ML20045B706
| ML20045B706 | |
| Person / Time | |
|---|---|
| Site: | Black Fox |
| Issue date: | 05/31/1993 |
| From: | Martin R EG&G IDAHO, INC. |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-L-1050 EGG-2677, NUREG-CR-5882, NUDOCS 9306180322 | |
| Download: ML20045B706 (68) | |
Text
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l NUREG/CR-5882 EGG-2677 TRAC 3 Taerma:-Hycrau:ic ;
Ana:ysis of the B.acic Fox l Boi:ing Wa~er Reac":or l l
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. I'repared by R. P. mnin j j
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'l Idaho National Engineering Laboratory I EG&G Idaho, Inc. ,
1 Preparul for U.S. Nuclear llegulatory Commission ,
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. i 9306180322 930531 [
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DISCLAIMER NOTICE j t
This report was prepared as an account of work sponsored by an agency of the Unitod States Govemment. g f Jetther the United Stames Govemment nor any agencythercof, or any o'ther employees, makes any warranty, expressed or imphed, cr assumes any !e;;al liability of responsbilty for any third party's use, or the resu!ts of I such use, of any inforrnat:On, apparatus, product or proCOCS disclosed in this report, or represents that its use s
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EGG-2677 R4- !
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i TRAC-B Thermal-Hydraulic l Analysis of the Black Fox :
Boiling Water Reactor :
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i Manuscript Completed: March 1993 ;
) Date Published: May 1993 <
b Prepared by i R. P. Martin ;
l Idaho National Engineering 12tboratory {
Managed by the U.S. Department of Energy [
EG&G Idaho, Inc. j Idaho Falls,ID 83415 i
- Prepared for !
Division of Systems Research j Office of Nuclear Regulatory Research !
U.S. Nuclear Regulatory Commission l
Washington, DC 20555 '
NRC FIN L1050 ,
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t ABSTRACT ;
Thermal-hydraulic analyses of six hypothetical accident scenarios for the >
General Electric Black Fox Nuclear Project boiling water reactor were perfonned >
using the TRAC-BF) computer code. This work is sponsored by the U.S. Nuclear i Regulatory Commission and is being done in conjunction with future analysis work at the U.S. Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and ;
benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient ;
conditions.
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FIN > o. L1050-l hermal-hydraulic analyses using TRAC-BFI computer code.
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CONTENTS ABSTRACT....................................................................................................................... iii EX EC UTI V E S U M M A R Y . . . . . . . . . . . . . . . . . . . ... . . . . .. .. . . . . . .. . .. . . . . . . . . .. . . . . . . ... . .. . . . . . . . . . . . . .. . . . . xi . . . . . . . . . . . . .. . .. . .
A C KN O WLE DG M ENTS . . . . . . . . .. .. . . .. . . . .. . . ... . . .. . .. . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... .. . . . . . . . .. .
ACRONYMS.................................................................................................................... xv
- 1. INTRODUCTION........................................................................................................... ,
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- 2. M O D EL D E S C RI PTI O N . . ... . . . . . . . . . . . . . .. . . . . . .. .. .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 . . . . . . .. . . ... . . . . . .
2.1 The rm al Hydraulle M odel ... .. . .. . .. .. . . .. .. ... .... .... . .. .. . . .... ..... .. ..... ...... . . . .. .... ....... . ... ... .. ... 2 2.2 Control S y st em M od el . .. ... . .. ... .. .. . .. . .. .. . .. .. .... . . . . . ... .. . .. ... .. ....... . ... . . ... ... .. ...... . . .. .. .. .... .. 5 2.3 S te ady S tate Conditi ons . . . . .. .. . . .. .. ... .. . ... .. . .. ... . . . . . ... . . . . . .. . .. .. .. . .. .. .. . ... .. . ... .... ... . . ... ... 7
- 3. SCEN ARIO 1: INA DVERTENT MSIV CLOS URE ....... .... .... ..... .. ... ............... . .......... .. . ... 8 3.1 S c enario De s cription ... .. . .. .. . . .... . . .. . . . .. . . . .. . . .. ...... ... . ... .. .. . . . .. ..... . .. .. .. . . .. .. .. . .. . ..... ... 8 3.2 Calc ul at i on A ss um ptions . .. .. . .. . .. .. . . . . . .. . . ... . . . .. . . . ... ..... . ... .. .. ..... .... . . . .. . . .. . . . . . .. .. . . . ..... 8 3.3 Calculation R e s ult s . .. ... . ... . . .. . ... . .. . ... . . .. . . . .. .. .. . . . . . . .. .... . . . ... . ... .... ..... ... .. ... .. ....... ........ .. .. 8
- 4. S CEN ARIO 2: IN ADVERTENT ADS ACTU ATION ...... .... ..................... ............. ........... ... .... 14 4.1 S e e n ario De s cri ption ... . .. . . .. . . .. . .. ... . . . . . . .. .... . . . . . .. . .. . . .. . . ... . ... . . . . . .. . .. ... .. .. .. . .. ... . ..... . . .. ... ..... . 14 4.2 Caleulation Assumptions... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 4.3 Cal c ulation R e s ult s ... . . . . . . . . . . .... . . . . . ... ... . . .. . . . .. ... . ... . .. .. . .. .. . . . .. . . . . . . . ... .. . . ......... .. . . ....... 14
- 5. SCENARIO 3: LARGE-BREAK LOCA IN RECIRCULATION LOOP ...... ......... ...... .. ......... .. 20 5.1 S cenario Descri ption ........... .......... ... ... ........ ........... .. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.2 Calculation Assumptions.. . ................................................................... 20 5.3 Calculation Re s ult s ............ ...... ................ . . . .. .. ............................................... 20
- 6. SCENARIO 4: SM ALL BREAK LOCA, WITH STATION B LACKOUT......... ...... ... ............... 28 6.1 S c e n ari o De s c ription . . . .. .. . . . . . . . . . . . . .. . . . . .. . . . . . . .. . . . .. . . . . . . .. .. . .. ... . .. . . .. . .. . . . ... ... ...... . .. ... . 28 L
6.2 Cal culat io n A s s um pt ion s . .. . .. .. .... .. .. . . . . . .. . .. . . . .. . . . . . . . . . . .. . . . . . .. . .. .. . . . . . .. . .. . .. . .. .... . . . . .. . 28 6.3 Calculat ion Res ult s . .. . . . .. ..... . . . .... ... . .... ....... . ...... . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 i
d'
- 7. SCENARIO 5: TURBINE TRIP WITilOUT SCRAM.... ... - . . . . . . . . . . . . . . . . . . . . 34 7.1 Scenario Description.. . . .. . . .. . . .. .. ... - 34 7.2 Calculation Assumptions. . . .. . . . . . .. .. . . . . . . . . . . . . . . . . . . . . 34 l i
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' 7.3 Calculation Results . . . . . . . . .. . . . . . . . . . .. . . . 34 :
- 8. SCENARIO 6: 40% POWER / 60% FLOW TURBINE TRIP ATWS . ....... . .... .. . . . 41 8.1 Scenario Description. . . . . . . . . .. . . . . . . . . 41 8.2 Calculation Assumptions.. . . . . . . . . . . . . . . . . . .. . 41 i 8.3 Calculation Results. . . .. . . . .. . .. 41
- 9. CONCLUSIONS AND RECOMMENDATIONS. . . . . .. .. . . . . . 47 -l ;
- 10. REFERENCES . . .. . . . . .. . .. . .. . . . . . 48 l
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i LIST OF FIGURES !
- 1. Black Fox vessel nodalization. . . _ . .. 3 l 2. Scenario 1: reactor power. . . 10
{
- 3. Scenario 1: system pressure. . .. 10 ,
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1 i l 4. Scenario 1: wide-range water level . . . . . .. . .. I1 1
1
- 5. Scenario 1:' core Dow . .. . .. I1 ;
i
- 6. Scenario 1: reactor temperatures. . . 12
- 7. Scenario 1: main steam Dow upstream of SRVs . . . . . 12 l
- 8. Scenario 1: feedwater flow. .. . .. . . . 13
- 9. Scenario 1: ECCS flow. . . .. .. . 13 ,
- 10. Scenario 2: system pressure. . .. 16 f l
1 j 11. Scenario 2: reactor power. 16
- 12. Scenario 2: wide-range water level.. . . . . 17 l
- 13. Scenario 2: core flow.. . .. . 17 ;
i
- 14. Scenario 2: reactor temperatures. . . . 18 ,
- 15. Scenario 2: core inlet void fractio. .. . . . . 18 ;
- 16. Scenario 2: SRV Cow. . . . . . 19 l
- 17. Scenario 2: ECCSHow. . ., .. . . . . j9
- 18. Scenario 3: break flow. .
m I 1
- 19. Scenario 3: system pressure. . . 22
- 20. Scenario 3: wide-rance water level..
. . . . 23 e
- 21. Scenario 3: reactor power. . .. 23 ,
- 22. Scenario 3: main steam and SRV Dow. . .. . . 24
- 23. Scenario 3: lower plenum mass. .. . . 24
- 24. Scenario 3: ECCS flow. .. . . 25 P
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- 25. Scenario 3: total break How. . . 25 1
- 26. Scenario 3: core flow. 26 vii NUREG/CR-5882
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- 27. Scenario 3 : intact loop flow. .. ...... .................... ...... . .... .. ......... . . ...... .... . .. ...... ........... ........ . ....... 26 l 2 8. Scenario 3 : reactor temperatures. .. ..... .... ...... . . . ... ............................. . ........ ...... . ............ .... 27 l 2 9. S c enario 4 : reactor power. ........ .... . . ... ... . ... ..... . .. .............. ...... ... ......... ... .... .......... . .. . ...... .. ... ..... 30
-l 3
- 30. Scenario 4: core now. .. .. .. ..... .. ... . . ......................................................................... 30 i
- 31. Scenario 4: wide-range water level..... .................................................................. 31 j F
- 32. Scenario 4: s y s te m p re s sure . . . . . . .. . . .. . . . . . .. . .. . . . . . ... . .. . . .. . .. . .. .. .. . . . . . . . . . .. . . . ... . . . . . . .... .. . . . . . 31 !
- 33. Scenario 4: fraction o f S R V s ope n. . . .. . .. .... ... ............ . .. .... . ..... . . .... .. .. .... ..... . 32 ,
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- 34. Scenario 4: ECCS fl ow. .. ..... ...... ... .. . . .... . . . . . . . . . . . . . . . ... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 4
- 35. Scenario 4: bre ak flow. ... . ..... . ...... . .. .. . .... . . .. ... . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 )
4
- 36. Scenario 4: reactor temperaturer .. .. . . . . . . . .. . . . . . . . . . 33 !
(
- 37. Scenario 5: system pressure. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . 36 [
- 38. Scenario 5: wide-range water level.. .. . .. . . . . . . . . . . .. 36 i
- 39. Scenario 5: core flow. . .. .. . . ~ . . . . . . . . . . . . . . . . . . . 37 ~ I I
- 40. Scenario 5: reactor power. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 37 ;
- 41. Scenario 5: main steam and SRV Dow. . . .. .. . . . . . . . _ . . . . . . . . .. .. .. 38 !
- 42. Scenario 5: reactor temperatures. .. ... .. . . . .. .. .. . . . 38 .
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- 43. Scenario.5: ECCS Dow., .. . . . . .. . . . . .. . . . . . . . 39 :
3
- 44. Scenario 5: core inlet void fraction... . . . . . . . . . . . . . . . .. . . . . . . . . . . 39 i
- 45. Scenario 6: system pressure.. . .. . . . . - . . . . . . . . . . . . .. . . . . . 43 !
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- 46. Scenario 6: reactor power. . . ..... .. . . . . . . . _ . . . . . . . ... 43 :
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- 47. Scenario 6: main steam and SRV flow. . . ... . . . . . . . . 44 l
- 48. Scenario 6: core flow. . . . . . . . .. . . . . . . - . . . . . . . . . 44
- 49. Scenario 6: reactor temperatures. . .. . . .. .. . . . .. 45 ;
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- 50. Scenario 6: wide-range water les el.. . . . . . . . .. .. . 45 i
- 51. Scenario 6: ECCS flow. . . . . . .. . . .. . . .. .. .. 46 j t
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NUREG/CR-5882 viii f
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i LIST OF TABLES
- 1. S u mmary o f sc en arios an alyzed. ...... ......... ...... . ....... .. ..... ... . ... ......... .. ... ....... ....... ........... . ....... 1
- 2. Correspondence between the physical and mathematical components in the TRAC B model of BFNP.. ..... . . .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .................... 4
- 3. Comparison of the TRAC-B and simulator initial conditions at full and reduced power ........... 5
- 4. Transient sequence information for Scenario 1.... .. ..... . . . .. . . . . . . . . . . . . . . . . . . . . 9
- 5. Transient sequence information for Scenario 2 . .. ....... ..... . .... .. . .. ...... ...... .. . .... ... . .... . 15
- 6. Transient sequence information for Scenario 3 . ..... .. . .... .. . . .......... . .......... .. ...... .. ........... 21
- 7. Transient sequence information for Scenario 4. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
- 8. Transient sequence information for Scenario 5 . .. .. . .. . .. .. . . . . .... .... . ... ... ..... . .. .... 35
- 9. Transient sequence information for Scenario 6. .. . . 42 i
= - . -.
EXECUTIVE
SUMMARY
l In 1979 the U.S. Nuclear Regulatory Commission (NRC) adopted recommendations from the Kemeny Commission requiring that all nuclear power plants have a plant specific simulator to use in l training operators. The simulator is required to be able to model plant operation and transients in an i environment closely resembling the plant control room. Today's simulators have evolved into reliable tools for simulating normal plant operational transients. Howes er, for many other types of transients and accidents. the current simulators base not yet been shown to produce reliable results or are unable to model the plant response altogether. Keeping to the recommendations of the Kemeny Commission, the NRC began a project that examines the capabilities of the current generation of simulators. The focus of this initiative is to evaluate simulator capabilities under uniquely challenging transient scenarios, measured against the predictions of advanced thermal-hydraulic system codes such as RELAP5 and ,
TRAC-B. Based on previous work. this measurement is expected to be in both directions (i.e., both ,
i system code and simulator will benefit by the comparison).
The simulator for the General Electric (GE) Black Fox Nuclear Project (BFNP) is located at the !
NRC Technical Training Center (TIC) and was modeled using TRAC-BFl. The two-loop boiling water reactor (BWR) model contains detailed thermal-hydraulic representations of the pertinent BWR systems, including the emergency core cooling system (ECCS), recirculation loops, and steamlines. The model also includes detailed models of key plant control systems. ;
i The TR AC-H model was used to analyze six separate transients, which had been selected to cover a }
wide range of possible thermal-hydraulic conditions that could occur in a reactor accident. The transients were:- (a) large-break loss-of-coolant-accident (LOCA). (b) small-break LOCA with a station blackout, (c) inadvertent closure of the main steam isolation salve, (d) inadvertent initiation of automatic depressurization system from 40% power /60% flow system conditions, (e) turbine trip without scram with the plant at full power, and (f) turbine trip without scram with the plant at 409 power /60% flow .
system conditions. Except w here noted, these transients w cre initiated from full power.
In general, the calculated TRAC-B trends were reasonable for the scenarios studied in the analysis, and will. judging by the review of experienced operators and plant analysts, provide a good basis for comparison with simulator data. Choices of boundary conditions and modeling assumptions used in these i analy ses might influence the results; however, such assumptions have been made consistent with the !
Black Fox simulator when possible.This practice will be considered in the final comparisons of simulator ;
and TRAC B results for these transients. e t
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ACKNOWLEDGMENTS The author gratefully acknowledges the assistance of Steve Roessler, Jan Griffin, and Jack Lewis at the NRC TTC for their advice and direction in formulating the accident scenarios in this document. In addition, the author would like ta acknowledge the contribution of Ken A. Matheny and the Simulator Project Staff at the Perry Nuclear Project, Perry, Ohio. The assistance and the critical review of this document by John Burtt, Don Fletcher, Jeff Borkowski, Rex Shumway, and Jay Larson are also appreciated.
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ACRONYMS ,
ADS automatic depressurization system -
ATWS anticipated transient without scram B&W Babcock and Wilcox BFNP Black Fox Nuclear Project BWR boiling water reactor ECCS emergency core cooling system FSAR Final Safety Analysis Repon l
GE General Electric flPCS high pressure coolant system ;
INEL Idaho National Engineering Laboratory i
LOCA loss-of-coolant accident ,
LPCS low pressure coolant system i
MSIV main steam isolation vahes ,
NRC Nuclear Regulatory Commission ,
i PNP Perry Nuclear Project RCIC reactor core isolation coolmg RilR residual heat removal RPS reactor protection system SRV safety relief valves TIC Technical Training Center i
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TRAC-B Thermal-Hydraulic Analysis of the Black Fox Boiling Water Reactor -
- 1. INTRODUCTION P
One of the lessons learned from the accident at Three Mile Island was recognition of the need for ,
effective training for reactor operators. The President's Commission on the Accident at Three Mile Island l Ohe Kemeny Commission)I recommended that all plants have access to a plant-specific simulator for training operators. Such a simulator should be capable of modelling plant operation and transients in an l environment closely resembling the plant control room. Additionally, the Kemeny Commission recommended that research and development be carried out on improving simulation and simulation !
systems in order 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. Today's simulators have evolved into reliable tools :
for simulating nonnal plant operational transients. However. for many other types of transients and accidents. current simulators have not yet been shown to produce reliable results or are unable to model ;
the plant response altogethes, thus reducing the cf fectiveness of the designed intent of the simulator. ,
$ Keeping to the recemmendations established in 1979, the NRC began a project to examine the '
capabilities of the current generation of simulators. Tbc focus of this initiative is to evaluate simulator capabilities under uniqueiy challenging transient scenarios measured against the predictions of advanced ;
I thermal-hydraulic system codes such as RELAP5 ae.d TR AC-B. The simulators to be evaluated reside at i the NRC Technical Training Center (TTC) in Chattanooga, Tennessee. The TIC uses three resident ,
simulators. representing specific Westinghouse, Babcock and Wilcox (B&W), and General Electric plants; in addition, a Combustion Engineering simulator will be installed by the summer of 1992. The ;
scope of this project insches developing advanced system code models of the represented plants, perfomaing a series of transient calculations with the models, and comparing the code results with simulator results. both before and after scheduled simulator upgrade.
l This repon documents the TRAC-B transient analysis of the simulator for the GE Black Fox Nuclear 1 Project (BFNPL Table I reports the six scenarios analyzed. This report will discuss ordy the code results; comparison with simulator data will follow in a later report. Section 2 contains a detailed description of
, the TRAC-B BFNP model used in the analysis. Sections 3 through 8 document the model changes, the
- assumptions specific to each scenario. and the calculated results for the six scenarios. A discussion of the conclusions drawn from the analyses is in Section 9, and a reference list is in Section 10.
1 Table 1. Summary of scenarios analged _ . _.__ _ _ _
l Scenario Initiating event -
1 Inadvertent closure of the main steam isolation vah e 2 Inadvertent initiation of ADS from 409 power /609 flow system conditions j 3 Large-break LOCA i
4 Small-break LOCA, with station blackout ,
5 Turbine trip without scram (with the plant at full power) l i
l 6 _ Turbine trip without scram (wjtighe plant at 401 power!609 flow system conditions) 1 NUREG/CR-5882 ,
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- 2. MODEL DESCRIPTION 1
Following is a desenption of the input model used to represent the Black Fox Nuclear Project (BFNP) I plant in the TRAC-B calculations. This discussion focuses on the modelled thennal-hydraulic components, the control system, and the steady-state initialization. The TRAC-BFl/ MODI (G2W2 ;
versioni computer code' was used to perfonn these calculations. Design specifications from the GE Perry i Nuclear Project (PNP) served as the foundation of the TRAC-B model of the BENP plant. Of the existing i boiling water reactor (BWR) plants, the PNP most resembles the BFNP design: therefore, design ;
specifications from the PNP provided an appropriate substitute for the BFNP design. The model does ,
incorporate available BFNP-specific information pathered from the TTC; however, since the Black Fox ;
facility w as never constructed, detailed infonnation is limited. !
2.1 Thermal-Hydraulic Model The TRAC-B BFNP model includes the major system components. Specifically modelled are the [
sessel and vessel intemals. recirculation loop How paths, main feedwater paths downstream of the main ;
feedwater salves, and the main steam paths upstream of the turbine stop valves, including the main steam i isolation valves (MSIVs). Modelling also included the emergency core cooling and extensive control systems. The BFNP 'lRAC-B steady ^ tate model used 26 components and 38 junctions to simulate the i nuclear steam supply system. Tables 2 and 3 summarize the correspondence between the reactor system i and the model components. Figure 1 illustrates the TRAC-B model nodalization scheme. i The BFNP model describes the thermal-hydraulic flow path, from the feedwater upstream of the ,
reactor vessel to the turbine stop vahe. Feedwater How is initialized from a boundary condition in the feedwater line. The model defines the feedwater flow rate based on steady-state conditions and a control system used during transients. Coolant enters the reactor vessel above the core and follows the downcomer to the low er plenum. Within the downcomer. the coolant is accelerated through 20 jet pumps.
4 The BFNP plant has two recirculation loops, A and B; both are represented in the TRAC-B model. As ;
shown in Figure 2-1, each modelled loop includes a pump suction leg, a reactor coolant pump, and a jet pump component (representing 10 jet pumps). During normal operational conditions, the thermal-hydraulic conditions are identical in both loops. Ilowever. to accommodate the modelling for the two losvof-coolant accident (LOCA) transients in a recirculation loop. modelling of both loops is necessary.
The reactor sessel is represented by a three-dimensional vessel component having three radial rings.
I1 asial levels. and one azimuthal division. Within the vessel component. models exist for the guide tubes. core, and separators. The guide tubes estend from the bottom of the lower plenum to the top of the ,
core plate and contain the control rods and drive mechanism. Two guide tube components represent the mass and fluid volume of 177 puide tubes. Three core channel components represent the 748 fuel i assemblies. One core channel represents the aserage of the 88 low-powered and two core channels represent 660 aserage-powered fuel assemblies. The active length of the fuel assemblies is 3.81 m (123 f t). Each fuel assembly contains 62 f uel rods and two water rods in an 8 rod by 8 rod array. j Abose the core, steam is collected and dried through 300 separators that remove liquid from the ;
steam and return it to the downcomer. The separators are modelled with one separator component. The model for the steam dryers is lumped with the separators. Steam leases the reactor sessel through four ,
main steam'ines which are modelled as one lumped steamline. Along this How path.19 safety / relief valves (SRV s) prevent over-pressuritation. The vah es dump excess steam to the suppression pool, w hich is modelled as a boundary condition. During normal operation. steam flows directly to the turbine. The #
MSIVs also reside in this How path and serse a safety function for mitigating the response to most rapid- l depressurization events. The turbine stop vahe is modelled with a control system m aibw the appropriate ,
amount of sicam to the condenser boundary condition for a given steady-state c.inditior' (e.g., the function of the turbine bypass is lumped imo the turbine stop vahet i
2 N UREG/CR-5882 ,
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-t Model Description. l l
Ring 1 Ring 2 Ring 3 !
i i i Level 11 l l 66 Safety 'l Valves !
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Figure 1. Black Fox sessel nodalization.
1 l
i 3 NUREG/CR-5882
Model Description l
l Table 2. Correspondence betu een the physical and mathematical components in the TRAC-B model of l BFNP. __
_ Physical component TRAC-B component (s) !
l .
Vessej 1 -i I
Guide tubes (1) 31 :
Guide tutes (2) 32 :
Core channels (1) 10 :
Core channels (2) 11 ;
Core channels (3) 12 Separators (1) 71 Recirculation laops i
A pump suction leg 42 A reactor coolant pump 44 A jet pumps 45 l B pump suction leg 52 i B reactor coolant pump 54 l B jet pumps 55 Feedwater line 'I;
'i Feedwater flow boundary condition 20 - !
Feedwater flow control valve 21 Feedwater line to vessel 23 '
Main steamline i !
Piping prior to safety / relief valve 6' Safety / relief valve 64 .
Main steam isolation valve 63 Flow control valve 68 ;
Condenser boundary condition 69 !
Suppression pool boundary condition 66 i
Emercency core cooline system 30- ;
Eressure tans :
I Reference tap 2 i Narrow range tap 4 Wide range tap 6
{
t r
i i
Model Description Table 3. Comparison of the TRAC-B and simulator initial conditions at full and reduced power. ;
i Simulator Simulator
_ _._PJant parameter (100/100) TRAC-B (40/60) TRAC-B - I Reactor pow er (MW) 3546.8 3579.0 1406.5 1406.5 Reactor pressure (psia) 1023 1025 943 935 Vessel narrow range 37 36 37 36 level (in)
Core flow (Mlb/hr) 1(M.0 102.8 59.5 59.8 Feedwater flow (Mlb/hr) 15.2 15.3 5.4 5.5 ;
Steam flow (Mib/hr) 15.2 15.3 5.7 5.5 Recirculation Dow (gpm) 40190 42832 21741 21642 Feedwater temperature (F) 420 420 347 351.8..
Steam temperature (F) 548.0 548.0 537.2 535.4 The model for the reactor coic applies the point kinetics option for a reactor kinetics model. A point kinetics model was chosen over a more detailed one-dimensional model because detailed kinetics data ,
were not immediately available and because the simulator does not use such a model. The decision not to -
use a one-dimensional kinetics model may influence the calculational result by not representing the
. correct Linctic response (or phenomena such as core instability). However, since the simulator does not have a one-dimensional kinetics model, using the point kinetics model climinates uncertainty when comparing TRAC-B and simulator calculations. The Perry Final Safety Analysis Repon (FSAR)3 and the Black Fru Tec/mology Manna 4/ provided the reactor kinetics information for the point kinetics model.
Information required by TRAC-B includes the reactivity coefficient for the void fraction, fuel temperature, and moderator temperature. Control rod reactivity was also incorporated in the model in tabular form. For the transients analyzed, only the two anticipated-transient-without+ cram (ATWS) events used the feedback mechanism in the reactor kinetics. For the other cases, the reactor power responded only to trip-initiated reactivity insertion from the scram table. The Perry FSAR provided the reactivity values for the scram table.
The original BFNP TRAC-B model describes the plant at full power conditions. A second model describes a low-power, low-core-flow version of the BFNP. This model was necessary to ensure a meaningful companson of Aimulator data to TRAC-B results for certain transients. The low-power, low-flow model hs power at ~40% nominal and core flow at -60% nominal. Minor changes in flow loss coefficients represent the cuent of the thennal-hydraulic modelling differences between the models.
2.2 Control System Model Because of the proprietary nature of most of these data, this discussion does not include detailed infonnation regarding relevant setpoints and time constants. locluded in the trip logie were setpoints from l data provided by the Perry Technical Specifications5 and the Perry FSAR. Because of the scope of this 1 project. not all BFNP control systems need to be modelled. For the transients documented in this report.
these control systems were assumed inoperative or not challenged and, therefore. not required in the ,
model. l l
5 NUREG/CR-5882 1
I l
Model Description The televant control sptems used to model the BFNP transients are the reactor protection system j (RPS), isolation controls, emergency core cooling system (ECCS). and the plant systems controls. Delay times in the physical process instrumentation that exist in an actual plant were included in the controls model.
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 -i i
these abnormal conditions. the RPS will signal for an automatic reactor shutdown. Relevant RPS parameters monitored include reactor power, reactor coolant pressure, vessel water level low /high, and ,
MSIV closure. Other RPS control logic not challenged in the transients documented was excluded from ;
the model. j i
The isolation control system handles the control of safety systems responsible for reducing the ;
consequences of the failure of specific components during an accident. The only feature of this system ,
necessary for the transients performed in this study is the main steam isolation. Relevant parameters ,
modelled include vessel pressure and water level. i l
l The ECCS controls safety systems responsible for reducing the consequences of core damage during an accident. The ECCS initiates the automatic depressurization system ( ADS), the high- and low-pressure ;
coolant supply systems fIIPCS and LPCS) and part of the residual heat removal (RilR) system. Since i much of the RilR system applies to long-term cooling and the calculations for this study are concluded l within 10 minutes or less the BFNP model does not include the RllR loops used for heat removal. t Opening a bank of eight SRVs actuates the ADS. ,
i The HPCS, LPCS, and R11R are modelled as a flow boundary condition adjusted by the control :
system dependent on system pressure in the reactor system. All the ECCS core cooling is assumed to j' inject in the core shroud dome at the k> cation of the llPCS and LPCS spargers. The RilR injects coolant from pipes mounted on the base of the core shroud dome directed toward the core bypass surrounding the !
fuel bundles. Because the contribution of reactor core isolation cooling (RCIC) is always less than 10% of ,
total ECCS for all cases except the small break LOCA with station blackout, it has been assumed that i RCIC output is within the error of total ECCS flow and, therefore, will not be specifically modelled.
l Ahhough these systems are located in slightly dif ferent places, modelling the ECCS and RHR coolant injection above the core should not affect the course of the transient significantly to discount calculational resuhs because the cf fectiveness of the RilR model is limited by the nodalization sensitivity of the vessel ;
at the core. The importance of this sensitivity is associated with how well the RiiR coolant mixes with the
- core bypass fluid. Unless the volume in which the RHR directly injects is selectively nodalized, no improvement in calculational resuhs would be expected. It is unreasonable to nodalize in this detail based ;
t on the scope of this study and the Courant time-step advancement limitation.
Plant systems contmis reduce the risk of mer-pressurization transients. These include trips for the turbine, the feedwater supply, and the recirculation pumps. Additionally, initiation of the containment !
spray sptem is handled by this control system. Because of the lack of design information, the SRV Low-Low Set Logic is not included, and the I eedwater Control Setpoint Sctdow n Logic is hardwired for the :
inadvertent MSIV closure scenario, where it plays a major role in the course of the transient.
To provide realisne accounting of the instrumentation used in evaluating reactor vessel water level, the dif ferential pressure transducers are modelled mechanistically. Three pipe components in the downcomer represent the pressure taps used for this measurement. Since thermodynamic property information from the TRAC-H 3-D VESSEL cells are assumed valid only for the cell center, control system inputs used for the water level calculation cannot be extracted directly f rom the model. Therefore.
1-D PIPE components had to be modelled within the sessel with cell centers at the elevation of the actual ,
tap. In consideration of how TR AC-B numerical methods are applied to the TRAC-B VESSEL ;
component, the TRAC-B components are modelled to optimite for accuracy and efficiency. When using ,
I NUREG/CR-5882 6 [
t
_ l
Model Description the VESSEL component. it is recommended that 1-D componens connect to the three-dimensional vessel I
in the center of axia'l boundaries or radial boundaries. To accornmodate this restriction, the two pipe ends of the PIPE component modelling the pressure are connected directly to the same axial boundary, while the center of the pipe is pulled from that elevation to the elevation of the actual tap. This requires _;
modelling the pipe with three cells. The center cell has a leak path to the vessel cell in which it resides.
Cell lengths are made long enough so that flow will not be Courant limited, and the pipe is modelled as frictionless to climinate any such effects. The lesel tracker model is active in the adjacent vessel cells.
This ensures correct soid traction in the pressure tap pipe model. Pressures are measured at the center of the middle cell of these pipes. Actual pressure measurements are filtered through a LAG control system to damp rapid pressure fluctuations that might inadvenently trip control instruments.
2.3 Steady State Conditions A steady state initialization was performed with the TRAC-H BFNP model. Table 2-2 presents the "
comparisons with the simulator data, representing full power and 40% power /609 flow conditions.
respectively. Quantitatise results, escept for the actual power magnitude, were available directly from steadyarate simulator results supplied by the TTC. Other supporting documentation provided by the TTC indicates that the BFNP simulator operates at a power of 99.1% of 3579.0 MW rated full power (3546.8 MW). The TRAC-B full-power calculations were performed at 100% power. The initial diff erence in full power is within a reasonabic uncertainty (<l9 L For comparative purposes. TRAC-H data presented are limited to information that would be immediately asailable to a reactor operator imm the control room instrumentation, except in a few instances where additional inf ormation is needed to support the analysis; Therefore. some infonnation that is typically analyzed for safety analysis calculations. such as fuel temperature,is not presented.
?
P P
i b
i j
- 3. SCENARIO 1: INADVERTENT MSIV CLOSURE The following section details the analysis of an inadvertent main steam isolation valve (MSIV) closure event in the BFNP plant design initiated at full power. The subsections contain a description of the scenario, calculational assumptions and an analysis of the results.
3.1 Scenario Description In an inadvertent MSIV closure, the valve closes spontaneously and remains closed. The assumption about the mode of failure is to exercise the capabilities of the simulator rather than to model a probable failure. The MSIV in this scenario is ramped closed in three seconds. No operator intervention was modelled for this transient.
3.2 Calculation Assumptions :
The basic BFNP model described in Section 2 was used to perform this calculation. The failed-closed .
MSIV transient relies on the RPS and the safety / relief valves to mitigate the potential breach of system integrity introduced by this event. At initiation of this transient, the reactor is operating at full power, and the MSIV is ramped closed in three seconds. The RPS will signal a reactor trip w hen the MSIV is closed beyond 90% area fraction. Feedwater flow rate will decay as a result of the dependence on steam flow to supply the turbine feedwater pumps. Feedwater flow is provided as a boundary condition from the TTC BFNP simulator data for this event. This will eliminate a variable during the comparison of the TRAC-B results and the simulator results. Although the initial MSIV closure will cause a short power spike prior to reactor scram, a negligible amount of energy will be contributed to the system. For this reason, kinetic Icedback will not be incorporated into this calculation (e.g., only kinetic feedback from scram allowed).
3.3 Calculation Results Tahle 4 presents the sequence of events for this transient. Ensuing from the failed-closed MSIV, which ramped closed by three seconds, the RPS initiated shutdown of the reactor at 0.3 seconds tFigure 2L The system pressure rapidly increased as the coolant vapor mass increased in the reactor vessel without being allowed to vent (Figure 3). The initial increase in system pressure collapsed much of the vapor in the two-phase mixture from the reactor core. A sharp decrease in the wide range level measurement (Figure 4) verified this event. The recirculation pumps responded to a Level 2 low-level signal by tripping of f and reducing the flow through the reactor core at 3 seconds (Figure 5). The Level 2 low-water-level signal also began ECCS preparation. As the vapor build-up continued, the system pressure increased bey ond the pressure setpoint for nine SRVs to open at 6 seconds. The wide range les el ,
recovered quickly as the initial pressure shock wave was absorbed by the opening of the SRVs.
Core outlet temperature followed saturation temperature (Figure 6), which increased with system pressure. Core inlet temperature followed core outlet temperature as the coolant recirculation propagated the temperature transient through the vessel. The temperature differential across the core decreased as the power contribution from the core decreased. At 9 seconds, eight SRVs closed as the pressure dropped below the SRV close setpoint, while one SRV remained open to continue over-pressurization relief (Figure 7h Initially, the one SRV was unable to handle the volume of steam generated and the system pressure rose again; however, by 15 seconds, the energy content of the steam decreased to a threshold that could adequately be discharged through one SRV, and the system pressure decreased. Meanwhile, at 11 seconds, feedwater flow had ended from the decay of steam pressure in the turbine pumps (Figure 8).
The core intet temperature increased and the core outlet temperature decreased as the reactor vessel ,
energy content remained nearly constant; that is, the energy released out one SRV was nearly equivalent 8 NUREG/CR-5882
Inadvertent MSIV Closure :;
Table 4. Transient sequence information for Scenario 1.
. Time (s) Event 0.0 h1SIV closure 0.3 Reactor SCRAh! on h1SIV position signal 3.0 Recirculation pumps trip on Level 2 signal l h1SIV fully closed i
6.0 SRVs open in relief mode 9.0 Eight SRVs close on low pressure 11.0 Feedwater flow ends [
15.0 Energy released from one open SRV greater than decay heat energy (e.g., pressure begins to drop) -
32.0 ECCS HPCS starts injecting 54.0 Last SRV closes -
60.0 Calculation terminated to the core decay heat. At 32 seconds, ECCS HPCS begins injecting (Figure 9); however, since the pressure was near operating pressure, flow was restricted and had little effect on the transient. At 54 [
seconds, the pressure deceased below the close setpoint for the last SRV.
Since no other unique phenomena were expected to be obsersed in a continuation of the calculation, it ,
was terminated at 60 seconds. Increasing and decreasing pressure from the SRV operation would continue until the pressure drops below the lowest SRV closure setpoint so that the ECCS HPCS and LPCS and RHR systems could handle the reactor core decay heat levels. The feedwater boundary condition may have some influence with the system pressure. However, its overall effect on the course of events in the calculation of applying this particular feedwater flow signature versus any other signature that reduced the feedwater flow to zero was trivial. The lack of reactor kinetic feedback will not identify a short power increase before the scram. Power increase would be the result of voids collapsing in the core, t decreasing the magnitude of negative reactivity contributed by the voiding. However, since the reactor scrams very quickly (~0.3 seconds), the additional energy integrated over time would not significantly '
influence the course of the transient. In consideration of the no4:inctic-feedback assumption and lack of significant uncertainties, this calculation represents a best-estimate prediction of an inadvertent h1SIV closure.
3 9 NUREG/CR-5882 i
- . - . ~ - . . . . .. . . . . -
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Inadvertent MSIV Closure !
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NUREG/CR-5882 10 l
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E Inadvenent MSW Closure i
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Figure 5. Scenario 1: core now. '
e 1I NUREG/CR-5882 ,
1 l
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F Figure 6. Scenario 1: reactor temperatures. . ,
t L
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NUREG/CR-5882 12 I
.. . . .=. .-
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- Inadvertent MSIV Clasure !
t f
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Figure 8. Scenario 1: feedwater flow. '!
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e r.s - -
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Figure 9. Scenario 1: ECCS flow.
i
+
13 NUREG/CR-5882 i i'
l i
i i
l 4
- 4. SCENARIO 2: INADVERTENT ADS ACTUATION !
t
, t This section details the analysis of an inadvenent actuation of the automatic depressurization system (ADS). The transient initiates with the reactor at 40% nominal power /60% nominal core flow. The i following subsections describe the scenario, calcuhtnonal assumptions, and an analysis of the results. ,
4.1 Scenario Description ,
The inadvertent ADS actuation event begins when the eight SRVs designated for ADS operation spontaneously open and remain open. The transient initiates on a failure in the control system that handles :
ADS actuation. The SRVs in this scenario are ramped open over three seconds. The SRVs discharge to i the suppression pool, which is at atmospheric pressure.
4.2 Calculation Assumptions i i
The 40% power /60% flow BFNP model described in Section 2 was used to perform this calculation.
l The inadvertent ADS actuation transient relies on the RPS and ECCS to mitigate the potential breach in _
system integrity introduced by this event. At initiation of this event, the SRVs are ramped open over three ;
seconds. Reactor vessel coolant mass will discharge through the SRVs to the suppression pool, which is .i exposed essentially to atmospheric pressure. The SRV model uses a form loss coefficient of 1.0, which j approximates the losses through the valves. This uncertainty may affect the siming for system l
- depressurization. Eventually, the reactor will shut down from an RPS signal related to low pressure; this [
will involve applying the scram reactivity table to the point kinetics model (only kinetic feedback from l
scram allowed h ;
J l 4.3 Calculation Results 1
Table 5 presents the sequence of events for this transient. Ensuing from the failed-open SRVs. system j.
- pressure (Figure 10) rapidly decreased as steam escaped from the reactor vessel to the suppression pool. i Flow at the SRV throat was choked as the pressure dropped from operating pressure to atmospheric - {*
pressure across the valve.
I At 12 seconds, a low reactor pressure signaled for the MSIV to close. This then signaled for a reactor shutdown (Figure i1). Water lesel began to drop as a result of coolant shrinkage as core power dropped f (Figure 12). Shutdown of the feedwater supply was assumed to lag about 8 seconds behind MSIV closure. !
Ily 20 seconds, the feedwater now entering the vessel was negligible. Loss of feedwater flow into the !
t vessel was reflected as a sharp drop in water level since the coolant mass contribution from the feedwater was no longer present. The loss of feedwater immediately affected the core flow since less coolant was !
then available to circulate in the reactor vessel (Figure 13).
For the next -100 seconds, system pressure continued to decrease as reactor vessel inventory j continued to leave the system. Phase change inertia prevented system temperatures from deviating from j the saturation temperature (Figure 14), which decreased with system pressure. With the system -i depressurizing, and without feedwater now, voiding in the vessel increased as a resuh of liquid Dashing to t steam (Figure 15). Similarly, SRV Gow decreased because of the reduction in the pressure drop across the ,
SRVs (Figure 16). During this phase of the transient, no mass entered the system, mass continued exiting ;
the vessel through the SRVs, and water level slowly decreased from boiling as a result of decay heat.
1 J
t 14 NUREG/CR-5882 !
t I
I
. . .- - . . = .. . -
~
Inadvertent ADS Actuation
-- p
.- Table 5. Transient sequence information for Scenario 2. ,
Time !
(s) Event
-t 0.0 Eight SRVs open on ADS actuation i i
t 12.0 Reactor SCRAM on MSIV closure from low pressure signal l
20.0 Feedwater Dow ends 1 105.0 Recirculation pumps trip on Level 2 signal l 132.0 ECCS begins injecting !
160.0 Calculation terminated .!
f At ~105 seconds, a Level 2 low water level signaled a recirculation pump trip and started ECCS following a 27-second delay. With the coast-down of the pumps, the core flow decreased. At 132 seconds, the ECC system began pumping coolant directly to the reactor core (Figure 17). The ECC coolant flowing i counter to the main flow further reduced the core Dow. This also decreased the water level by reducing the dynamic contribution of the pressure drop measurement for water level. j l
At 160 seconds, the calculation was terminated, since all the significant events relevant to analysis !
had occurred. Without detailed pressure drop data on the SRVs, uncertainty exists for the depressurization :
timing because of unknown losses through the SRVs. Additionally, without the kinetic feedback model !
initiated in this model, energy input into the system may be overestimated, since core voiding would bc ;
expected to decrease core power before the scram. The no-kinetic-feedback assumption conservatively l represents the power transient. Since no unique phenomena can be attributed to the higher power, it can :
be extrapolated that the transient signatures would be very similar had the kinetic feedback option been .l' incorporated. Considering these uncertainties, this calculation represents a best-estimate prediction of the .
events following an inadvertent ADS actuation based on the known design.
l l
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Figure 10. Scenario 2: system pressure.
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Figure 11. Scenario 2: reactor power.
t NUREG/CR-5882 16 t
Inadvertent ADS Actuation
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Figure 13. Scenario 2: core now.
Inadvenent ADS Actuation
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Figure 14. Scenario 2: reactor temperatures. ,
it 0.750 , , !
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NUREG/CR-5882 18 l
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This section details the analysis of a double-ended guillotine rupture in the BFNP recirculation loop, specifically, in the pump suction piping. 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.
5.1 Scenario Description The pump suction large break for BFNP is the instantaneous non-isolatable, double-ended guillotine .
rupture located in the pump suction piping of recirculation loop A. On initiation of the transient, the ,
coolant released from the break will choke at the throat from the high pressure differential across the break exit. Emergency core cooling is expected to mitigate the effects of this event and prevent core damage. No operator intervention was modelled for this transient.
5.2 Calculation Assumptions i The basic BFNP model described in Section 2 was used to perfonn this calculation. For the large- ;
break LOCA in recirculation loop A. the BFNP input model includes modelling of the break in the pump suction pipe component. The model does not include the containment. An accurate containment model would ensure that the break flow is accurate following sufficient system depressurization to unchoke the break plane and that the RPS could monitor drywell pressure, which is one of the variables examined to -
determine whether a reactor shutdown is necessary. However, since events in a large-break LOCA occur ,
rapidly and the course of the transient is short. these containment contributions are negligible. A form loss coefficient of 1.0 is applied for the abrupt expansion at the break plane. The form loss will affect the :
system depressurization rate and represents an uncertainty, it is expected that the pump will inhibit
. reverse flow in the pump-side break path. The rapid loss of coolant from the system will trigger many ,
control system responses to low-water-level trip signals. These will include MSIV closure, ECCS injection, and ADS actuation. No time delay was assumed between an ADS trip signal and actual ADS actuation. A reactor shutdown signal from the MSIV closure will result in a negative reactivity '
representing control rod insertion to be applied to the reactor kinetics calculation (only kinetic feedback from scram allowedL 5.3 Calculation Results Table 6 provides a summary of the sequence of events that occuned in the large-break LOCA in the !
pump suction piping in recirculation loop A. At the initiation of the LOCA, the break flow from the ['
- vessel side immediately choked, while break flow from the pump side was inhibited by the pump rotor (Figure 18) (e.g., did not choke). The massive loss of coolant resulted in a rapid decrease in the system pressure and water level (Figures 19 and 20). At 3 seconds, the reactor was shut down (Figure 21)in response to a Level 3 low-water-level signal. By 5 seconds, the control system had signaled Level 3.
Level 2. and Level 1 low-water-level trips. These signals resulted in closure of the MSIV, shutdown of the recirculation pumps, and the preparation for high pressure coolant injection.
In the first 40 seconds, the system pressure dropped nearly linearly from 7.2 MPa to 2.0 MPa. This 7 resulted from the massive loss of fluid out the break and from the primary coolant contraction resulting ;
from the sudden reduction in thermal energy supplied by the reactor. Simultaneously, flow out the break and the SRVs (ADS actuation) (Figure 22) removed much of the mass out of the reactor vessel, as demonstrated by the depressed amount of mass in the lower plenum (Figure 23). Ilowever, following the t
~ _. ___. . .__ _ __ _ _
j :
Large-Break LOCA in Recirculation Loop - 'i Table 6. Transient sequence information for Scenario 3. j Time I (s) Event ;
i 0.0 Large -break LOCA .
4.0 Reactor SCRAM on Level 3 low-water-level signal i 5.0 Recirculation pumps trip and ECCS activates on Level 2 '
signal 33.0 MSIV closes on Level I signal. ADS actuates j 40.0 liPCS initiates t
120.0 ECCS coolant injection begins to compensate for lost mass l 160.0 Flow instability in core ,
240.0 Calculation terminated l initial blowdown of the reactor vessel, the break flow stabilized at a rate that the ECCS HPCS could - ll f
compensate (Figures 24 and 25). At 120 seconds, a flow oscillation developed in the core as a result of the ECCS reflood in the core (Figure 26). This oscillation was driven by the reflood of the vessel, which ;
generated a local pressure oscillation in the core as a result of steam being condensed from the ECCS. r Behavior of the flow in the intact loop (Figure 27) was normal, as coolant flow decreased with the coast- [
down of reactor coolant pump and with the loss of flow circulation when the level of coolant water in the ;
reactor vessel dropped below the recirculation loop intake lines. Throughout the transient, phase change ;
inertia prevented system temperatures from deviating from the saturation temperature, which decreased i with system pressure (Figure 28).
1 At 240 seconds the calculation was terminated. Continuation of this transient would eventually result .}
in a steady state condition, with flow from the ECCS matching the flow out the break with the core ;
partially reflooded. Simulating this stage of the transient should be deferred until more information is j obtained about the performance of the BFNP simulator during the early phases of this transient. Since no j delay time was modelled for the ADS, the ADS is actuated earlier than would be expected; however, the ;
impact of this assumption is minimited since the system rapidly depressurizes from liquid mass lost out :
the break. In conclusion, while uncertainty exir,ts about the interaction of a containment model and' ,
depressuritation rate dependence on the break plane form losses, this calculation represents a best- .j estimate prediction of the events following a large-break LOCA based on the known design. ;
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- 6. SCENARIO 4: SMALL-BREAK LOCA, WITH STATION BLACKOUT This section details the analysis of a small-break loss-of-coolant accident, coincident with a station power blackout (i.e., loss of offsite and diesel backup power) in the BFNP plant design. 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.
6.1 Scenario Description This calculation models a 2% area small-break LOCA in the pump suction piping in the recirculation loop. On initiation of the transient, the coolant released from the break will choke at the throat from the high pressure differential across the break exit. Normally, emergency core cooling is expected to mitigate the effects of this event and prevent core damage; however, the ECCS will be unavailable as a result of -
the loss of power. No operator intervention was modelled for this transient.
6.2 Calculation Assumptions The basic BFNP model described in Section 2 was used to perform this calculation. For the small-break LOCA in the recirculation loop pump suction piping, with concurrent station blackout, the BFNP input model includes modelling the break in the recirculation loop A. The initial break flow will be equal to the critical flow rate for the break area and pressure for the vessel side break path. As with the large-break LOCA transient, the model does not include the containment. The uncertainty of not modelling the containment may influence the sequence and timing of events predicted by TRAC-B; however, without specific design information on the containment cooling mechanism and confidence in the TRAC-B condensation model, not modelling the containment minimizes the uncenainties. A form loss coefficient of 1.0 is applied for the abrupt expansion at the break plane. The form loss will affect the system depressurization rate and represents an uncertainty. The MSIV will fail closed because of the loss of power, and the loss of power will discharge a solenoid vahe in the control rod drives, allowing reactor pressure to hydraulically drive the control rods into the core (only kinetic feedback from scram allowed).
The recirculation pumps will trip, and the ECCS pumps will not be able to respond to a Level 2 low-level signal during the station blackout; however, the reactor core isolation cooling (RCIC) turbine pumps will be able to provide some coolant. Simulator data of the flow rate from the RCIC system have been approximated and incorporated into the input model as a boundary condition during the calculation. The RCIC pumps will respond following a 10-second delay. The station blackout will also prevent the SRVs from opening in relief mode and in ADS mode.
6.3 Calculation Results Table 7 provides a summary of the sequence of events that occurred in the small-break LOCA in the recirculation loop A pump suction piping, with station blackout. The initial response to the LOCA with station blackout was a reactor shutdown on the loss of power (Figure 29). This loss of power discharged a solenoid valve in the control rod drives allowing reactor pressure to drive the control rods into the core; the power loss also caused the MSIV to fail closed. Additionally, the loss of power tripped the recirculation pumps; therefore, core flow began to decrease with the pump coast-down (Figure 30). The water level decreased with the loss of fluid out the break and the coolant shrinkage from the sudden reduction in thermal energy (Figure 31). Because of the MSIV closure, the system pressure (Figure 32) initially increased as the energy generated in the reactor core was greater than the energy lost out the break. At 19 seconds, the increasing pressure surpassed the threshold for opening eight SRVs in safety 28 NUREG/CR-5882 1
_- _ - __ _ - __ - a
Small-Break LOCA with Station Blackout Table 7. Transient seguence information for Scenario 4.
Time (s) _
Event 0.0 Small-break LOCA, reactor SCRAM and recirculation pumps trip 3.0 MSIV closure from low-water-level signal 19.0 Eight SRVs open 23.0 RCIC begins injecting 320.0 Last safety-related system (ADS) inoperable without power 500.0 Calculation terminated.
mode. The valves quickly closed as pressure dropped below the valve-cic, sing threshold, set ~240 KPa below the opening threshold in safety mode (Figure 33). At 23 seconds, the RCIC began injecting coolant into the core (Figure 34). liowever, without the other ECCS pumps contributing to the cooling of '
the core and replacing lost inventory, the water level and system pressure continued to decrease.
Following closure of the eight SRVs, the reactor vessel began a steady pressure oscillation, which -
continued for the duration of the transient. The pressure oscillation was driven by the decay heat energizing the vessel while the SRVs relieved the high pressure. The break flow (Figure 35) remained choked during this time and varied directly with the system pressure. Throughout most of the transient, phase change inertia prevented core outlet temperature from deviating far from the saturation temperature, which oscillated with system pressure (Figure 36). Core inlet temperatures increased toward the saturation temperature from the decay heat input.
At 320 seconds, the water level dropped below the Level 1 low-water-level threshold. Had electric power been available, ADS would have actuated. This represents the last safety-related system that would have worked to mitigate this transient. At 500 seconds, the calculation was terminated. Continuation of the transient would eventually result in the complete boil-off of the reactor vessel coolant, fuel rod dryout, and subsequent cladding temperature excursions. Simulating this stage of the transient should be deferred -
until more information is obtained about the performance of the BFNP simulator during the early phases .
of this transient. Since the reactor shutdown signal occurred in the beginning of the transient, the uncertainty in the reactor shutdown timing with respect to a high containment pressure signal is negligible. In the later stages of the transient, the lack of a containment model would not have influenced the mass lost from the system, since the break remained choked throughout the transient. Therefore, considering the available design information, the calculation represents a best-estimate prediction of a small-break LOCA, with station blackout.
o 29 NUREG/CR-5882
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NUREG/CR-5882 32 i
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Figure 36. Scenario 4: reactor temperatures. i 33 NUREG/CR-5882 l
1
- 7. SCENARIO 5: TURBINE TRIP WITHOUT SCRAM This section details the analysis of a turbine trip accident without reactor scram.-The transient occurs with the reactor at full power. The following subsections contain a description of the scenario, i
calculational assumptions, and an analysis of the results.
7.1 Scenario Description This transient involves tripping the turbine, with an additional malfunction that prevents the automatic scram signal from tripping the reactor. The turbine trip transient is the closure of the turbine stop valve that is designed to protect the turbine from over-pressurization and liquid contamination. No operator action was modelled for this transient.
7.2 Calculation Assumptions ;
The basic BFNP model described in Section 2, with the addition of a feedback kinetics model, was used to perform this calculation. In contrast to previous calculations, turbine trip without automatic scram requires a Linetics 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. This kinetics model incorporates the reactivity coefficients for a point kinetics model representing beginning- !
of-life sy stem conditions for the Black Fox Nuclear Plant. The reactivity coefficients and point kinetics assumption constitute an uncertainty in this scenario.
Failme to scram is the result of a hypothetical lockup of the rod drise system, so that the control rods contribute zero reactivity in the TRAC-B simulation. All other control systems will work correctly during this calculation. No time delay was assumed after an ADS trip signal and actual ADS actuation. Based on ;
information available during model development, a Level 8 trip was assumed to actuate HPCS.
The closure of the turbine stop valve requires an addition to the BFNP model. The addition includes a flow control vahe downstream of the MSIV. On initiation of the transient, the flow control valve reduces ,
the flow area until the flow rate that passes to the BREAK component is equivalent to the capacity of the turbine bypass v.alves at full power. Flow that is allowed to pass to the BREAK component is assumed to be that Dow that would nonnally go directly to the main condenser through the turbine bypass valves. The capacity of the bypass valves is approximately 359 of the Dow at full power. The available design ;
information did not provide an absolute value for this; therefore, this represents an uncertainty for this calculation.
7.3 Calculation Results Table 8 presents the sequence of events for this transient. Following the initiating event (turbine trip),
the system pressure rapidly increased (Figure 37) beyond the relief valve setpoints for opening nine SRVs. Water lesel fluctuated inversely with pressure as voids collapsed and formed (Figure 38); at 7 seconds, high pressure tripped the recirculation loop pumps and the feedwater pumps and began ECCS HPCS preparation for injection. This could be seen as a reduction in core flow as the pumps coast down (Figure 39) and a reduction in water lesel as the feedwater io longer contributed a dynamic contribution to the differential-pressure-based level instrumentation. The reactor power responded to the initial pressure shock wave by dramatically increasing from collapsed voids, before retreating as voiding increased from the high power (Figure 40). Because the RPS was not operating normally, a reactor shutdown signal was not received, initially for closure of the turbine stop valves, and later for high neutron Out indicated by the high power, or the Level 8 high water level. The power Ductuated near 1.2 .
GW, depressed fron. the nominal power by the large negative reactivity addition from the core voiding.
r Turbine Trip Without SCRAM -l I
Table _8 Transient _se_quence information for Scenario 5._ .!
Time (s) Event 0.0 Turbine trip f
2.0 Nine SRVs open on high pressure; recirculation pumps trip, and ECCS IIPCS prepares I on Level 8 high-water-level signal; feedwater pumps trip. :j i
5.0 Power adjusts to ~1.2 GW
<30.0 Eight SRVs shutter with pressure ] L 32.0 ECCS injection begin ,
i 110.0 MSIV closes. ADS initiates on Level 1 low-level signal !
160.0 Calculation terminated t i
At this power lesel, the turbine bypass and main condenser could handle the load. Eight of the SRVs shuttered on and off over the initial 30 seconds until the excess energy from the initial power spike dissipated through the SRVs (Figure 41). Reactor core outlet temperature (Figure 42) was equivalent to the saturation temperature and fluctuated with the system pressure. As the power decreased, the temperature differential across the core decreased.
l Because the RPS was not operating normally, a reactor shutdown signal was not received for the Level 3 low-water-level signal that occurred at 18 seconds. Meanwhile, ECCS began injecting coolant ;
from the llPCS and the LPCS at 32 seconds (Figure 43). The core temperature differential was reduced .j further following IIPCS initiation because the colder ECCS coolant and the coolant that had come i through the core mixed. Core voiding increased as the system pressure dropped faster than the reactor _[
vessel liquid inventory could cool to prevent phase change (Figure 44). Throughout the rest of the ' '
transient, phase change inertia prevented system temperatures from deviating from the saturation temperature, w hich decreased with system pressure.
For the next 80 seconds, the total reactor vessel energy decreased as the system adjusted to the lower power level. This is evident from the decrease in pressure. Water level decreased as steam continued to be ,
generated and leave the reactor vessel. At -110 seconds, the MSIV closed and the ADS initiated on a l
Level 1 low-water-level signal. This resulted in an increased depressurization rate and a lower power '
lesel from voids created from depressurization-driven coolant flashing. Again, had the RPS been _;
operating normally, a reactor shutdown signal would have been received from the MSIV closure.
At 160 seconds, the calculation was temiinated. In conclusion, the response of the kinetics model had the .' ;
. greatest influence on this anticipated transient without scram. The kinetics model defined how much :
energy was entered into the vessel. Thislaffected the rates at which the system pressure and water level !
reached the setpoints for the control system needed to mitigate this transient. The importance of the ;
kinctics model accentuates the uncertainty' associated with using a point kinetics model, since it is !
possible that dimensional effects could affect the direction of this calculation. However, the resuhs .
demonstrate that point kinetics is a necessary minimum requirement for this transient. Additionally, the ' !
capacity and control of the turbine bypass valves influenced the timing of pressurization rates. The ;
assumption of controlling the bypass valves to allow a maximum of 35% of the full-power flow will }
}
f 35 NUREG/CR-5882 !
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- w
,0 3.76 0
- C .
i
' 2.50 0
p i
~
1.2 S v %- . 1 I i "
= ^%, .
i 0.00 ' -
9 0.0 20.0 40.0 60.0 80.0 100.0 120.0 140,0 180.0 180.0 '
Tim e (s) i Figure 40. Scenario 5: reactor power. !
t I
i I
37 - NUREG/CR-5882 '
i
]
Turbine Trip Without SCRAM 2000.0, , , , , , i i i c 3 Steam
- 1800.0 -
g 3 ggy
, 4
- tr, 1200.0 a
800.0
- 400.0 2
- b -c ,
3 ; C ; WC- 0 3 -
0.0 t
-400.0 O.0 20.0 40.0 60.0 B 0.0 100.0 120.0 140.0 180.0 180.0 Time (s)
Figure 41. Scenario 5: main steam and SRV flow.
600.0 , .
o---o Co re Inlet c 3 Core Outlet
, ----- Saturation 58 G.0 b -
2 ~
660.0(L -
s.
"+ ,
c 540.0 -
e 520.0 - '
a 500.0 O.0 20.0 40.0 , 80.0 80.0 100.0 120.0 140.0 180.0 180.0 Time (s)
Figure 42. Scenario 5: reactor temperatures.
1 I
Turbine Trip Without SCRAM i i
I I
360.0 , , , , , ,
i l
l 300.0 -
i T
N 250.0 -
?
=
5 ,
3 200.0 -
2
=
0 160.0 -
- ~i 2 100.0 -
[ -
[
l no.O -
/ -
I 0.0 :, c' : '
0,0 20.0 40.0 50.0 80.0 100.0 120.0 140.0 160.0 180.0 Time (s) .i t
Figure 43. Scenario 5: ECCS flow. !
r 0.760 , i r '
l l ,
0.62S )h !
+
0.600 - ~
c o !
s U i 2 0,3?S -
- [ !
E o $
> i 0.250 -
- f I
L 0.125 - ~
s l
4 0.000 6 '
'0.0 20,0 40,0- 80.0 80.0 100.0 120.0 140.0 160.0 180.0 1
~
Tim e (s) .
i i
Figure 44. Scenario 5: core inlet soid fraction.
i f
39 NUREG/CR-5882 :
+
+-. c,- , , , -
-r -- y
l l
l Turbine Trip Without SCRAM ensure that coolant mass is lost at the maximum possible rate, thus advancing the transient more quickly than if the calculation were performed with a detailed turbine bypass model. Unlike the kinetics model that describes the energy input for the vessel, this quantity defines a limit for energy departure from the vessel; thus, its influence on the course of the transient is more understood. Without a more detailed kinetics model, this calculation represents a best-estimate prediction of the turbine trip with ATWS. I liowever, if it is determined that training simulators need a more accurate kinetics model, this study !
should be redone to resolve this uncertainty. !
1 i
f l
i i
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j
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.: t t
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L
- 8. SCENARIO 6: 40% POWER / 60% FLOW TURBINE TRIP ATWS This section details the analysis of a turbine trip accident without reactor scram. The transient occurs with the reactor at 409 nominal full power. The following subsections contain a description of the ;
scenario. calculational assumptions, and an analysis of the results. '
8.1 Scenario Description This transient involves tripping the turbine. with an additional malfunction that prevents the automatic scram signal from tripping the reactor. The turbine trip transient is the closure of the turbine stop valve that is designed to protect the turbine from over-pressurization and liquid contamination. This transient is unique to scenario 5 with respect to the reactor power (40% nominal) and recirculation loop Dow rate (609 nominal). No operator action was modeled for this transient. )
8.2 Calculation Assumptions The basic BFNP model described in Section 2, with the addition of a kinetics model, was used to ;
perfomi this calculation. In contrast to the previous calculations, the power level is 40% nominal and recirculation now rate is 60% nominal. This kinetics modelincorporates the reactivity coefficients for a point kinetics model representing beginning of life system conditions. The reactivity coefficients and ;
point kinetics assumption constitute an uncertainty in this scenario. Failure to scram is the result of a hypothetical lockup of the rod drive system, so that the control rods contribute zero reactivity in the TRAC-B simulation. All other control systems will work correctly during this calculation. No time delay ,
was assumed between an ADS trip signal and actual ADS actuation. Based on infonnation available i during model development, a Level 8 trip was assumed to actuate HPCS. The closure of the turbine stop vah e requires an addition to the BFNP model. This addition includes a now control valve downstream of the MSIV. On initiation of the transient, the flow control valve reduces the flow area until the flow rate that passes to the BREAK component is equivalent to 359 flow at full power. Flow that is allowed to pass to the BREAK component is assumed to be that now that would normally go directly to the main condenser through the turbine bypass valves. The available design information did not provide an absolute value for this; therefore, this represents an uncertainty for this calculation. j 3
8.3 Calculation Results i Table 9 presents the sequence of events for this transient. Following the initiating event (turbine trip),
the reactor pressure increased at a lower rate (Figure 45), relative to the ATWS response in Scenario 5.
Likewise, during system pressurization, the power increased as the moderator void reactivity decreased from voids collapsing in the core (Figure 46). Because the RPS was not operating nonnally, a reactor shutdown signal was not received initially for closure of the turbine stop valves. Pressure increased until, at 32 seconds, nine SRVs opened in relief mode (Figure 47). Additionally, before the opening of the SRVs the pressure increased above the setpoint to trip the recirculation pumps. This resulted in a decrease in core flow (Figure 48). Following these events. reactor power dropped near 1 GW and temporarily remained stable. At this power level the turbine bypass and main condenser could handle the i load. Pressure decreased in response to the lower power level and to the coast-down of the recirculation t
pumps. Reactor core outlet temperature (Figure 49) was equivalent to the saturation temperature and fluctuated with the system pressure. As the power decreased. the temperature differential across the core ;
~
decreased. Meanw hile, reactor water level (Figt.re 50) maintained a steady increase, as feedwater supplied more coolant than steam was removed from the vessel. until, at 185 seconds, the feedwater pump tripped in response to a Level 8 high-water-level signal. The reduction of mass in the downcomer resulted in a i decrease in water level measurement. The Level 8 high-water trip also activated the ECCS HPCS. which <
began injecting coolant to the core after the normal 27-second delay (Figure 51). Because the RPS was 41 NUREG/CR-5882
?
40% Power /60% Flow Turbine Trip ATWS l l
1 Table 9. Transient sequence information for Scenario 6.
Time (s) Event .
0.0 Turbine trip l h
32.0 Nine SRVs open on high pressure; recirculation pumps trip j i
45.0 Power adjusts to ~1.0 GW l I
185.0 No feedwater on Level 8 high-water level 208.0 ECCS injection begin: MSlV closes j
]
308.0 ADS initiates on level I low-level signal 350.0 Calculation terminated
.i t
i not operating nonnally, a reactor shutdown signal was not received for the Level 3 low-water-level signal that occurred at 205 seconds. i i
Following ECCS activation, the MSIV closed in response to low pressure in the steam line. With the .i closure of the MSIV, the pressure rapidly increased beyond the setpoint to open nine .SRVs. As the ,
)
pressure increased. the core temperature differential increased as the core outlet temperature followed the-increase in saturation temperature, which is dependent on system pressure. Core inlet temperature slowly -
increased to match the core outlet temperature as fluid recirculated in the vessel. Again, had the RPS been i operating normally, a reactor shutdown signal would have been generated for the MSIV closure. The l combination of ECCS injection and the closure of the MSIV established a feedback loop in which reactor ;
pressure responded to changes in reactor power, which responded to changes in the ECCS IIPCS i injection, which responded to changes in reactor pressure. Meanwhile, reactor water level decreased with ;
the loss of reactor vessel inventory from the SRVs. At 308 seconds, a Level I low-water-level signal f initiated ADS and the nine SRVs remained open. This interrupted the feedback loop response. The- {i continued loss of mass from the reactor vessel quickly reduced the reactor pressure. i At 350 seconds the calculation was terminated. The response of the transient was similar to Scenario 5. in which the kinetics model had the greatest influence: As before, the kinetics model defined ;
how much energy was entered into the vessel. This affected the rates at which the system pressure and water level reached the setpoints of the control systems needed to mitigate this transient. The imponance of the kinetics model accentuates the uncertainty associated with using a point kinetics model, since it is '
possible that dimensional effects could affect the direction of this calculation. Ilowever, the results i demonstrate that point kinetics is a necessary minimum requirement for this transient. Additionally, the capacity of the bypass valves influenced the timing of pressurization rates. As stated in Scenario 5, its ,
inDuence on the course of the transient is of lesser importance than that of the kinetics model. Without a more detailed kinetics model, this calculation represents a best-estimate prediction of the low power / low Dow turbine trip with ATWS. flowever, if it is determined that training simulators need a more accurate kinetics model, this study should be redone to resolve this uncertainty.
40% Power /60% Flow Turbine Trip ~ATWS t
D 7 s.o , , , ,
- 7. 0 - -
i
% ~
c :
1 8.0 - -
}
3 i
s.0 - - )
4.0
\ !
5 0.0 40.0 80.0 120.0 180.0 200.0 240.0 280.0 320.0 -360 0 !
Time (s) i Figure 45. Scenario 6: system pressure. !
i t !
p 5.0 1 , , .
t 4.0l -
t e 3.0 - - f E :
t I E k ,
e.0 ( j
=- -
hf 4 -
- ^
3.0 h % - -c
{
f .
'J dt,ih u(1,, !
0.0
, , , b -;
0.0 40.0 80.0 120.0 160.0 200.0 240.0 280.0 320.0 360.0 t Time (s)
Figure 46. Scenario 6: reactor power. ;
i 43 NUREG/CR-5882 ;
t 40% Power /60% Flow Turbine Trip ATWS l
.t l
1800.0 ,
C 3 Steam !
1880.0 "
W SRV -
g }
f <
1000.0 -
2 !
g ,
3 6 l l 3 vs 0.0 L . !
n & '
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he i h i s a ;a h. M == I 0.0 IM
- : : : g. _- - C .' ,
-280.0 ' ' ' ' i t
0.0 40.0 80.0 120.0 100.0 200.0 240.0 260.0 330.0 360.0 t Time (s) .
i Figure 47. Scenario 6: main steam and SRV flow.
?
8000.0 , , , , , ,
I 7000.0N % -
6000.0 - -
= <
6 <
u S00 0.0 {> -
3
- 8. 4000.0 - 0 : : M -
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J- 4 2 3000.0 - J -
, p.
l.
2000.0 - -
'1 1000.0 0.0 40.0 80.0 120.0 180.0 200.0 240.0 280.0 320.0 360.0 Time (s)
Figure 48. Scenario 6: core flow.
40% Power /60% Flow Turbine Trip ATWS 800.0 ,
I o----c Core Inlet 0---o Co re Outlet S gt.6 y
Saturatton j Q 878.0 -
E 1
! "* 5 ONN. i E,
W 560.0 -
l nat.s - - -
f.2 5. 0 0.0 40.0 80.0 120.0 100.0 200.0 840.0 880.0 330.0 360.0 Time (s)
F!gure 49. Scenario 6: reactor temperatures.
240.0 j v Y ~
y -
2 0 0.0 4R l e 0.0 -
y e l\ -
c
=.
g 120.0 - +
> A D
ac.0 -
{; ,
40.0 r '
(
O.0 O.0 40.0 50.0 120.0 180.0 200.0 240.0 280.0 020.0 380.0 Time (s)
Figure 50. Scenario 6: wide-range water level.
I 1
l 45 NUREG/CR-5882 i
40% Power /60% Flow Turbine Trip ATWS i
l l
280.0 . , , . . . . i
, j a .i 240.0 -
~
T N
200.0 - I r
ec ,
6 .
i r
3 1 e 0. 0 -
j e !
i
.2 120.0 - .
m <
m ,
e ~
2 80.0 -
, :7
)'\ \ 'bh\
40.0 - 0 -
f
+ ..t I I L i I gg 3 .; 7 3 0,0 40.0 80.0 120.0 160.0 200.0 240.0 280.0 320.0 380.0 ?
Time (s) .
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Figure 51. Scenario 6: ECCS flow.
e i
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I NUREG/CR-5882 46 ;
r f
1
- 9. CONCLUSIONS AND RECOMMENDATIONS Analyses of six BFNP accident scenarios were performed with the TRAC-B computer code. These calculations were perfonned in order to benchmark the TTC BFNP simulator with TR AC-B 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. Computational information presented in this report is a sample of a much more detailed data base calculated by the TRAC-B code. Additional data for these simulations are stored on magnetic tape and maintained at the INEL These data will be used for future TTC simulator / TRAC-B benchmark comparisons.
While limits in some of the modeling assumptions have been identified during analysis of the transient set (i.e., application of a general containment model and point kinetics model) the results presented have been judged by experienced analysts to represent a reasonable best-estimate prediction of the thennal-hydraulic behavior expected in the reactor during the abnonnal transient scenarios. Since the ,
Black Fox simulator at the TTC shares similar limitations, increasing the detail in the TRAC-B model to remove these limits is not expected to significantly improve the applicability of these results for comparison with the simulator results. For comparison purposes, it is useful to maintain some similar modeling strategies, so that other more esoteric differences can be identified. Although comparison with -
simulator data will be used primarily to assess simulator performance. it will also be used to assess the '
code's ability to model the more mechanistic phenoinena of plant behavior (i.e.. boundary conditions like feedwater flow, core power. etc.) that the simulator may do better. A detailed comparison study will probably result in recommended improvements to both TRAC-B and the simulator model.
In general, the predictions of TRAC-B were reasonable for the scenarios 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. Choices of l' boundary conditions and modeling assumptions used in these analyses might influence the results; howeser. such assumptions have been made consistent with the Black Fox simulator when possible. This practice will be considered in the final comparisons of simulator and TRAC-B results for these transients. ,
4 i
+
l i
- 10. REFERENCES
- 1. Report of the President's Commission on the Accident at Three Mile Island, Washington, D. C.,
October 1979.
- 2. M. M. Giles, et. al., TRAC BFil MOD 1: An Adeanced Best Estimate Computer Programfor BWR Accident Analysis.,NUREG/CR-4356. EGG 2626, May 1992.
- 3. Cleveland Electric Illuminating Company, Perry Nuclear Plant. FinalSafen Analysis Report.
l
- 4. U.S. Nuclear Regulatory Commission Black Fox Technology Manual, Nuclear Regulatory ;
Commission Technical Training Center Documents, February 1980. l 1
- 5. Cleveland Power Authority. Perry Nuclear Power Plant Unit 1 Technical Specifications, November 1985.
t I
I NUREG/CR 5882 48
geo== ss us.=vetsAm nacutATonY eeuwasion i.~gT,wys IIE8E """
BIBUOGRAPHIC DATA SHEET rs , .n a,.,.e,,,,,,, i NUREG/CR-5882 2.Vitta A~c sustif t EGO-2677 TRAC B Thermal Hydraulic Analysis of the Black Fox Bolling Water Reactor 3, ,,,,,,,,,,,y, ,,,
w0 ta *6.s g
May 1993 4, fin QR GRANT NvM64 A r L1050 E. ALTMQO6)
- 6. TYPE OF RsPQRT R. P. Martin Tbchnical L P L A eoD cow E R & D , awwww pee =,
- 6. F F A ,dra h4AI10N - AAMI Aho ALLR EE5,,e.ac.p.e see oesma, cum, or A.psa, u snar.a es,weewr cm anermamme empees a so.awsee peews, Idaho National Engineering Laboratory EO&O Idaho,Inc.
P.O. Box 1625 Idaho Falls, Idaho 83415 3.gcgRgsAwaA1.os - $w: A8o Aconus ,, ..c. 3 w, .n -.ac e. o, . u m w ,c-DivisionofSystems Research Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Cormnission Wash!ngton, D.C. 20555
- 10. sUPPLEME NT ARY NCTEE j
i1. Abst RACT trao som w s.a.,
Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electdc Black Fox Nuclear Project bolling water reactor were performed using the TRAC-BF1 computer code. 'Ihis work is sponsored by the U.S. Nuch Regulatory Commission and is being done in coglunction with future analysis work at the U.S.
Nuclear Regulatory Commission Tbchnical Training Center in Chattanooga, Tbnnessee. These accident scenados were chosen to assess and benchmark the thermal hydraulle capabilities of the Black Fox Nuclear Project simulator at the 'Ibchnical TYaining Center to model abnormal transient conditions.
I a. n a v woaesect 6Ca.eica s n.,-. . , . . . ,,,, ..t..,utvitate m t is
'Ihermal-hydraulle capabilities, abnormal transient conditions, plant simulator TRAC DFl Un11m!ted 14.&ECw01t W E 8#L.T 4h IThierar.:
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