplies to that generator only.
2.3 STEADY STATE CONDITIONS A steady-state initialization was performed with the RELAPS WNP1 model.
Table 4 presents the comparisons with the simulator data, representing full-power conditions.
Quantitative results, except for the actual power magnitude, were available directly from steady-state simulator results supplied from the TTC. Other supporting documentation provided by the TTC indicates that the WNP1 simulator operates at a power of 98% of 3800 MW rated full power (3684.8 MW).
P 16
j Table 4.
Comparison of the RELAP5 and simulator initial conditions I
P1anLf aramet er ELmulAter BR!LP_L Reactor power (MW) 3684 3 3684.8 Primary pressure (psia) 2210.0 2210.1 Pressurizer level (in.)
220.0 220.3 Primary loop flow (lb/s) 20969.4 20959.0 Average hot leg 629.3 629.0 temperature (*f)
Average cold leg 572.4 571.4 temperature (*f)
Steam generator 1061.7 1077.2 outlet pressure (psia)
Steam generator level (in.)
80.0 80.5 feedwater temperature (*f) 464.3 464.3 Steam temperature ('f) 602.2 600.0 17
3.
SCENARIO 1:
LOSS OF AC POWER The following section details the analysis of a loss of alternating current (ac) power in the WNP1 plant initiated at full power.
The subsections contain a description of the scenario, calculational assumptions, and an analysis of the results.
3.1 SCENARIO DESCRIPTION The loss of ac power involves the loss of offsite power accompanied by a failure of the diesel generators.
f ailure of the diesel generators made the pumped ECCS unavailable when the reactor coolant system (RCS) pressure reached the appropriate setpoint (e.g., from a loss of primary-coolant inventory).
The loss of offsite power resulted in the automatic shutdown of the reactor in addition to the automatic shutdown of the turbine, isolation of the lcidown and charging makeup system, deactivation of the pressurizer heaters, and coastdown of the reactor coolant, feedwater, and condensate pumps.
No operator intervention was modeled for this transient.
3.2 CALCULATION ASSUMPTIONS The basic WNP1 model described in Section 2 was used to perform the calculation.
Direct current (de) battery power is assumed available for the duration of the transient for the operation of equipment controlled by the ECl system (e.g., steam generator icvel controller).
Following initiation of this event, reactor coolant pumps lose power and coast down, a reactor trip occurs on the loss of the electromagnetic charge holding the control rods, and the turbine stop valves close on the signal of the reactor trip.
Loss of power also causes the loss of the condensate pumps (not modeled).
The effect of this loss is low suction head for the turbine-operated main feedwater pumps, resulting in a trip of these pumps.
Since these pumps are not explicitly modeled, a coastdown table is used to ebulate this event.
The assismed trip time for these pumps is five seconds after initiation of the event.
The atmospheric dump valves controlled by the ICS and the turbine bypass valve will fail closed and remain in that position.
With these pressure relief paths removed, the only means of energy release from the steam generators will
be from the two banks of safety valves and the modulating atmospheric valves, which open at 1280, 1250, and 1220 psia, respectively.
The turbine operated auxiliary feedwater pump (motor auxiliary feed pumps will not be available) will then respond to an Ecl signal from the loss of both main feed pumps i
following a startup delay of 15 seconds.
The auxiliary feedwater provides enough feed flow to keep a steam generator liquid level of 6 f t until the actuation of an ESFAS signal.
The [SFAS will typically respond to low steam generator pressure (<600 psig), low RCS pressure (<l600 psig), or high containment pressure (>4 psig).
When any of these conditions are true, closure of both the main fecawater and main steam isolation valves isolate the steam generators.
Concurrently, the CSfAS signal terminates the steam generator level control and auxiliary feed f'ow signals to provide the maximum 4
flow (600 gpm per steam generator).
This will lead to an overfill of the OTSG.
3,3 CALCULATION RESULTS i
Table 5 provides a summary of the sequence of events that occurred in the loss of ac power transient (the calculated times are rounded off to the nearest second).
The immediate effect of the loss of ac power was the release of the control rods into the core, which automatically tripped the reactor.
The turbine stop valves then closed on the turbine trip signal that actuated from the reactor trip, Primary system pre!.sure (figure 5) decreased from 2210 psia to 1970 psia because of the sudden reduction in thermal energy supplied by the reactor.
Likewise, hot leg temperature (Figure 6) and pressuriter level (Figure 7) decreased.
In the secondary system, pressure (figure 5) increased in response to the closure of the turbine stop valve.
No flow bypassed the turbine to the condenser, since the turbine bypass valve failed closed on loss of electrical cower.
The loss of ac power to the condensate pumps caused the feedwater pumps to trip on suction pressure, lhe steam generator liquid level (Figure 8) initially increased because of reduced heat transfer from the primary system and the f act that the turbine stop valve had 4
closed, resulting in a sudden pressure increase.
A quick decrease in steam generator level followed as the main feed pumps coasted down and the main steam safety valves (open at 1250 psia) and modulating atmospheric dump valves (oran at 1220 psia) relieved high secondary pressure.
As the main feedwater 19
Table 5.
Transient sequence information for Scenario 1 Time L11_
t:ent 0
Loss of offsite power results in reactor trip, reactor coolant pumps trip, charging and letdown system is isolated, pressurizer heaters trip, feedwater pumps trip, turbine stop valve closes, steam driven auxiliary feedwater signal is generated 5
Steam generator modulating atmospheric relief valves and safety relief valves open 20 Temporary repressurization of primary side begins 400 Primary loop flows complete transition to natural circulation conditions; primary system begins slow depressurization and cooldown 1800 Calculation is terminated i
l 4
20
o e,i...
c
-.......e.,
2 1
2 t i....
tuu i...
i...'.
ti m M i.............. s e........ i.... n o e. io,e i.co e i...
Tim. (..c)
Figure 5.
Scenario 1:
primary and secondary pressure.
...e u
l 2 c.i. t..
n.u..
/
- N e. i...
5 3
..o.o K
E e
..a o j\\
.o o..
.........io....i....
i.... i.... i....
Tim. (..c)
Figure 6.
Scenario 1:
primary hot and cold leg temperature, 21
o w
i...
7c.
i....
.n nni. (...)
Figure 7.
Scenario 1: pressurizer level.
- i.... l
-i....
?.... n 3,...
f).vg_
_ - - _ A
_ d n m. t..c)
Figurc 8.
Scenario 1:
steam generator narrow range levels.
4 22
..-a
. * ~ -
I a
pumps lost their effectiveness to remove the decay heat from the primary system, the steam generator level dropped as the feedwater boiled.
In response to the loss of the main feedwater pumps, the turbine-operated auxiliary pump began to supply feedwater from the condensate storage tank (and, as a backup, from the essential raw cooling water tank) following 3 20-second startup delay after the signal.
1he auxiliary feedwater :ystem then supplied enough feedwater (figure 9) to maintain the steam generator level at approximately 6 ft (5.7 ft) as signaled from the LCl system.
Following the initial primary depressurization and hot leg t aperature decrease, pressure and temperature began to increase as a result of the coasting down of the reactor coolant pumps and the reduction in loop flow (figure 10).
This coastdown resulted in a reduction in primary to secondary heat transfer, which allewed an increase in hot leg temperature and consequently, an increase in primary pressure.
The steam generator level control system stabilized heat transfer from the primary system to the secondary system (figure 11) by 400 seconds, allowing loop natural circulation flow-to be achieved.
As a result of this transition to natural circulation, f
primary pressure and hot leg temperature also stabilized.
for the remainder of the transient, system conditions preserved this equilibrium, Primary pressure and temperatures slowly decreased as the system e
cooled with the decrease in decay heat.
The pressurizer and reactor vessel 4
provided additional cooling, as heat transfer can also be extended from these
[
components to the environment through the metal mcss.
Secondary pressure oscillated between the open and close pressure setpoints of the modulating atmospheric dump valves.
At 1800 seconds, the transient was terninated when it was concluded that general primary and secondary thermal-hydraulic trends were established.
i in the absence _of meaningful uncertainties for this transient, the results resented represent the best estimate predicatinn of RELAPS.
With the primary Md secondary systems achieving a state of thermal equilibrium within 500 seconds after the loss of ac power, and with the primary coolant remaining subcooled throughout the event, it can be concluded that the loss of ac power
[
transient represents a benign event that is adequa:.ely handled by the built-in s a fe ty - sy s.t em.
L 23 1
o al
1 1
1 o
- to.
a.
t...
- i h
H j....
h,., !,khdk I O( % F~ e h
~
l 9
v.a x~~t-y a:.*t..
u
..n.
..... i.....................
rim.(..c)
Figuro 9.
Scenario 1:
auxiliary feedwater flow.
2
- t. 7. I]
c
.t..
no t.
7..
3 is..
1 C
4,.
-c c
p n
.3.
~
......i...............i.....
Time (..c)
Figure 10. Scenario 1:
primary hot leg mass flow rates.
24
..o o ;
m-i
?
c o Cete in Prlteert i <-~ rrse u to seerneser Olo-o t too.o 7
y l6e o w
t
!, o oo o
\\
ioc o t
f ico o y
mu-T oc o eo soo.o.......... ooe.o tooe.o iseo o i.co.o ione o soon.e um (o.c)
Figure 11.
Scenario 1:
system heat transfer.
25
i 4.
SCENARIO 2:
SMALL BREAK LOCA WITH LOSS OF AC POWER I
The following section details the analysis of a cold leg small break loss of-coolant accident (LOCA) coincident with a loss of ac power in the WNP1 plant. The transient occurs with the reactor at full power.
The subsections contain a description of the scenario, calculational assumptions, and an analysis of the results.
4.1 SCENARIO DESCRIPTION This transient is identical to Scenario 1, with the exception that a small break LOCA accompanies the loss of ac power. On initiation of the transient, the break is to release 1000 gpm of reactor coolant from one uf the cold legs in loop A.
Emergency core cooling is unavailable as a result of the loss of power; thus, the event will eventually lead to core uncovery.
No operator intervention was nod? led for this transient.
4,2 CALCULATION ASSUMPTIONS The basic WNP1 model described in Section 2 was used to perform this calculation, for the cold leg LOCA with concurrent loss of ac power, the WNP1 input model includes modeling of the break in cold leg A-1.
The break size provides an initial break flow of 1000 gpm. A critical flow calculation
- determined break size to be-approximately equivalent to a 1 in.-diameter hole.
The inlet junction to cold leg component 180 incorporated this break model.
For the loss of ac power, the reactor trips from the loss of the electromagnetic charge holding the control rods. Consequently, the reactor trip signals a turbine trip, which results in closure of the turbine stop valve. With the closure of the turbine bypass and stop valves, the modulating i
atmospheric dump valves and main steam safety valves release core energy from the steam generators. Main feed pumps will trip from low suction pressure from the loss-of condenser booster pumps. The turbine-operated auxiliary-feed pump will respond to this signal by-. delivering enough feedwater to maintain the steam generator liquid level at 6 ft. Actuation of the ESTAS will override the steam generator level controller and the_ auxiliary feed pump will 26 l
deliver its maximum capacity.
A test calculation determined that the ESfAS would respond to high containment pressure (>4 psig).
lo accommodate this situation, the containment is represented as a single volume with a volume equivalent to that referenced in the IVA Bellefonte FSAR (see Reference 6).
This assumption represents an uncertainty to the calculation, since steam condensation on the containment wall might prolong actuation of the signal.
4.3 CALCULATION RESULTS Table 6 provides a summary of the sequence of events that occurred in the small break LOCA with loss of ac power transient.
1he initial response to the 9
cold lag LOCA with concurrent loss of ac power was the decrease in primary system pressure (figure 12) from 2210 psia to 1920 psia, lhis resulted from the primary coolant contraction due to the sudden reduction in thermal energy supplied by the reactor and the fluid lost out of the break.
Consequently, the hot leg temperature (figure 13) and pressurizer level (figure 14) also decreased. As a result of the closure of the main flow path through the turbine stop valves, secondary pressure (figure 12) increased.
lhis increased the saturation temperature on the secondary side, thus, increasing cold leg temperature in the primary system.
The rapid increase in steam generator pressure with the heat transfer mismatch from the drop in reactor power resulted in an increase in steam generator level (figure 15). Main feedwater pumps tripped from the built-in "cause and ef fect" assumption of low suction pressure from loss-of-condenser booster pumps-and thus, signaled the turbine-operated auxiliary feed pump to start after a delay of approximately 70 seconds. As a result of the coastdown of the main feed pumps and secondary pressure relief through the main steam safety valves (open at 1250 psia) and the modulating atmospheric dump valves (open at 1220 psia), the steam generator level dropped as the feedwater boiled.
Auxiliary feedwater supplied an equivalent flow (Figure 16) to maintain the steam generator level at I
approximately 6 f t as signaled from the Ecl systcm.
Following the initial primary system depressurization and hot leg temperature decrease, temperatures began to increase as a result of the
-coasting down of the reactor coolant pumps and the reduction in loop flow (Figure 17).
Pressure continued to decrease with loss of coolant from the 27
Table 6.
Transient sequence information for Scenario 2 Time 1sj_.
Event 0
Cold leg break opens, loss of offsite power results in reactor trip, reactor coolant pumps trip, charging and letdown system is isolated, pressurizer heaters trip, feedwater pumps trip, turbine stop valvo closes, steam driven auxiliary feedwater signal is generated 5
Steam generator modulating atmospheric relief valves and safety relief valves open 450 Hot leg high point begins to void 650 ESFAS is activated on high containment pressure 800 Steam generator narrow range is at full level 1050 Natural circulation is interrupted, primary pressure and temperature increases, voiding at the break 1450 Flow is induced in steam generators, primary pressure and temperature drop sharply 1800 Calculation is terminated f
28
. ~. -
l l
P flan. r p I>W
.....,4.,,
l 7
j' 3i.....
t D
- i.....
ts i.....
i..... l%
W g-
'""...............................i........
ri m. <..c)
Figure 12.
Scenario 2:
primary and secondary pressure.
- u.i t..
a c.i. t. g.
c j
.n.
c l
j
[j L.,..,%v=
c k,
.............i..i.........i.....n..
Time (sec)
Figure 13.
Scenario 2:
primary hot and cold leg temperature.
29
l i
?c
- i...
b Tim. (.ee)
Figure 14.
Scenario 2:
pressurizer level, i...
- t..,7 N.rrow kar g.
meumum.v.:
.0.l>
79 h
.I
, 1
..............................i....i,.......
Tim. (..c)
Figure 15.
Scenario 2:
steam generator narrow range levels.
30 e.e--d v-w-
.e,w-
=
m M
u--egr y
rs
+-
+ - -F-
> +-----
-g,r
-6..,.i
- i.., )
T...
}
\\ maurnu n eutu,ary gs.
s nu 5,.
a e h...
d X
i..
- a-
.................. i......... i.... i.... i s...
Tim.(..c) figure 16.
Scenario 2:
auxiliary feedwater flow.
)... p.
-t..,.
3.-. t..e u
mt
} i..
.c
&j ii.
I k
C w6... i.......4...
~~
=
t i.
is i...
Tim.t..c)
Figure 17.
Scenario 2:
primary hot leg ma:s flow rates.
I 31
i i
i cold leg.
The steam generator level control system stabilized heat transfer l
from the primary to the secondary by 200 seconds, allowing loop natural j
circulation flow to be achieved.
System pressure controlled the rate of the choked break flow. At 600 seconds, voiding began in the upper head (figure 18), restricting further RCS depresturization.
At 650 seconds, the containment pressure reached an ESFAS setpoint (>4 psig) and tripped the turbine-operated auxiliary feedwater pump to discharge maximum flow (1200 gpm). This resulted in a decrease in cold leg temperature as greater heat transfer was occurring in the steam generators.
At 1050 seconds, voiding at the hot leg high point (figure 19) interrupted natural circulation.
This event prevented adequate primary to secondary heat transfer, resulting in an increase in cold leg temperature from steam that discharged from the reactor vessel vent valves into the cold leg.
Since the cold leg is elevated above the exit of the steam generator and the inlet into the reactor vessel, lottlized voiding occurred near the break as this coolant drained from the sys'am.
This resulted in a decrease in break mass flow (figure 20).
With the 1
ic.,s of heat transfer to the secondary, primary pressure increased from coolant expansion from the core decay heat, which caused volumetric break flow to increase (figure 20).
At 1450 seconds, heat transfer to the secondary was sufficient to cool the standing coolant on the primary side of the steam generator so that the hydrostatic pressure forces of this column of liquid were greater thaq the downstream hydrostatic forces.
This force imbalance resulted in a slug of coolant to drain from the steam generator (figure 21).
I An increase in break flow resulted from the induced flow.
Voiding reestablished as the steam generator entrance interrupted loop flow again, allowing system pressure to increase as it hai done before.
At 1800 seconds, the calculation was terminated.
Continuation of this transient would eventually result in the termination of primary loop flow natural. circulation (caused by continued primary mass loss out-of the break), core boil off, fuel rod dryout, and subsequent l
cladding _ temperature excursions.
However, simulating this stage of the
-transient should be deferred until more information is obtained about the performance of the WNP1 simulator.
In conclusion, specific uncertainties characterized this calculation.
The primary uncertainty was the timing of the ESFAS signal, which RELAP5 predicted would result from high containment pressure.
Since the containment volume was not modeled to consider 32
mmm.
i
,..+
i.
p
- ..ei........1 o
t E
i..
;M.'..................i.............i.....
tim. <..o Figure 18.
Scenario 2:
reactor vessel upper head void fraction, i.
f**
l E
i..
1
..=.=...e..
Time (..c)
Figure 19.
Scenario 2 hot leg high point void fraction.
33
l
...u a
c,
-~
I
- - olum.tric Fl..
- v... n..
7 ""
ll
.... 7
.s.
j "
.... 1 i: ;
7 e
20-dr.a now \\
3
~,
e a
j '" "w_
i
.... j f
i
,i f. ~,; (.,,, j 1
............i...=............
..o Tim. (s.c)
Figure 20. Scenario 2:
break mass and volumetric flow rates, u
c.a...
t.
l e
o o
.*u
............... i. ;... i.;... i..........
Tim. (e ec)
Figure 21. Scenario 2:
representative steam generator void fraction.
34
containment wall condensation from environmental cooling, an ESFAS signal may l
have been generated early, Condensation would likely delay an ESFAS signal en high containment pressure.
Specific information or' containment cooling was not available to use in the model.
35
5.
SCENARIO 3:
STUCK-OPEN PRESSURIZER SAFETY VALVE The following section details the analysis of a stuck-open pressurizer safety valve simulation. The transient initiates with the reactor at full power.
The following subsections contain a description of the scenario, 3
calculational assumptions, and an analysis of the results.
5.1 SCENARIO DESCRIPTION The stuck-open pressurizer safety valve is characterized with the valve ope,ing spontaneously and remaining open.
The assumption about the mode of valve failure is to test the capabilities of the simulator rather than to model a probable f ailure.
The safety valve in this scenario is ramped open in one second, No operator intervention was modeled for this transient.
5.2 CALCULATION ASSUMPTICNS The basic WNP1 model described in Section 2 was used to perform this calculation. The failed-open pressurizer safety valve transient relies on the RPS and the ESFAS to mitigate the pr ential breach of system integrity a
introduced by this event. At initiation of this transient, the reactor is operating at full power.
The safety valve is ramped open in one second.
The 2
full open area is equal to 0.03 ft.
A subcooled and two-phase discharge co' fficient of 0.75 is necessary to approximate actual break flow as documented in the Bellefonte FSAR (see Referenue 6).
The RPS will signal a reactor trip when the pressure drops below 1981 psig.
This will signal a turbine trip on the secondary side. When pressure drops below 1600 psig, the ESFAS will actuate.
This signal will initiate the HPI (af ter a delay of 15 seconds), isolate the steam generators, and initiate auxiliary feedwater finw (after a delay of 15 seconds) at the maximum capacity (1200 gpm).
The model incorporates a flow table to simulate the HPI pump performance.
m 36
l l
5,3 CALCULA*(ION RESULTS Table 7 presents the sequence of events for this transient.
Ensuing from the failed-open pressurizer safety valve, which ramped open over the first second of the calculation, system pressure (Figure 22) rapidly decreased as steam escaped from the top of the pressurizer.
As vapor escaped from the pressurizer, liquid from-Loop A replaced the lost vapor, causing the pressurizer level to rise slowly (Figure 23).
Flow from the open safety valve (Figure 24) initially increased to nearly 155 lb/s, then decreased with the depressurization of the primary system.
At 15 seconds, the primary system pressure dropped below 1987 psig, which activated a reactor trip from the RPS.
A trip signal to the turbine followed, which resulted in closure of the turbine stop valves on the secondary side.
Primary pressure and pressurizer level rapidly decreased from the primary coolant contraction, With the reduction in core power, loop temperatures (Figure 25) also dropped sharply.
Secondary pressure increased to the modulating atmospheric dump valves and main steam safety valve setpoint of 1220 and 1250 psia, respectively, as a response to the closure of the turbine stop valves. -This resulted in an increase in the steam generator level (Figure 26) as main feedwater continued to enter the steam generators and vapor condensed from the increased pressure.
Steam generator PORVs opened at 14 and 15 seconds, respectively.
The increase in secondary pressure increased the saturation temperature of the steam in the steam generators;-this resulted in an initial increase in p_rimary cold leg temperature.
At 28 seconds, system pressure dropped below 1600 psig, which signaled the ESFAS.
The ESFAS responded by closing the main steam and main feed isolation valves in the steam generators.
This tripped the main feedwater pumps and actuated the auxiliary feedwater (Figure 27) and ECCS high pressure (Figure 24) pumps to supply coolant to the steam generatocs and the primary system (after a 15-second delay), respectively.
By 96 seconds, the ECCS had compensated for the lost coolant (i.e., pressurizer level was back to normal).
37
Table 7.
Transient sequence information for Scenario 3 Time 10_.
Event 0
Pressurizer safety valve begins to open 1
Pressurizer safety valve is full open 15 Reactor is tripped off on low pressure signal and turbine is tripped 28 ESFAS is actuated on low RCS pressure, steam generators are isolated, main feedwater pumps aia tripped, auxiliary feedwater and ECCS pumps are activated 43 Auxiliary feedwater and ECCS pumps are at full operation 180 Pressurizer normalized level is at 100%
300 Calculation is terminated f
38
W Prim a ry l W Seeendary]
......i' 2
3 s.....
t
=
- i.....
t s
l.....
1.....
i...
i...
Tim. (.ec)
Figure 22.
Scenario 3:
primary and secondary pressure.
=
=
=
=
\\ pressuriz.c top
^g ne..
t i.
3.....
i s...
Tim (eec)
Figure 23.
Scenario 3: pressurizer level.
l I
39
1 l
l
-.i, v.i..
o--e aec.
7 g__
}...
s
.j no.
g 3
- 5.....
. sow c
\\ tr.nsiuon to
- '"Y is..
i...
Time (..c)
Figure 24.
Scenario 3:
safety valve and total ECCS mass flow rates.
..u
- m.u..
- c.
g...."7 t
B.....
1 E
i
_ ~
i...
i...
Tim. (..c)
Figure 25.
Scenario 3:
primary hot and cold leg temperature.
40
- t.., a.
.---.t..,
.u
?:
3
... N is..
i........
Tim. (..c)
Figure 26.
Scenario 3:
steam generator narrow range levels.
eu
. - z....
o---o La ep 8 T
4
.u 3-
,Ec....
2 a.u oc G
l..
i.'...
is..
..o......
u...
tim. (..e>
Figure 27.
Scenario 3:
audliary feedwater flow rate.
41
Pressurizer level, responding to the open safety valve and increased system mass from ECCS injection, increased after a sharp decrease from primary-coolant contraction following the reactor trip.
Near 130 seconds into the -
calculation,- flow from the open safety valve began to transition from a two-phase liquid / vapor mixture to release only liquid coolant from the pressurizer and by 180 seconds, the pressurizer was completely full.
With the filling of the pressurizer, flow from the open valve had increasea; however, the ECCS could still compensate.
Throughout the calculation, coolant-entering the pressurizer and leaving through the safety valve and flow from the ECCS directly influenced the variations of the flow in the coolant loops l
(Figure 28). At 300 seconds, the transient was terminated when it was determineo that all major events had occurred.
In the absence of meaningful uncertainties for this hypothetical transient, the results presented represent the best estimate prediction of RELAP5. The failed-open pressurizer safety valve transient also represents a benign event (i.e., the core is not in danger) that is adequately handled by the built _in safety systems.
5 42
-1
tp a4.o c
2 i., 4 3 t..., a E
4 c.
E
$ to.o 5
t a:
3
,o se e a
a r
se.o ite 40.0 30.0 180.0 180 0 300.0 840.0 g
-Figure 28.
Scenario 3.
primary hot leg mass flow rates.
I i
i 43
i 6.
SCENARIO 4:
MAIN STEAMLINE BREAK WITH STEAM GENERATOR TUBE RUPTURE The following section details the analysis of a double-ended guillotine rupture of a WNPI main steamline upstream of the main steam isolation valves (MSIVs), with the concurrent rupture of a single steam generator tube.
The transient initiates with the reactor at full power.
The following subsections contain a description of the scenario, calculational assumptions, and an analysis of the results.
6.1 SCENARIO DESCRIPTION The main steamline break is the instantaneous nonisolatable, double-ended guillotine rupture of a main steamline upstream of the MSIV with the reactor at full power.
Concurrently, a steam generator tube rupture (SGTR) occurs in the steam generator adjacent to the break.
The location of the tube rupture is at the bottom of the tube sheet adjacent to the primary outlet plenum.
No operator action occurs during the course of this ennt.
6.2 CALCULATION ASSUMPTIONS The double-ended guillotine break in the main steamline and SGTR are modeled to occur simultanecusly on the secondary side of Loop A.
The system will respond with a signal from the ESFAS on low secondary pressure (pressure taps are modeled immediately downstream of the steam generators).
This signal will trip the reactor, which will generate a turbine trip signal, initiate the ECCS, and close the main feed and steam isolation valves.
The ESFAS will also initiate auxiliary feed, which will provide feedwater based on the F0GG logic.
For the steamline break, modification of the basic WNP1 model described in Section 2 was necessary to provide greater detail in this region (Figure 29).
The original input model combined the two steamlines coming of f the steam generator.
Renodalization of the steamline provided for both of these lines, and for the break in one line.
The break model involves the insertion of two valve components attached to Volumes 321 and 330.
At 44
812 Atmosphere 811 Safety Valves 315 Boundary Condition l
l 2
320 10 3 1
810
/
/
9
'/
2 l
/
Modulating 809 9 I
Atmospheric fi g
'l j
3 Valves e
7
$ )
4 341 I
h MSIV 6
j
/
5 Steamline 336 2
? l 5
s, j 6
f, ;';
340 4
7 3
335 MSIV j
3 3
- s ',
8 i
793 330 2
/
9 Y' -
351 350 79~,
u2 7'
331 310 795 326 I
?
10 f
f, 794 4
7, p
Break Model
- )
I f,
Steamline 2
'e I 11 360 Steam lieader I I 951 300 311 I h
V 365 Turbine Stop 1
1 i
'$ j 12 A
Valve a
z Feedwater I
I Boundary 370 Condition SGTR Stearn Boundary Model M Condition Figure 29.
Secondary renodalization for steamline break.
45
l initiation of the transient, these valve components trip open.
A third valve, connecting Volumes 321 and 330, closes at that time, thus, isolating the steam header from the affected steam generator.
Communication between the two l
l secondary loops through the common header that exists upstream of the turbines was not incorporated since geometric information about this portion of the systam was not available.
This represents a prime uncertainty in this 4
calculation.
Simulation of the tube rupture in the A steam generator also required modifications to the model.
The original steam generator model simulated the primary tube assembly as a single pipe volume.
For the SGTR event, a single tube was modeled from the primary inlet plenum to the outlet plenum.
At the outlet plenum junction, a valve component connects the single tube to the inlet of the secondary tube shell.
Likewise, a valve component connects the primary outlet plenum to the inlet of the secondary tube shell.
These valves open at the beginning of the transient.
This configuration of the SGTR model provides the correct break flow as documented in the Bellefonte FSAR (see Reference 6).
Control systems will operate in their automatic modes.
The F0GG logic will ensure auxiliary feed to a steam generator until its pressure decreases below 600 psig (if both steam generator have pressures less than 600 psig, auxiliary feed will resume to the dry stsam generator).
6.3 CALCULATION RESULTS Table 8 presents the sequence of events for this transient.
The immediate result of the steamline break was the rapid decrease in pressure in the A steam generator.
Within one second, an ESFAS signal actuated from a signal of low secondary pressure in the A steam generator (Figure 30).
This resulted in a reactor trip, which signaled a turbine trip.
The main steam and feedwater isolation valves began to close.and the ECCS and auxiliarv feed pumps received the signal to begin pumping, in the B steam generator, l
pressure increased to the safety valve setpoint following closure of the main steam flow path (turbine).
Saturation temperature increased in the B steam generator, causing an increase in the cold lea.emperature in the primary B loop (figure 31).
Flow from the break quickly reached sonic speed and choked.
46
m Table-8.- Transient sequence information-for Scenario 4 Time isI-Event 0
Main steamline loop A and one steam generator tube rupture 0.05 Reactor scrams on ESFAS signal, triggered on low steam generator pressure; turbine is tripped; steam generators are isolated; auxiliary feedwater and ECCS pumps are signaled for operation i
5-MSIVs are completely closed 15 Auxiliary feedwater and ECCS pumps are at full operation 36 ECCS has made up for coolant lost through the steam generator 120 Calculation is terminated 47
W 3 G &.
i...
2 io...
=
b C
-...'i... u.. " ii..
Time (s ee)
Figure 30. Scenario 4:
steam generator pressure.
M Het Leg l C
3 Cold Let!
g.....
t.
f.....
E E# ee.
.,' s
.................. iii. i...
.. i.....
Time (esc) l Figure 31. Scenario 4:
Loop B hot and cold leg temperature.
48
The rapid exhaust of steam from the A steam generator temporarily increased heat transfer in that steam generator.
This resulted in a significant decrease in coolant temperatures exiting from the primary outlet plenum of the A steam generator (Figure 32). As the pressure in the A steam generator decreased, coolant flashed and boiled on the secondary side resulting in a rapid decrease in the liquid level (Figure 33).
Excessive cooling in the A steam generator beyond the normal heat transfer rate resulted in a reverse in heat transfer in the B steam generator (fluid temperatures in the secondary system exceeded primary coolant temperature, thus heating the primary system),
as shown in Figure 31.
Blowdown in the A steam generator vias terminated at 46 seconds and the break flow unchoked.
Concurrently, che steaaline pressure stabilized at 20 psia.
Primary pressure decreased sharply as secondary coolant escaped through the break (thus, overcooling the primary system) and from primary coolant contraction following the reactor trip (Figure 34).
These two components dominated the pressure response, hiding any effect of the SGTR.
At 15 seconds into the transient, the ECCS HPI had reached full operation and began compensating for the lost coolant (Figure 35 and 36).
Auxiliary feedwater also began injection to the steam generators.
However, the conditions in the A steam generator signaled the F0GG system to bypass feedwater to that steam generator, while allowing feedwater to the B steam generator.
The auxiliary feedwater momentarily reduced the liquid volume in the B steam generator as the cold auxiliary feedwater mixed with hot boiler water.
The mixing resulted in a denser liquid, thus reducing the secondary liquid volume (Figure 33).
By 40 seconds, the ECCS had fuit.y compensated for the lost primary coolant. As the HPl continued to supply emergency coolant, primary pressure increased.
Concurrently, the mass flow rate from the broken stea.iline had decreased to only that from the ruptured steam generator tube (Figure 37). At 120 seconds, the transient was terminated when it was decided that the major events in the transient had occurred.
In conclusion, the general phenomena exhibited in the simulation of the steamline break with concurrent SGTR have little dependence on the uncertainties attiibuted from the calculation assumptions.
Since no communication between the two secondary loops was modeled, it is possible that 49
.... e - -r-- :
.---e u.i s.~al
- W Cold Leg)
..u E
i L.se e
=
=
f
~
in su
.u
.u
,u
.. i....
n.
i..
Time (s ec)
Figure 32.
Scenario 4:
Loop A hot and cold leg temperature.
in.
W SS 4 o---- e : o i s...
2
.. e a
?
i c.
Auzmary feedwater
/ at mexunum
\\.
n..
. _.- we in n..
en u..
.u n.
.o i.u un in.
Time (s ee)
Figure 33. Scenario 4:
steam generator narrow range levels, s
50
sou g.....
a i....
i.....
... i....
n.
............ i... ii.. i...
Time (.ec)
Figure 34.
Scenario 4:
pressurizer pressure, n..
i...
I
} i...
.c 5m 2
mj
-m i.....
n.
.................. ii.....
Tim.(.u)
Figure 35.
Scenario 4: -ECCS mass flow rates.
51 i
. -.. ~ -
_. 5,.u..
l k...
Y a
ej... p,_ ;
~
i..
1 n..
.o t.
i... in. i..
Time (..c)
Figure 36.
Scenario 4:
steam generator tube rupture mass flow rates.
t
(......;
E i
e;.....
1n......
Oc:
i s....
o-
.u u.
.1.
i...
n..
un a
n.
.u Tim. (s.c)
-Figure 37.
Scenario 4:
steamline break mass flow rate.
52
I the intact secondary loop might be temporarily affected by the steamline
- break, However, since the ESFAS responded so quickly to the event, isolation of the broken loop steamline would also occur quickly.
Timing of the ESFAS represents another important factor defining behavior of the transient.
Ilowever, since the A steam generator blows down so quickly, timing error should be insignificant. Overall, the safety and control systems of the WNP1 facility adequately handled the steamline break and SGlR transient.
53
E l
7.
SCENARIO 5:
LOSS OF MAIN FEEDWATER WITH DELAYED SCRAM The following section details the analysis of a loss-of-main-feedwater accident with a delayed reactor scram.
The transient occurs with the reactor at full power.
The following subsections contain a description of the scenario, calculational assumptions, and an analysis of the results.
7,1 SCENARIO DESCRIPTION This transient is the complete loss of main feedwater with an additional malfunction that prevents the automatic scram signal from tripping the reactor.
The loss of-main-feedwater transient is the simultaneous loss of power to all pumps within the secondary feedwater train. A RELAP5 time-dependent trip to scram the reactor simulates the only operator action for this transient.
7,2 CALCULATION ASSUMPTIONS The basic WNP1 model described in Section 2, with the addition of a kinetics model, was used to perform this calculation.
In contrast to the pre /ious calculations, the loss of main feedwater without automatic scram required a kinetics model to ensure that after the loss of heat sink, the increase in moderator and fuel temperatures and the decrease in moderator density would result in a correctly simulated power result.
This kinetics model incorporates the reactivity coefficients representing beginning of life system conditions.
The reactivity coefficients constitute an uncertainty in l
this scenario.
Failure to scram is the result of a lockup of the rod speed j
controller, so that the control rods contribute zero reactivity in the RELAP5 L
simulation until the manual scram.
This also precludes any rod movement contribution by the ICS to mitigate this event.
The manual scram represents the only operator intervention action simulated in this calculation.
All other modeled automatic control systems will work correctly before and af ter generation of the automatic scram.
The pump coastdown approximation for the main feedwater cumps is another uncertainty affecting the timing of events in this transient.
'he assumption is that flow from the main feedwater pumps 54
completes the coastdown with;n 8 seconds following the loss of feedwater train.
This assumption was necessary in the absence of sufficient information, 7.3 CALCULATION RESULTS Table 9 presents the sequence of events for this transient.
Following the initiating events (loss of feedwater train), auxiliary feedwater pumps responded to the EC1 signal from a low flow condition in the secondary feedwater train.
At 12.5 seconds, the RPS signaled for a reactor trip on high system pressure (Figure 38), which failed to actuate; however, a turbine trip did successfully respond to this signal.
Following the turoine trip, secondary pressure increased toward the modulating atmospheric dump valve setpoint of 1220 psia.
Concurrently, primary pressere began to rapidly increase as heat transfer to the secondary system decayed with the loss of feedwater, lhe pressurizer relief and safety valves opened on high pressure at 10 and 15 seconds (Figure 39), respectively.
This affected mass flow in the loops by increasing loop flow in loop A (pressurizer loop) and decreasing loop flow in loop B as the pressurizer siphoned coolant from the loops (Figure 40).
In the steam generators, the auxiliary feedwater system was ineffective in controlling the steam generator level.
Since the auxiliary feedwater system can handle only decay heat power levels, most of the original feedwater had boiled off by 15 seconds (Figure 41),
This created a power mismatch between the primary and secondary (Figure 42), resulting in a sharp increase in temperature and pressure as core energy continued to heat the primary with reduced heat transfer to the secondary.
At 20 seconds, conditions in the primary had nearly reached the critical point.
Concurrently, reactor power began to drop significantly (Figure 43) as moderator and fuel temperature increased and moderator density decreased, which all contributed negative reactivity to the nuclear core.
Power reduced to nearly 85% of nominal before the manual reactor scram.
Since RELAP5 was having severe performance problems with the calculation, it was necessary to initiate a manual scram to continue the calculation.
Primary pressure and temperature peaked at 21 seconds, reaching a maximum of 3120 psia and 656'F, 55
Table 9.
Transient sequence information for Scenario 5 Time (s)
Event 0
Power lost to feedwater train and auxiliary feedwater signal is generated 10 Pressurizer PORV opens on high pressure 12,5 Reactor fails to scram on high system pressure 15 Pressurizer safety valve opens on high pressure and steam generators boil dry 20 Manual scram is initiated 21 Reactor pressure peaks at 3120 psia 28 Pressurizer is completely solid 100 Calculation is terminated (primary system is cooling down) 56
l l
- Pete.rF W leeendart
...6 T
ej 4....
t D
4 i.....
i,...
..................,......... i....
Time (. 0 Figure 38.
Scenario 5:
primary and secondary pressure, W....,
W.sitef 7.....
Eac 1g....
i
- i...
m T-.......
. ~...
. i.......... i....
i..
~
~
~
Time (sec)
Figure 39.
Scenario 5:
pressurizer safety and relief valve mass flow rates.
57
.=..
.--.o..,.
--.t..,.
n..
b) o
.o i
c.
j u..,
eo c....
,:m i...
Time (sec)
Figure 40.
Scenario 5:
primary hot leg mass flow rates.
W S0.
C
- SC.
C.,
e d
.Q..
I i
49..
fn.Iitnufn.u11h.fy l
l i aw.t r nu l
v.'.... '..
Time (see)
Figure 41.
Scenario 5:
steam generator narrow range levels.
i 58 J
W Cet. Le Peta.ory W Primary to Seeendery r.....
E t
Ba......
n=n
=
=
p-
.. i.................,......... i...
nme (sec)
Figure 42. Scenario 5:
system heat transfer at steam generator.
r.....
5:
ae....-.
io....
...r
... i.....
........ i...
Time (sec)
Figure 43. Scenario 5:
reactor power.
59
respectively.
The system began to depressurize following the reactor scram, as the auxiliary feedwater flow in the steam generators removed core energy and the open pressurizer relief valves released coolant to the containment.
The steam generator level rose slowly following the reactor trip, because the energy stored in the primary system was still great enough to boil the auxiliary feedwater entering the steam generators.
The pressurizer responded slowly to the reactor trip and its level continued to increase until 28 seconds, when the pressurizer was completely solid (Figure 44).
As the pressure continued to decrease, the pressurizer safety and relief valves closed at 30 and 36 seconds, respectively. Cold leg temperature was temporarily greater than the hot leg temperature as a result of the time delay of fluid traveling from the hot to the cold leg (Figure 45).
At 100 seconds, the transient was terminated when it was determined that all relevant transient events had occurred.
In conclusion, the amount of coolant in the steam generators prior to the transient had the greatest influence on this anticipated transient without scram.
Once this coolant had boiled off, little heat transfer could take place between the primary and secondary systems, resulting in the pressure and temperature excursion in the system.
The timing of when boil-off in the steam generators had reached its maximum is dependent on the initial liquid quantity and the coastdown of the feedwater pumps. The assumption that these pumps coast down within eight seconds was necessary in the absence of specific information. However, feedwater coming into the steam generators following a pump trip should only represent a small fraction of the tot 61 liquid already in the steam generators; therefore, inlet feedwater will not delay the loss of heat transfer much beyond the time it takes to boil of f the initial steam generator inventory.
Since this transient is driven by the initial steam generator coolant inventory, uncertainties about the kinetic response of the core do not significantly affect the timing of the transient.
60
w,_
i 1
i i
i
+
=
3....
I..
u
.... J. d.... J.. '..... i....
Tim e (sec)
Figure 44.
Scenario 5: pressurizer level.
-..u.. I W f ate Leg l g....
e 3
Ea....
i
- a. - k.... a. J..... A....
Time (sec)
Figure 45.
Scenario 5:- primary hot and cold leg temperatures.
.i 1
61
8, CONCLUSIONS AND RECOMMENDATIONS Analyse-five WNP1 accident scenarios were performed with the RELApS computer code.
ine basis of these calculations derives f rom the need to benchmark the 11C WNP1 simulatcr with RELAPS to determine how well the simulator can model a wide range of accidents and to identify where simulator software technology and system code capability can improve.
Computationai information presented in this report is a sample of a much more detailed data base calculated by the REL APS code.
Additiona' data for these simulations are stored on magnetic tape and maintained at the lhEL.
These data will be used for future TTC simulator /RELAP5 benchmark comparisons.
In general, the predictions of the P. FLAPS were reasonable for the scenario 3 in this report and will provide a valid basis for comparison with simulator data.
This conclusion was based on extensive review of the scenario data by experienced reactor operators and systems analysts at the INEL.
Vncertainties relative to boundary conditions, delay times, and instrumentation orocess deley times still have not been resolved and could potentially affect simulator /REL AP5 result s.
Although comparison with simulator data will primarily be used to assess simulater performance, it will also be used to assess the code's capability to model the more mechanistic phenomena of plant behavior (e.g., boundary conditions sucS as feedwater flow, ccre power, etc., that the
- mulator may do better).
1 62
9.
REFEREN(t.S 1.
Report of the President's Ccamission 00 the Accident at Threc Mile Island, Washington, 0.0., October 1979, 2.
C. H. A311 son et al., SCDAP/R[ LAP 5/M003 Code Manual, NUREG/CH-6536, fGG-2596, June 1990.
3.
P. D. Bay 1ess et al., feedwater Transient and Small Break loss of Coolant Accident Analyses for the Bellefonte Nuclear Plant, NUREG/CR 474).
EGG 2471, March 1987.
4.
American flational Standard for Decay Power in Light Water Reactors, ANSI /ANS-5.1, 1979.
5.
kNP-1 Operating System Description, Washington Public Power Supply System, January 1979.
6.
Tennessee Valley Authority, Bellefonte Nuclear Plant, final Jafety Analysis Report, Docket 50 438, December 1982.
I 63
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too 2633
, c. v -, 4.. u t,in.a.s w ia RELAP5 %ennal-llydraAic Analysis of the WiiP1 Precurized Wales Heat tot 3
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Idaho National Engineering Laturatory EG&O idaho. Inc.
P.O. Box 1625 Idaho Falls,laaho 83415
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Division of Systems Researth Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington.D.C.20555 10 SutPLE ME Nian t NCT EL 11 Ab t f R Ali sm.e,o.
Thermal-hydraulic analysen of five hypothetical accident ccenarion were perf ormed with the RI:1APL/ MOD 3 computer code for the Babcock and Wilcox Company Washington Fuclear Project Unit 1 (WNP1) pressurized water renetor.
This work vaa uponsored by the U.S.
Nuclear Regulatory Commission (HRC) and in being performed in conjunction with future analysis work at the NHC Technical Training Center in Chattanooga, Tennoscoe.
The accident scenarios were chosen to acceso and benchmark the thermal-hydraulic capabilitien of the Technical Training Center WNP1 almulator to model abnormal transieni er id it ioris.
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RELAP5/ MOD 3 DUEEified WNPI thermal-hydraulic analysis DUEGfied hypotheticalaccident scenarios is so.ut a o raon 10 iH1'yL M C P OW W )Jb (2 8 9,
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