ML20056D872: Difference between revisions
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| number = ML20056D872 | | number = ML20056D872 | ||
| issue date = 08/31/1993 | | issue date = 08/31/1993 | ||
| title = Nonproprietary WCAP-13811, BVPS Unit 1 30% SG Tube Plugging Analysis Program Engineering & Licensing Rept | | title = Nonproprietary WCAP-13811, BVPS Unit 1 30% SG Tube Plugging Analysis Program Engineering & Licensing Rept | ||
| author name = Gerlowski T | | author name = Gerlowski T | ||
| author affiliation = WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. | | author affiliation = WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. | ||
Latest revision as of 07:09, 13 November 2023
| ML20056D872 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/31/1993 |
| From: | Gerlowski T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19310D604 | List: |
| References | |
| WCAP-13811, NUDOCS 9308180185 | |
| Download: ML20056D872 (214) | |
Text
{{#Wiki_filter:__. Westinghouse Class 3 WCAP-13811 DUQUESNE LIGHT COMPANY BEAVER VALLEY POWER STATION UNIT 1 30 PERCENT STEAM GENERATOR TUBE PLUGGING ANALYSIS PROGRAM ENGINEERING AND LICENSING REPORT AUGUST 1993 Prepared by: T. J. Gerlowski Approved by: o fd
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Nuclear and Advanced Technolcgy Division P. O. Box 355 Pittsburgh, PA 15230 { I C 19')3, Westinghouse Electric Corporation, all rights reserved. 9308180185 930811 R PDR ~ADOCK'05000334 Ri P: pyn. _ _ , _ . g l
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.l TABLE OF CON'ENTS l
SECTION TITLE' PAGE !
1.0 INTRODUCTION
1-1
'f 1.1 Purpose 1-1 l , 1.2 Background 11 j l.3 Scope 1-3 2.0 POWER CAPABILITY PARAMERIR EVALUATION 2-1 ,
2.1 Power Capability Parameters 2-1 2.2 Best Estimate Steam Generator Tube 2-4 l Plugging Levels -l' 2.2.1 RCS Flow Measurement Uncertainty 2-4 2.2.2 Overall Steam Generator Tube 2-5 Plugging Levels ; 2.2.3 Steam Generator Tube Plugging _2-5 Levels on a Per Loop Basis 3.0 NSSS SYSTEMS EVALUATION 3-1 3.1 NSSS Fluid Systems 3-1 ! i 3.2 Reactor Protection and Reactor 3-1 Control Systems > 3.3 Low Temperature Overpressure Protection System 3-2 [-- 4.0 - NSSS DESIGN TRANSIENTS EVALUATION 4.1 Design Transients 4-1 4-1 1 5.0 N3SS COMPONENTS EVALUATION 5-1 j 5.1 Reactor Vessel 5-1 _ j 5.2 Reactor Internals 5-5 5.3 Reactor Coolant Pumps 5-11 l 5.3.1 Structural Evaluation 5-11 5.3.2 Motor Evaluation 5-11 1 5.4 Control Rod Drive Mechanisms- 5-15 5.5 Pressurizer 5-15 5.6 Reactor Coolant Loop Piping and Primary 5-15 , Equipment Suppons : 5.7 Reactor Coolant Loop Isolation Valves 5-16 l 5.8 Steam Generators 5-16 i 5.8.1 5.8.2 Structural Integrity Evaluation Thermal-Hydraulic and U-Bend Vibration 5-16 5-16
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. Evaluations l 5.8.3 Tube Crush Evaluation 5-16 l 5.9 Auxiliary Equipment 5-17 !
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9 Y TABLE OF CONTENTS (CONT) SECTION TITLE PAGE 6.0 SAFETY ANALYSIS AND EVALUATION 6-1 6.1 Background 6-1 6.2 Licensing Approach and Scope 6-2 6.3 System and Components Evaluations 6-2 6.3.1 NSSS Fluid Systems 6-3 , 6.3.2 Reactor Pressure Vessel System 6-3 6.3.3 Reactor Coolant Pump and RCP Motor 6-4 6.3.4 Main Loop Stop Valves 6 6.3.5 Control Rod Drive Mechanism and Capped Latch Housing 6-5 6.3.6 Pressurizer 6-5 6.3.7 Steam Generators 6-5 6.3.8 NSSS Auxiliary Equipment 6-6 6.4 Non-LOCA Evaluation 6-7 6.4.1 Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal from a Suberitical Condition (USFAR 14.1.1) 6-9 6.4.2 Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal at Power (UFSAR Section 14.1.2) 6-9 6.4.3 Rod Cluster Control Assembly 6-10 s Misalignment (UFSAR Section 14.1.3) 6.4.4 Uncontrolled Boron Dilution (UFSAR Section 14.1.4) 6-1I 6.4.5 Partial Loss of Forced Reactor Coolant Flow (UFSAR Section 14.L5) 6-1l-6.4.6 Startup of an inactive Reactor Coolant Loop (UFSAR Section 14.1.6) 6-13 6.4.7 Loss of External Electrical Load and/or Turbine Trip (UFSAR Section 14.1.7) 6-14 6.4.8 Loss of Normal Feedwater (UFSAR Section 14.1.8) 6-15 6.4.9 Excessive Heat Removal Due to Feedwater System Malfunctions (UFSAR 14.1.9) 6-16 6110 Excessive Load Increase Incident (UFSAR Section 14.1.10) 6-17 6.4.11 Loss of Offsite Power to the Station Auxiliaries o (Station Blackout)(UFSAR Section 14.1.11) 6-18 6.4.12 Accidental Depressurization of the Main Steam System (UFSAR 14.1.13) 6-19 6.4.13 Accidental Depressurization of the Reactor Coolant System (UFSAR 14.1.15) 6-19 6.4.14 Spurious Operation of Safety Injection System at Power (UFSAR 14.1.16) 6-20 WPFIE02D:1Do50693 ij
r TABLE OF CONTENTS (CONT) ; SECTION TITLE PAGE , ~ 6.4.15 Major Secondary Side Pipe Rupture i (UFSAR 14.2.5) 6-21 j 6.4.16 Feedwater System Pipe B.reak (UFSAR Section 14.2.5.2) 6-22 6.4.17 Rupture of a Control Rod Drive Mechanism , Housing (UFSAR Section 14.2.6) 6-22 6.4.18 Reactor Coo; ant Pump Shaft Seizure (Locked Rotor)(UFSAR Section !4.2.7) 6-24 , 6.4.19 Complete Loss of Forced Reactor toolant Flow (UFSAR 14.2.9) 6-25 6.4.20 Single Rod Cluster Control Assembly (RCCA) t Withdrawal at Full Power (USFAR 14.2.10) 6 27 ; 6.4.21 Steam Line Break Mass / Energy Release l Inside/Outside Containment 6-27 ; 6.4.22 Setpoint Impact 6-28 6.4.23 Non-LOCA Conclusion 6-29 : 6.4.24 Non-LOCA References 6-29 , 63 Steam Generator Tube Rupture (SGTR) _6-30 , Evaluation 6.5.1 Unit 1 SGTR Evaluation 6-30 j 6.5.2 Results 6-31
.' 6.5.3 Conclusions 6-31 6.6 Loss of Coolant Accident (LOCA) Evaluation 6-32 6.6.1 Large Break LOCA 6-32 6.6.2 Small Break LOCA 641 ,
6.6.3 Blowdown Reactor Vessel and Loop Forces 6-47 l 6.6.4 Post-LOCA Long-Term Cooling, ! Suberiticality Evaluation 6-48 i 6.6.5 Small and Large Break LOCA PCT Tables 6-49 , 6.6.6 Section 6.6 LOCA References 6-49 6.7 Assessment of Unreviewed Safety Questions 6-57 6.8 Conclusions - 6-59 7.0 NUCLEAR FUEL EVALUATION 7-1 'i 7.1 Core Design 7-1 ) 7.2 7hermal-Hydraulic Design 7-1 7.3 Fuel Rod Performance 7-2 8.0 NSSS/ BALANCE OF PLANT INTERFACE EVALUATION 8-1 ;
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9.0 REFERENCES
9-1 l l APPENDIX A FSAR UPDATE A-1 ! j i I l WPF1802D;1D/0MM3 til j l l
i LIST OF TABLES 0
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- - TABLE TITLE PAGE
,- 2.1-1 Unit 1 NSSS Power Capability Parameters for 2-3 .
Increased S/G Tube Plugging ; 2.2 Unit i RCS Total Flow and Overall 2-7 .* i Steam Generator Tube Plugging Level 2.2-2 Unit 1 RCS Loop Flows and Steam 2-8 ! Generator Tube Plugging Levels ; 5.2-1 Unit i Normalized rms Displacement 5-10 l Amplitudes for Core Barrel and Thermal j Shield Modes 5.3-1 -Unit 1 RCP Motor Loads 5-14 I
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1.0 INTRODUCTION
- 1.1 PURPOSE 'f The purpose of the Steam Generator Tube Plugging Analysis Program is to perform the analyses and i evaluations needed to verify that Beaver Valley Power Station Unit 1 (BVPS-1) is functionally and .f structurally capable of continued reliable and safe operation with: , j
- 1) the tubes in the steam generators plugged to the most limiting plugging level of: a) a maximum plugging level of 30% in any steam generator or b) a plugging level that results j in the reduction of the Reactor Coolant System (RCS) loop flow rate to the Thermal Design !
Flow (TDF) limit, .j
- 2) a reduction in the TDF from 88,500 gpm per loop to 87,200 gpm per loop, and ,
. 3) the incorporation of loop flow asymmetry of up to 5% in the analyses and evaluations in which RCS flow rates are imponant. t r
1.2 BACKGROUND
The TDF limit is a conservatively low design parameter that is used in the thermal-hydraulic design of the RCS and a variety of system, component and safety analyses. The TDF limit is selected to be l conservatively low with respect to the actual RCS flow rate so that flow margin is available over that. assumed in the design and analysis of the Nuclear Steam Supply System (NSSS) and the NSSS components. Once operational data is obtained and the amount of flow margin is known, this margin can be subsequently used to accommodate reductions in RCS flow due to modifications such as steam generator tube plugging. The'TDF flow rate for BVPS Units 1 is 88,500 gpm per loop (265,500 gpm total), and will be reduced to 87,200 gpm per loop (261,600 gpm total). The original thermal-hydraulic design and analysis of systems, components and accidents consistently utilized the TDF design parameter. However, the design and analysis did not always utilize consistent ; assumptions with respect to the level of steam generator tube plugging. Assumptions of the steam generator tube plugging level in the original design and analyses for BVPS Units 1&2 ranged from 0% ! to 5% depending on the analysis. ! WPF1802D:lD051793 11 _j i
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in 1987, Duquesne Light Company performed a Steam Generator Tube Plugging Analysis Program for , 1 Unit I to justify operation with as many as 10% of the tubes plugged in any steam generator, not to % exceed the plugging level that would reduce the RCS flow rate below the TDF limit. The results of. N the Unit I analysis program are presented in WCAP-11591, Beaver Valley Unit 1, "10 Percent Steam i - Generator Tube Plugging Licensing Report" (Reference 1). f in 1990, Duquesne Light Company performed a Steam Generator Tube Plugging Analysis Program to,- increase the allowable level of steam generator tube plugging for Beaver Valley Power Station Units . ; 1&2 (BVPS-1, -2) to a limit of 20%. In general, this tube plugging analysis program was structured ' l to systematically perform the analyses and evaluations to permit the steam generators in Units 1&.2 to be pluggtd to a maximum level of 20% but not higher than the level that preserves the TDF limit on a~ per loop basis. Where possible, common analyses and evaluations were performed to envelop.both units. The scope of this program was limited, however, since the Unit 2 LOCA analyses were [ excluded from the effort. The results of the 1990 program are presented in WCAP-12966, Beaver i Valley Power Station Units 1&2, "20 Percent Steam Generator Tbbe Plugging Program Engineering - and Licensing Report" (Reference 2). ; o In 1992, Duquesne Light Company initiated a program to increase the allowable level of steam generator tube plugging for Unit 1 to 30%, to decrease the TDF for both Units 1&2 to 87,200 gpm per loop, and to incorporate up to 5% loop to loop flow asymmetry in the analyses. As in the 1990 effort, l common analyses and evaluations were performed to envelop both units where possible. In some cases, such as for the LOCA (Loss-of-Coolant-Accident) analyses, separate analyses and evaluations were performed for each unit As part of the effort started in 1992, Westinghouse performed { evaluations for the reduced TDF for each unit (Reference 3 for Unit 1, Reference 4 for Unit 2). These j evaluations were used by DLC to assist in preparation of a combined (Unit 1/ Unit 2) request for a Technical Specification change to reflect the lower TDF value (Reference 5). ; 7 This report provides the results of this program as they pertain to Unit I and identifies whether the o analysis or evaluation is common to both units or separate. , 1.3 SCOPE , Westinghouse scope for this program includes the NSSS safety, system and component analyses and l evaluations necessary to increase the allowable steam generator tube plugging level for BVPS-1 to the -l r wPFl*02D:lD051?93 1-2 ; i
i {
' 30% limit. The program encompasses all aspects of the NSSS design and operation which are.
impacted by the increased steam generator tube plugging, reduced TDF limit, and loop to loop RCS l M flow asymmetry, 'Ihe scope of the program includes the NSSS safety analyses, the functional l 1 capability of the systems for normal and abnormal plant operations, and the mechanical design of the 1
.. NSSS components and structures. A detailed technical description of the integrated (Units 1&2) )
Steam Generator Tube Plugging Analysis Program may be found in Westinghouse letter DLW-92-147, j (Reference 3), which contains the program technical description, a list of deliverables, and program , commercial information. ! Certain analyses and evaluati~1s are excluded from the Westinghouse scope of this program. Among those are: r 1, The Post-LOCA containment mass and energy release analysis; l
- 2. The analysis of Post-LOCA hot leg switchover time to prevent potential boron precipitation in j
- the reactor vessel following boiling in the reactor core; and b
- 3. Analyses and evaluations of Balance of Plant systems. To assist in this area, Westinghouse scope includes an evaluation of BOP system interfaces. j The analyses and evaluations included in this program were performed in accordance with the l following criteria. -;
I. Safety analyses included in the program were performed to FSAR quality standards, using i current NRC approved analytical techniques, and were evaluated in accordance with the criteria and standards that apply to the current BVPS Unit 1 operating license. , r
- 2. NSSS system and component designs were evaluated in accordance with the regulatory. ;
l s, ' requirements, codes, and standards which were applicable to BVPS Unit 1 when it was , 1
- l; originally licensed plus any subsequent criteria specifically applied to BVPS Unit I by the - ;
!. - NRC. i i ! I wPriso2n;1D/051793 13 .
f I ,.5.~ The steam generator tube plugging increase program for Beaver Valley Unit I was based on the following assumptions and design inputs: The analyses and evaluations assumed a reactor coolant system TDF of 87,200 ppm per loop and a maximum flow asymmetry of 5%. The analyses assumed a steam generator tube plugging level of 30%. , l t NSSS performance parameters were determined which minimized the impact of the tube .i plugging allowance on steam pressure beyond the steam pressure reduction accounted for in WCAP-12966. .
+
t
- The analyses and evaluations were based on the fuel loading anticipated for Cycle 10. (275 psig non-IFBA Vantage 5 Hybrid without IFMs, (fresh and burnt),1500 psig once-burnt center assembly STD, and 100 psig and 200 psig IFBA V5H (burnt and fresh). Future cycles l t
.. are expected to insert only fresh V5H (non-IFBA and IFBA) except for one 500 psig STD center assembly reclaimed from the spent fuel pit.)
- The analyses assumed that the thimble plugs in the reactor internals are removed. .I 3
-l
- The analyses were based on the industry and regulatory codes and standards which were applicable to Beaver \ alley Unit I when the plant was originally licensed plus any subsequent i criteria specifically applied to Beaver Valley by the NRC. ,
The safety analysis methodology used was consistent with the current evaluation model methods (i.e., BASH and NOTRUMP). ; The TDF Reduction Repon addressed the reduction in TDF of 1.5% from 265,500 gpm to , e 261,600 gpm and its attendant effect on the "DNB PARAMETERS TABLE 3.2-1" in the Technical Specifications. The revised design parameters and the results of the evaluations and analyses performed during the program are presented in the following sections. WPF1802D:lDO51793 ]$
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I 2.0 POWER CAPAHILITY PARAMETER EVALUATION i 1 m' 2.1 POWER CAPAlllLITY PARAMETERS - BVPS Unit 1 is designed and licensed to operate at an NSSS power level of 2660 MWt with a TDF of i 88,500 gpm per loop (265,500 gpm total flow). The original NSSS power capability parameters j associated with this design condition are shown in the first column of Table 2.1-1. , The power capability parameters for Unit I were revised in 1987 as part of the analysis program performed to permit steam generator tube plugging to the 10% level in combination with the design , t modification of the reactor internals to convert to upflow an:1 to remove the thimble plugs. De power capability parameters resulting from this analysis are shown in the second column of Table 2.1-1. '{ Comparison of the revised power capability parameters with the original parameters shows that the { only parameters to change were the steam generator (secondary side) parameters. Steam pressure, steam temperature and steam flow decreased as a result of the decrease in steam generator heat transfer j
. surface area that accompanied the increase in the allowable steam generator tube plugging level to the .;
10% limit. The power capability parameters for Unit I were further revised in 1990 as part of the analysis program performed to permit steam generator tube plugging to the 20% limit, or to the : plugging level that preserves the TDF limit on a per loop basis. The power capability parameters resulting from the 1990 analysis are shown in the third column of Table 2.1-1. As with the power ; capability parameters for the 10% plugging limit, this table shows that the parameters are the same , with the exception of steam generator steam pressure, temperature and flow which continue to decrease as a result of the continued decrease in heat transfer area that accompanies increased plugging levels. These parameters supported retention of the existing design transients for both Units 1&2. ; Consequently, the final parameters from the 1990 effort permitted the plugging of tubes in any steam 'l generator up to either the 20% limit or to the level that maintains the TDF limit, whichever is most j limiting. j
.l . The power capability parameters used in this program reflect the level of 30% steam generator tube =
plugging, a thermal design flowrate reduction of 1.5%, and loop-to-loop flow asymmetry of up to 5%. These parsmeters are shown in the fourth column of Table 2.1-1. Since the RCS thermal design - l
- )
flowrate was reduced, the RCS temperatures reflect the change in heat load on the reactor coolant.. T increased, and T decreased, slightly due to the lower RCS flow through the coce. T,,' however, . remained the same. The steam temperature, pressure, and flowrate also decreased due' to the increased WPF1802D:IDt051793 2-1 l l
t i i i
- i steam generator tube plugging, and its attendant reduction in heat transfer area. As noted in j Table'2.1-1, however, the steam generator condidons were assumed to remain the same as the j
- . parameters for the 20% SGTP case for analysis considerations. Based on current plant operating conditions, there is expected to h operating margin to accommodate the anticipated steam pressure )
i
. . reduction caused by the higher plugging level. Since a steam pressure reduction below 760 psia would ] i effect a large change in the design transients necessary to envelop the design, the decision was made j to ' analytically' limit the steam pressure reduction to 760 psia and use actual plant operating margin to !
i maintain steam pressure above the 760 psia limit. Consequently, the analyzed parameters now cover ,j
- up to 30% steam generator tube plugging, reduced thermal design flow (87,200 gpm per loop), and j -
loop-to-loop flow asymmetry of up to 5%. '!
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i , . .. . . TAllt,E 2.1-1 IIEAVER VAI,1,EY POWER STATION - Unit i NSSS Power Capability Parameters for increased S/G Tube Plugging 10 % 20% 30 % Original S/G Tube S/G Tube S/G Tube Design Plugging Plugging, Plugging Conditions TDF Maintained TDF Maintained TDF Reduced NSSS hwer, MWt 26M) 2660 26N) 2660 Reactor Power, MWt . 2652 2652 2652 2652 Thermal Design Flow, gpm/ loop 88,500 88,500 88.500 h7,2tX) Reactor Coolant Pressure, psia 2250 2250 2250 2250 Reactor Coolant Temperature, "F Vessel Outlet 609.9 609.9 609.9 610.4 Vessel Average 576.2 576.2 576.2 576.2 Vessel / Core Inlet 542.5 542.5 542.5 542.0 Steam Generator Steam Temperature *F 516.8 512.3 512.4 506.5 (1) Steam Pressure, psia 790 760 760 721 (I) Steam Flow, lif Ib/hr Total Il.61 11.60 l1.60 11.59 Tinha Plugging, % 0 10 20 30 Zero Load Temperature, 'F 547.0 547.0 547.0 547.0 Core Bypass Flow, % 4.5 6.5 6.5 6.5 Footnote: (1) Steam conditions are limited to minimums of 760 psia and 512.3 'F due to component design transient conditions. (2) Analyses using these parameters also include consideration of up to 5% loop flow asymmetry. WPFl802D:llM)$l?93 2-3 - _ _____. _ ._ __ . _ ._ . -. . ...._. . . . . _ . . . . . . . . . . ~ . - . _ . _ . . - . . _ , . . . _ - _ _ _ _ _ . . . _ . . . _ . . . . . . . . - . - - . _ _ , . - . . . .-.._....._.~.w2.__ _ . -
l 2.2 IIEST ESTIMATE STEAM GENERATOR TUBE PLUGGING LEVELS ;
'o The BVPS Unit 1 Steam Generator Tube Plugging Analysis Program has been stmetured to permit the steam generators to be plugged to the 30% limit or to the plugging level that maintains RCS flow at l . the reduced 'Ihermal Design Flow (TDF) limit of 87.200 gpm per loop (261,600 gpm total flow), !
f including consideration of flow measurement uncenainty as applicable Also included in the program is the allowance of loop-to-loop flow asymmetry of up to 5%, which is defined as follows. The . . { flowrate in any kiop must be greater than 82,840 gpm (54 below 87.200 gpm); the sum of the ; flowrates in the two lowest flow loops must be greater than 170,040 gpm; and the sum of all three l loop flowrates must be equal to or greater than the reduced TDF limit of 261,600 gpm. The various ! i system, component and safety analyses assume either the minimum loop flow of 87.200 gpm (or the l flow of 82,840 considering asymmetry) or the minimum total flow of 261,600 gpm depending on l
}
whether they are k>op-related or system-related evaluations. This program structure requires that the ; e TDF limit be met on both a loop flow and a total flow basis. Consequently, total flow cannot be permitted to decrease below the TDF limit of 261,600 gpm and loop flow cannot be permitted to j
) . decrease below the TDF limit of 87,200 gpm (or 82,840 gpm considering asymmetry), including l t
margin to accommodate flow measurement uncenainty. Steam generators may be plugged based on l 1 the flow margin available in their respective loops but may not utilize flow margin available in other - loops. j l t 2.2.1 RCS Flow Measurement Uncertainty
- t i
Technical Specification 3.2.5 requires that actur.1 RCS total flow be greater than the Thermal Design l Flowrate. Recently, plant-specific calculations of flow measurement uncenainty were performed by Westinghouse. (Reference 6 letter DLW-92-140,9/23/92, Simmers to Halliday,"RCS Flow Uncenainty Calculation") These calculations indicated that the flow measurement uncenainty is 1.95% t for Unit 1. Experience in revising the subject Technical Specification on other plants has shown that the NRC recommends a flow measurement penalty for feedwater venturi fouling. The net effect of l
. this consideration would increase the calculated flow measuremer uncenainty by approximately 0.1%.
For the purpose of this report, the value of 2.05% for the uncertainty will be used for both total and loop flowrates, j i WPFIE02D:tD/051793 2-4 ! l
1 1 I 2.2.2 Overall Steam Generator Tube Pluccine Levels ; i The prediction of best estimate overall steam generator tube plugging levels for BVPS Unit 1 considers both the existing plugging levels plus the additional plugging levels that can be acideved ! .. based on the present RCS flow margin. The results of the most recent flow measurement tests show l that total flow for Unit I was measured as 281,271 gpm (Reference 7). The overall steam generator j I tube plugging levels that existed for Unit I at the time of these tests was 10.4% (Reference 8). The - , most recent RCS flow measurement test results and the existing steam generator tube plugging levels l at the time of these tests are summarized on Table 2.2-1. To obtain percent flow margin, the flow margin is divided by the TDF flow limit. To obtain the best l esumate plugging level, several iterations of the NSSS code were made to indicate the plugging level j in each generator which resulted in a loop flowrate equal to the TDF value plus 2.05% uncertainty. -l Re tube plugging levels calculated in this manner based on existing flow margin are termed the best estimate final tube plugging levels. f Based on the measurement uncertainty of 2.05%, the best estimate final plugging level is calculated to ! be 273% for Unit 1. The calculation of this best estimate additional plugging level is summarized in Table 2.2-1. Best estimate additional plugging levels for Unit 1 would be 16.9%, also based on a j 2.05% uncertainty. j De best estimate calculation of overall steam generator tube plugging levels is based on overall flow j margin with respect to the TDF of 261,6'O gpm total flow. Since the assumptions of this program f also require that the TDF of 87,200 gpra be maintained on a per loop basis, best estimate plugging j levels on a per loop basis can be calculated based on the flow margin in the respective loops. ; I i 2.23 Steam Generator Tube Pluccine Levels on a Per Loop Hasis f t
'i . As discussed in Section 2.2 the Unit I steam generators are permined to be plugged to the 30% limit i
or to the plugging level that reduces RCS flow to the TDF limit of 87,200 gpm per loop (or to } 82,840 gpm per loop considering asymmetry), including consideration of flow measurement ] uncertainty as applicable. Consequently, the best estimate prediction of tube plugging levels for each f i steam generator is based on the plugging levels that exist in each steam generator plus the best ; estimate prediction of additional plugging that can be achieved based on the present RCS flow margin 6 wmm:n.inenn 2-5 i l
in the associated loop. The results of the most recent Dow measurement tests on a per loop basis, a summary of the existing steam generator tube plugging levels per steam generator, and a calculation of
;* best estimate additional plugging levels for each steam generator based on present RCS Dow margin .
are presented in Table 2.2-2. e
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TABLE 2.2-1 i
- BEAVER VALLEY POWER STATION i
Unit 1 , 4 RCS Total Flow and , Overall Steam Generator Tube Plugging Level , i l Measured Row, gpm 281,271 (1) i Overall Steam Generator Tube Plugging Level. 1,054 f Number of Tubes (1) ; i L Overall Steam Generator Tube Plugging Level, % 10.49c ! + (1) t Minimum Measured Flow, gpm 266,963 _; i (2) l RCS Total Flow Margin, gpm 14,308 : i RCS Total Flow Margin, % 5.5 Best Estimate Final Tube Plugging Level, % 27.3 Best Estimate Additio'nal Tbbe Plugging Level, % 16.9 l - Notes: (1) Tube plugging level and measured flow after outage 8R. (2) Flow measurement uncertainty = 2.05% WPF1802D:1DOSI793 ~2 7 j i
( TABLE 2.2 2 BEAVER VALLEY POWER STATION Unit i RCS Loop Flows and Steam Generator Tube Plugging Levels , Measured Flow, gpm 93,193 93,962 94.116 (1) (1) (1) Steam Generator Tube Plugging Level, 508 301 245 Number of Tubes (1) (1) (1) Steam Generator Tube Plugging Level, % 15.0 8.9 7.2
. (1) (1) (1)
Minimum Measured Flow, gpm 88,988 88,988 88,988 (2) (2) (2) Flow Margin, gpm 4205 4974 5128 Flow Margin, % 4.8 5.7 5.9 Best Estimate Final Plugging Level, % 29 27 26 Best Estimate Additional Plug'ging Level, c4 14 18.1 18.8
. . Notes:
(1) Tube plugging level and measured flow after outage BR. (2) Flow measurement uncenainty = 2.05%. WPF1802D:lD21793 2-8
I 3.0 NSSS SYSTEMS EVALUATION 3.1 NSSS FLUID SYSTEMS g As shown in Table 2.1-1 the increase in allowable steam generator tube plugging levels to either the 309 limit or the new TDF limit, or the loop-to-loop flow asymmetry of .5% does not result in ; significant changes to the power capability parameters for NSSS power, reactor power, RCS flow, , RCS pressure and RCS temperatures (i.e.. Tw oT w, T,3 and T,%). The design parameters for RCS flow, pressure and temperature are major parameters with respect to the design of the RCS and the NSSS auxiliary systems such as the Chemical and Volume Control System, the Safety injection System and the Residual Heat Removal System. Since these major parameters do not change signifiert.tly due to the increase in allowable steam generator tube plugging level or the TDF reduction { or loop-to-loop flow asymmetry, increased tube plugging to either the 30% or the revised TDF limit . does not affect the design of the RCS or the NSSS auxiliary systems that service the RCS. l b . The steam generator parameters for steam temperature, pressure and flow do change as a result of the .! increase in allowable tube plugging level. The Ange to these parameters does not affect NSSS j systems design. An evaluation of the impact of these changes on the performance of the steam ; generators and the design of the Balance of Plant systems is presented in Sections 5.8 and 8.0, respectively. : e 3.2 REACTOR PROTECTION AND REACTOR CONTROL SYSTEMS
.i The increase in allowable steam generator tube plugging level to either the 30% limit or that widch ;
results in RCS flow approaching the TDF limit or the loop to-loop flow asymmetry condition do not necessitate any changes to the reactor protection system or the reactor control system setpoints. j The performance of the various protection systems . vill remain within their design envelopes and will i
.- not be materially affected. Revision to the protection system setpoints is not required since there are no changes to the RCS operating parameters of thermal design flow, delta temperature and average temperature. See Section 6.0 for the safety analysis and evaluation of design basis transients and i accidents which confirm that changes are not required to protection system setpoints to satisfy safety j analysis acceptance criteria.
wPFIs02D:lDo$1793 3-1 ; l
C e Based on an overall review of the reactor control system operation and experience on plants that have performed tube plugging, the performance of the various control systems will remain within their design envelopes and will not be materially a.iected. Dere will be no loss !a plant operability or degradation in the plant response to normal expected transients. Th- response of level in the steam a generators to unequal thermal loads associated with load swings will not be materially affected and the potential for steam generator level trips shoul<t not be increased as a result of the increar.e in allowable steam generator tube plugging level. , 3.3 LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEM j De effect of increased SGTP is to reduce the volume of the RCS and to reduce the heat trans'fer surface area in the steam generators. The effects that these changes have on the design basis heat injection and mass injection transients were evaluated to determine the continued validity of the PORV setpoints utilized in the Lnw Temperature Overpressure Protection System (LTOPS). A brief , description of the analyses condxted ' 'I the heat injection and mass injection transients is described
- below. ,
For the mass injection transient, the basic relationship for setpoint/ overshoot determination is: AP = AP(ref)*F(v)*F(s)*F(z) where , 1 AP = se: ,nt overshoot ; AP(ref) = mass injection rate fuuion for a reference overshoot ! F(v) = RCS volume fuor . F(s) = relief valve setpoint factor , F(z) = relief valve opening time factor 5 Only the F(v) term will be significantly affected by reduced SG tube volume resulting from increased
- SGTP. Smaller RCS volumes result in greater pressure increases for the mass injection transient.
Given that the overshoot is equal to the Appendix G pressure limit minus the PORV setpoint (AP = Po - SP), with a higher overshoot and a constant Appendix G limit,it fcellows then that the new PORV setpoint must be lower. The new PORV setpoint is equal to the initial setpoint minus the' 1 initial overshoot adjusted for the RCS volume change. Using the worst case large setpoint overshoot ; l (80.9 psi) combined with the relatively small setpoint (421.3 psi) at 48 EFPY as the limiting case, the l wermerr19/05i793 32 ~ E
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maximum reduction in the LTOPS setpoints is calculated to be 1.31%, or approximately 5.5 psi. l Ilowever, considering that the PORY LTOPS setpoint has already been reduced by 4 psi to account -
- . for the previously analyzed (Reference WCAP-12966) 20% SGTP conditions leaves only a maximum l of approximately 1.5 psi to be accounted for as the effect of increasing to 30% SGTP from 20%
.- SGTP. Dis difference is considered to be sufficiently small so as to be absorbed within the conservatism inherent in the metimdology used to generate the setpoints. Derefore, further adjustment - c' the PORV LTOP setpoints to account for the effect of 30% SG'IP should not be necessary in order ' to maintain the appropriate level of protection against reactor vessel brittle fracture. he cor. sequence of increasing the steam generator tube plugging level, with respect to heat injection events, depends on the rate of reduction of heat transfer surface area relative to the rate of reduction of the RCS volume. De effect of plugging steam generator .Nbes is to always result in reduced PORV . setpoint overshoot (system gressure minus PORV setpoint) due to heat injection events. Therefore, the setpoint adjustment for the mast injection transient developed above is bounding. ; . -j t f a
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4.0 NSSS DESIGN TRANSIENTS EVAI,UATION 4.1 DESIGN TRANSIENTS o Based on the power capability parameters presented in Section 2.1, the currently applicable Beaver Valley Unit I design transients which are based on those presented in Systems Standard Design j Criteria (SSDC) 1.3, Revision 1 (Reference 7) and Revision 2 (Reference 8) remain bounding for , increased steam generator tube plugging levels up to either the 30% or the revised TDF limit. The evaluation and discussion of design transients which are in Section 4.0 of WCAP-12966, continue to - , apply for the 30% SGTP, reduced TDF case. That same evaluation, which also addresses the cold ; I overpressure and auxiliary design transients, remains valid. This conclusion is valid only to the minimum steam conditions of 760 psia and 512.3'F due to component design considerations. , Operation at steam conditions lower than these values will require additional analyses to confirm component integrity. t I
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l i l l 1 i l 5.6 NSSS COMPONENTS EVALIIATION , l l 5.1 REACTOR VESSEL i i
- This section evaluates the maximum primary plus secondary stress intensity ranges and maximum }
cumulative fatigue usage factors resulting from the revised operating parameters of 30 percerit steam l generator tube plugging with reduced thermal design flow. Additionally, the effects of operating , temperatures near vessel outlet temperature (Ts) in the upper head region were evaluated. j
'I The original design basis assumed that the internal surfaces of the closure head are in contact with j inlet water. Tids origi. o assumption is not conservative for the main closure flange assembly and the CRDM housings. Therefore, the higher temperature effects were also included herein. The reactor vessel analytical report including the previous addenda and this addendum report indicates that operation of the reactor vessel in accordance with the conditions associated with 30 percent steam f generator tube plugging (SG1T>) and reactor coolant system thermal design flowrates (TDF) of 87,200 f - gpm does not re.sult in stress intensities nor fatigue usage factors which exceed the acceptance criteria of the 1968 Edition of Section til of the ASME Iloiler and Pressure Vessel Code through the Winter [
i 1968 Addenda. j lhe design and operating parameters for the reactor vessel were revised as a result of 30 percent f i SGTP and reduced TDF as follows: l r Design Pressure 2485 psig i t Normal Operating Pressute 2235 psig { Design Temperature 650*F 542.0'F I Normal Operating inlet Temperature I Normal Operating Outlet Temperature 610.4*F l l (610.9'F conservatively used) ; I a Zero Load Temperaturc $47'F j
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The reactor vessel design transients and design external loads are assumed to be unchanged with the c exception of the Plant Loading and Unloading transients which are affected by the changes in the i i normal operating temperatures. ;
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i l Table 2.1 1 identifies revised normal operating parameters for the Beaver Valley Unit No. I reactor coolant system (RCS). De revised parameters are applicable for a 30 percent steam generator tube : r - plugging level with the thermal design flowrate reduced to 87,200 rpm. The resulting reactor vessel normal operating outlet temperature is within the range of temperatures for which the reactor vessel . was previously analyred in the analytical report (Reference 12) and Addenda (References 13,14, and 15). De vessel inlet temperature decreased from 542.5'F to 542.0*F m the power capability parameters of Table 2.1-1, and this temperature is below 543.5*F for which the reactor vessel is
- t analyzed in References 12 through 14 and also below 542.5'F which is considered in Reference 15. ,
herefore, the regions of the reactor vessel which are in contact with vessel inlet water during normal reactor operation are subject to the evaluations in this report. No evaluation of the reactor vessel ; outlet nozzles is required. ! Reference 16 specifies that only the reactor vessel outlet nozzles are in contact with vessel outlet water while the balance of the reactor vessel and closure head internal surfaces are in contact with vessel inlet water, flowever, a subsequent study, documented in Reference 17, shows that the closure head . internal surfaces are in contxt with reactor coolant at temperatures near vessel outlet temperature. As ; a result, the design basis assumption of vessel inlet water in the closure head is not conservative. Therefore, evaluations were performed to justify operation of the reactor vessel closure, flange . i assembly and CRDM housings (head adapters) at the higher temperatures. Dese evaluations along i with the evaluations of other regions for the normal operating vessel inlet temperature change are ., r documented in this report. ; i i The reactor vessel main closure flange assembly and CRDM housings were evaluated for a reactor . coolant temperature of 598.5*F in the closure head. De upper head temperature is 81.9 percent of the difference between an inlet water temperature of 542.0*F and the revised outlet water temperature of l 610.9'F. De only design transients which were affected by the increased upper head temperature are ; the Plant 1.oading and Unloading transients. Since the upper head region does not directly -
- communicate with the outlet plenum, the cold leg transient temperature variations with which the :
J
. closure head sections we originally analyzed are assumed adequate, liowever, the steady state '
effects of the higher temperature were accounted for by scaling the steady state thermal stresses for the lower temperature. The Plant Loading and Unloading stress intensities are obtained by either directly calculating the " skin effect" stresses or by scaling the existing transient thermal stress information. De maximum ranges of primary plus secondary stress intensity and the maximum cumulative fatigue usage factors for the closure flanges, closure studs and CRDM housings are updated accordantly, wPF180 D:1Dm51793 52
l 1 i i The small change in the vessel inlet temperature from 543.5*F to 542.0*F was viewed as affecting only the Plant Loading and Unloading transients. Reference 19 indicates that the other design
- - transients remain bounding and are, therefore, unchanged. De changes in the stress intensities for the T ,w regions due to Loading and Unloading are obtained by scaling the stress intensities from ,
. Reference 12 using a factor of 1.429 obtained from the ratio of the temperature changes from zero L
load temperature to the respective vessel inlet temperatures. The maximum fatigue usage factors for the T,,,m regions were then revised in accordance with the calculated stress intensities. , The regions of the reactor vessel which were assumed to be in contact with vessel inlet water during normal operation are outside the bounds of the previous thermal and stress analyses in References 12 through 15. He 542.0*F vessel inlet temperature associated with 30 percent SGTP and reduced TDF is less than the 543.5'F Ta considered in the analytical report. Derefore, the maximum ranges of , primary plus secondary stress intensity and cumulative fatigue usage factors for the To regions were re-evaluated, and the evaluations are documented in calculation PCE-93-0016 in Appendix A. The regions include the inlet nozzles, vessel wall transition, bottom head-to-shell juncture, core support
. guides and bottom head instrumentation tubes. De evaluations indicate that the maximum ranges of stress intensity and maximum cumulative usage factors reported for these regions in References 12 ;
through 15 are unchanged by 30 percent SGTP and reduced TDF. ; i The outlet nonles which operate at vessel outlet temperature are within the bounds of the previous , thermal and stress analyses in References 12 through 15. Berefore, the outlet nozzle maximum range of stress intensity and maximum cumulative usage factor are unchanged. De reactor vessel main closure flange assembly and CRDM housings were evaluated to account for higher operating temperatures in the closure head in addition to the 30 percent tube plugging and reduced thermal design flowrate effects. Reference 17, Table 4-4 shows that the measured mean upper { head temperature in the J. M.'Farley reactor vessel head is 81.8 percent of the difference between Ta i and Tw. This closely agrees with the predicted value of 81.9 percent of the difference. Since Beaver j
. Valley I has a similar reactor vessel, reactor internals and core bypass flow as J. M. Farley, die 81.9 1 percent predicted mean upper head temperature was used in calculation PCE-93-0016 in Appendix A. l Assuming T is $42.0*F and Tw is 610.9 F, an upper head temperature of 598.5'F is obtained. His j
- i i
temperature was used in the evaluations of the flange assembly and CRDM housings. Consideration of the higher closure head temperature is responsible for the significant increases in the maximum i WPFikO2D;1Dx151793 5-3
cumulative usage factors as opposed to 30 percent tubes plugging / reduced TDF. De results of the evaluations performed for the upper head region are summarized as follows: Slain Closure Flange Assembly The maximum range of stress intensity in the closure flanges occurs in the vessel Dange. The range reported for the vessel flange in Reference 12 was found to be unaffected by the 30 percent SGTP and reduced TDF. Additionally, the increased upper head operating temperature does not change the maximum range of stress intensity which is attributable to the heatup and cooldown transients. The maximum range of stress intensity in the vessel flange remains 73.1 ksi which is less than 3 Sm of 80.1 ksi. The maximum service stress in the closure studs is also unchanged at 104.44 ksi which is less than 3 Sm at 116.1 ksi. The maximum cumulative usage factor in the Dange assembly occurs in a closure stud at the bottom of the stud shank. The revised usage factor was calculated to be 0.9800 which is much greater than 0.7864 from Reference 12, but remains less than the 1.0 allowable. The increase in the closure stud usage factor is due to the increase in the assumed upper head temperatures. The effect of 30 percent SGTT' and reduced TDF is insignificant. CRD51 Housings The maximum range of primary plus secondary stress intensity for the CRDh1 housings is 50.65 ksi as a result of the increased upper head temperature. This value is less than the 3Sm allowable of 69.9 ksi. The maximum cumulative fatigue usage factor for the CRDh1 housings is reported to be 0.5688 in Reference 2. Howe'ver, review of the usage factor calculation revealeo conservatism in this value. A less conservative calculation which included the effects of the increased closure head
. operating temperature yielded a maximum cumulative usage factor of only 0.0972.
5.2 REACTOR INTERNAIS N wPflho:D.!Do5 t ?93 5-4 \ ^
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i 1 5.3 REACTOR COOLANT PUMPS i 5.3.1 Structural Evnluation De structural design of the reactor coolant pumps utilites as design inputs the power capability . parameters for RCS normal operating temperatures (i.e., RCS hot leg temperature and cold leg temperature) and the NSSS design transients. As discussed in Section 2.0 and 4.0 the RCS normal ; operating parameters and design transients do not change as a result of the increase in steam generator a tube plugging level to either the 30% or the revised TDF limit. The reactor coolant pumps are also { exgosed to higher resistance due to the increase in allowable steam generator tube plugging level. De ~
. design and analyses of the reactor coolant pumps remain applicable for tube plugging levels up to i cither the 30% or the revised TDF limit, ne hydraulic loads associated with operating at the higher resistance do not have any adverse effect on the pump internals. I i
f 5.3.2 Motor Evaluntion Revised loads have been calculated for the Beaver Valley Unit I reactor coolant pump motors as shown on Table 5.3-1. De new loads have increased due to both revised performance estimates (RCS temperatures) and the effects of the proposed tube plugging (best estimate RCS flowrate). he ability of the RCP motors to operate with these new loads is as follows
- 1. Continuous operation 'at a new hot loon load of 6250 HP. -%is represents a 4.2% increase j over the nameplate rating of the notor. Original test data for this design indicates that there is ;
. sufficient thermal margin to allow for operation with this revised load without exceeding i
National Electrical Manufacturers Association (NEMA) temperature rise limits (NEMA allowable rise is 75*C). !
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- 2. Continuous oxration at a new cold loon load of 8030 HP. His represents a 7.1% increase over the nameplate rating of the motor. Original test data for this design indicates that there is ;
*~ sufficient thermal margin to allow for operation with this revised load without exceeding NEMA temperature rise limits (NEMA allowable rise is 10(PC).
- 3. Rotor windine temperatures durine worst case startine scenario with revised loads. The worst case starting scenario is a cold loop,80% voltage stan with full back flow. For this condition, the calculated rotor winding temperature rises exceed the design allowance of 30(P C for bars and 50* C for rings and are therefore, unacceptable. De consequence of exceeding the design allowances for rotor winding temperature rise is an accelerated rate of mechanical aging (fatigue) which could result in a rotor winding failure before the 40 year operating period of the motor has been achieved. It should be noted that failure in this case means a crack developing in the resistance ring which would, if not corrected, eventually cause a failure of '
the motor to start. There is no impact on the safety-related function of the motor (i.e., , coastdown). Although the potential failure of an RCP motor to start on Unit I is not a safety concern, it is an operational concern. Several possible solutions exist to alleviate the operational concern, including a change to the minimum starting voltage, a more detailed finite element analysis, rotor modification or additional operational precautions and limitations. These possible solutions are outlined below: i) Minimum Startine Voltace he rotor windings would meet the design limits for , temperature rise if the minimum allowable starting voltage were raised to 89% (it is currently 80%). If the actual voltage drop at the plant is not greater than 11%, l then the motors are acceptable as is and nothing else need be done. > ii) Finite Element Analysis A more detailed finite element type analysis could be
- performed which would yield an estimate of the starting cycles to failure for the rotor. His analysis would give specific information~ but may not change the ;
result. 1 iii) Rotor Modification The rotor could be reworked to make it the same as the Unit 2 rotor. I wPF1802D:lD!051793 ~ 5-12 l
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I i i t iv) Ooerational Precautions and Limitations Alternatively, the choice could be made j to do nothing and allow the Unit 1 motors to operate as is. If this option is i a chosen, the following precautions are strongly recommended:
- I
. a) The starting duty would be ch: aged to allow NO CONSECUTIVE STARTS. A cooling period of 20 minutes running or 45 minutes standing l idle would be required before the next start attempt. .
b) The rotor should be inspected visually for cracks in the resistance rings and braic joints at a nunimum interval of every 5 years if cracks were found, an immediate repair would be recommended. l P The RCP motor evaluation v .s based on an RCS best estimate flow rate of 89,100 gpm per loop (hot) l as shown on Table 53-1. The increase in loads due to temperature changes (associated with a -l reduction of RCS flow rate to the 87,200 gpm per loop revised TDF limit) was also evaluated. It was
- determined that this increase in loads is not large enough to change the results of the analysis for the j 89.100 gpm per loop best estimate case, t
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l TABLE 5.3-1 l BEAVER VALLEY POWER STATION . .' Unit ! ' P RCP Motor Loads ; 07c Pluccing 309c Pluccing Cold - Ilot Cold ' Hot - l Flow, gpm 91090 96900 86760 89100 BHP 7900 6050 8030 6250 l Torque: , e ft-lbs 35000 26730 35611 27626 I i _ NOTES: A cold flow rate equal to 94% of the hot flow rate, was assumed. Impeller hydraulic thrust is as given in the motor specifications.
'i The above BHP / Torque values are higher than those in the motor specifications, and are from revised {
i estimates. ; i 4 i werisozo:inc51793 5-14 1
i 5.4 CONTROL ROD DRIVE MECIIANISMS r - The design of the control rod drive mechanisms utilizes as design inputs the power capability j parameters for RCS hot leg temperature (Tw) and the NSSS design transients. As discussed in Section 2.0 and 4.0, the RCS normal operating parameters and design transients do not change as a result of the increase in steam generator tube plugging level to either the 30% or the revised TDF limit. Derefore, the design basis analysis and evaluation of the BVPS Unit I control rod drive mechanisms remain applicable for tube plugging levels up to either the 30% or the revised TDF limit. 5.5 PRESSURIZER Key parameters in the design of the pressurizer are the RCS hot leg temperature (Tw), the RCS cold , leg temperature (T,), the RCS no-load temperature and the pressurizer design transients. Based on the design parameters presented in Section 2.0 and the conclusion that there are no changes to the ; design transients of record which were used in the original analysis, the design of the pressurizer was evaluated via Reference 12. De evaluation Showed that there were no significant effects on the component stresses and fatigue analysis as presented in the original stress reports. De Unit I and 2 pressurizer components meet the ASME Code Section 111 stress analysis and fatigue analysis requirements for operation with allowable steam generator tube plugging levels up to either the 30% or the revised TDF limit. 5.6 REACTOR COOLANT LOOP PIPING AND PRIMARY EQUIPMENT SUPPORTS The design of the reactor coolant loop piping, the primary equipment supports and the primary equipment nozzles utilizes as design inputs the power capability parameters for RCS normal operating temperatures (i.e., RCS hot leg temperature and cold leg temperature), the NSSS design transients, and- l the LOCA hydraulic forcing f' unctions. As discussed in Section 2.0 and 4.0, die RCS normal .i operating parameters and design transients do not change as a result of the increase in steam generator fi .- tube plugging level to either the 30% or the revised TDF limit. Also as discussed in Section 6.8, the -; LOCA hydraulic forces do not change due to the increase in tube plugging level. Therefore, the l design basis analysis and evaluation of the BVPS Unit I reactor coolant loop piping, primary ' equipment supports and primary equipment nozzles remain applicable for tube plugging levels up to - - l
-l either the 30% or the revised TDF limit.
5-15 i WPFIs02D:1D/051793 . t i
5.7 REACTOR COOLANT LOOP ISOLATION VALVES
- Re design of the reactor coolant loop isolation valves utilizes as design inputs the power capability parameters for RCS normal operating temperatures (i.e., RCS hot leg temperature and cold leg
. temperature) and the NSSS design transients. As discussed in Section 2.0 and 4.0, the RCS normal operating parameters and design transients do not change as a result of the increase in steam generator tube plugging level to either the 30% or the revised TDF limit. Therefore, the design basis analysis and evaluation of the BVPS Unit I reactor coolant loop isolation valves remain applicable for tube plugging levels up to either the 30% or the revised TDF limit.
1 I 5.8 STEAM GENERATORS 5.8.1 Structural Intecrity Evaluation f l j De structural integrity evaluation of the steam generators performed in Referena 2 was determined to i
- i. remain applicable for the 30% plugging conditions. In summary, the results of the structural evaluation for the 30% tube plugging conditions indicate continued compliance with the steam generator component structural integrity requirements.
5.S.2 Thermal.Ihdraulic and U-Bend Vibration Evaluations he thermal-hydraulic and U-bend vibration evaluations discussed in Reference 2 were reviewed for applicability to the operating conditions for a steam generator tube plugging level of 30%. The 30% SGTP operating conditions show that a minimum steam pressure of 760 psia and the NSSS rated power of 2660 MWt will be maintained. Under these restrictions, the evaluations of Reference 2 (thermal-hydraulic and U-bend Vibration) continue to apply at the level of 30% tube plugging. SM.3 Tube Cru(h Evaluation (Flow Area Reduction Under Cembined LOCA + SSE Loads) For the combined LOCA + SSE (Safe Shutdown Eanhquake) condition, the potential exists for yielding of the tube support plates, followed by permanent deformation of tubes and subsequent loss of flow area through the steam generator tubes. Typically this reduction in flow area has been conservatively assumed to be 5% of the total flow area. An analysis has been performed for the Beaver Valley Unit I steam generators considering the effects of the combined LOCA + SSE loading WPFI802D 1D051793 5.]6
l f condition to assess the actual reduction in flow area. De analysis indicates that the total flow area .I reduction for the limiting line break, for the Beaver Valley Unit I steam generators is 0.23 W. This
- result has been factored into the LOCA analysis results. Please refer to Section 6.6, "LBLOCA j Results," for further discussion of the use of the actual value for flow area reduction.
*. l 5.9 AUXILIARY EQUIPMENT l l
he design of the NSSS auxiliary equipment utilizes as design inputs the power capability parameters for RCS normal operating temperatures (i.e., RCS hot leg temperature and cold leg temperature) and l the NSSS design transients. As discussed in Section 2.0 and 4.0, the RCS normal operating parameters and design transients do not change as a result of the increase in steam generator tube , plugging level to either the 30% or the revised TDF limit. Derefore, the design basis analysis and evaluation of the BVPS Unit 1 NSSS auxiliary equipment remain applicabl: for tube plugging levels ! i up to either the 30% or the revised TDF limit. 1 I
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t 6.0 SAFETY ANAI,VSIS AND EVAI,UATION . t E
6.1 BACKGROUND
He effect on the NSSS of an increased steam generator tube plugging level will primarily be a decrease in the primary to secondary side heat transfer area, resulting in reduced secondary side steam j pressure and temperature. Also, there will be a slight increase in the pressure drop through the steam l generator tubes and a small decrease in the RCS active volume. f Beaver Valley Power Station Unit 1 is currently licensed to operate at a Nuclear Steam Supply System [ (NSSS) power level of 2660 MWt, with a Reactor Coolant System (RCS) Bermal Design Flow (TDF) l of 88.500 gpm per loop. De original and revised NSSS design parameters for Beaver Valley Unit I are shown in Table 2.1-1. , i Previously, Duquesne Light Company performed a Steam Generator Analysis Program for Unit I to ' f f - justify operation with as many as 20% of the tubes plugged in any steam generator, not to exceed the plugging level that would reduce the RCS flow rate below the TDF limit of 88.500 gpm. De TDF limit is a conservatively low design parameter that is used in the thermai-hydraulic design of the RCS l and a variety of system, component, and safety analyses. De TDF limit is selected to be f conservatively low with respect to the actual RCS flow rate, so that flow margin is available over that i assumed in the design and analysis of the NSSS and NSSS components. Once operational data is obtained and the amount of flow margin is known, this margin can be subsequently used to accommodate reductions in RCS flow due to modifications such as steam generator tube plugging. Recently, Duquesne Light Company initiated a program to decrease the TDF for Unit I to 87,200 gpm per loop. The evaluations performed to justify a reduced thermal design flow of 87,200 gpm per loop , are documented in References 3 and 4. 1
. i ne purpose of this safety evaluation is to provide the necessary licensing support for Beaver Valley - Unit I foi operation with an increased steam generator tube plugging (SGTP) level of 30%. Actual plugging levels will be limited to a level at which the TDF of 261,600 gpm will be maintained with a .
maximum flow asymmetry of 5%, up to a maximum of 30% tube plugging. 1 WPF 1802D:lDA15179'l 6-1
i t 6.2 I.ICENSING APPROACli AND SCOPE l l i
- Re increase in the steam generator tube plugging levels at Unit I represents a change to the plant as j described in the UFSAR. Title 10 of the Code of Federal Regulations, Section 50.59 (10 CFR 50.59) .- allows the holder of a ticense authorizing operation of a nuclear power facility to make this type of change without prior Nuclear Regulatory Commission (NRC) approval. Prior NRC approval is not required to implement the change if the proposed change does not involve an tmreviewed safety question or if it does not result in a change to the plant technical specifications incorporated in the !
i license,11 is, however, the obligation of the licensee to maintain a record of the change or [ modification to the facility, to the extent that such a change impacts the FSAR. Additionally, ! 10CFR50.59 further stipulates that these records shall include a written safety evaluation which l provides the basis for the determination that the change does not involve an unreviewed safety , question. It is the purpose of this section to support the requirement for a written safety evaluation. He safety evaluation provided in this section has been prepared pursuant to the requirements of l t
. 10CFR50.59. He scope is limited to an evaluation for operation of Beaver Valley Unit I at increased steam generator tube plugging up to 30% maximum tube plugging level or up to the plugging level corresponding to the RCS Thermal Design Flow Limit, whichever is lower, centering on any effects ;
said operation may have on existing plant equipment or any unreviewed safety questions that may be i identified. De criteria stated in Chapter 14 of the Unit 1 UFSAR were used in establishing the continued applicability of the licensing basis safety analysis by demonstrating that the conclusions for - I the safety analyses remain valid. q r 6.3 SYSTEM AND COMPONENTS EVALUATIONS q The following sections present the system and components evaluation for the steam generator tube , plugaing/ thermal design flow limit program for Beaver Valley Unit 1. This program includes the evaluation of increasing the tube plugging to 30% for any steam generator or to the level i
. corresponding to the Hermal Design Flow limit. These evaluations indicate that the revised ;
i parameters associated with the increased tube plugging do not affect the NSSS design transients. : I wifiko2D:tnes1793 6-2 i
'4 i
c 1 i 6.3.1 NSSS Fluid Systems The RCS and certain NSSS auxiliary systems provided by Westinghouse were evaluated to verify their
- 'l continued adequacy for plant operation with increased steam generator tube plugging. Since it has l i
o been determined that the 30% SGTT) parameters cause no changes to the NSSS or Auxiliary j Equipment Design Transients, thae is no effect on the fluid systems designs; the RCS, CVCS, RHRS, f and SIS are acceptable ior 100% power operation. 6.3.2 Reactor Pressure Vessel System ( De reactor pressure vessel system consists of the reactor vessel, the reactor upper and lower internals '[ assemblics and the reactor core. Since these components are interdependent from a thermal-hydraulic l and structural viewpoint, they were evaluated as a system. Although the reactor pressure vessel system is sensitive to variations in the reactor coolant system flowrate, it is not directly impacted as a , 1 result of steam generator tube plugging. Herefore, the conclusions presented in Reference I assessing ;
. the impact of the reduced Thermal Design Flow on the reactor pressure vessel system remain valid for 30% SGTP.
In summary, new Dows and pressure drops were calculated for the various flow paths within the reactor pressure vessel system. De results showed that the changes in pressure drops associated with , . the new operating conditions are evenly distributed throughout the reactor internals, and that the total pressure drop acmss the internals would decrease by an insignificant amount. Since the internals flow l and pressure drop changes are not changed significantly by the new operating conditions, detailed - calculations of the effect on core bypass flow, hydraulic lift forces, flow induced vibration and Rod Control Cluster Assembly (RCCA) rod drop times were not performed. Additionally, the temperature .I rise across the reactor vessel is bounded by the original structural analyses of the Beaver Valley Unit I ! internals.
.. De evaluation of the reactor pressure vessel system indicated that there would be no advery impact l on the performance of the system by the proposed reduction in thermal design flow.
j a b i wPrisozo:inestm 6-3 l y - m . -- --
6.3.3 Reactor Coolant Pump and RCP Motor
- De current design transients remain bounding, therefore only the effects of the changes to the Power Capability Parameters were evaluated. A review of the Power Capability Parameters shows that the changes which would affect the RCP are very small. The reactor coolant temperature change is small, and there is no change in pressure. Compliance with the design criteria is not affected.
De RCP motor evaluation shows that operation with the revised loads, caused by the revised Power Capability Parameters, will not exceed NEMA temperature rise limits. Also, the rotor winding temperature rises, during worst case starting scenarios with the revised loads, have been evaluated (Reference 6). De calculated rotor winding temperature rises (based on a conservative all heat stored analysis) exceed the design allowances for bars and for rings. The consequence of exceeding the design allowances for rotor winding temperature rise is an accelerated rate of mechanical aging I ('atigue). The accelerated aging could potentially result in a rotor winding failure, meaning a crack developing in the resistance ring, which could eventually cause a failure of the motor to start. There is. I
. no impact on the safety-related function of the motor, if the motors are operated as is, the following precautions should be taken:
- The starting duty should be changed to allow no consecutive starts. A cooling period of 20 minutes running or 45 minutes standing idle should be required before any start attempt.
- The rotor should be inspected visually for cracks in the resistance rings and brare joints at a minimum interval of every five years. If cracks were found, an immediate repair is recommended.
Please refer to Section 5.3.2 for additional discussion of the RCP motor evaluation. 6.3.4 Main I,oop Stop Valves The design parameters for the 30% SGTP were reviewed to determine the impact on the loop stop l l . valves. De operating temperature for the valves was reduced below the operating range specified in the design specification. However, the reduction has no significant impact on valve pressure boundary integrity or on the valves
- operation.
wPriso:.D:stwsim 6-4
E 6.3.5 Control Rod Drive Mechanism and Capped I.atch flousinn - A review of the Power Capability Parameters shows that the changes which would affect the CRDM and CL11 are very small. De small temperature change would have a negligible effect on the analysis o of the pressure boundary components, and there is no change in pressure. Derefore, it is concluded , that compliance with the design criteria is not affected. i 6.3.6 PresSuriier ne increased tube plugging limit has been evaluated for the impact on the pressuriier. De evaluadon indicated that the pressuriter components continue to meet the ASME Code, Section III stress analysis and fatigue analysis requirements. 6.3.7 Steam Generators . The steam generator evaluation for the revised conditions was d.vided into three areas - a thermal-hydraulic evaluation, a U-bend vibration evaluation, and a structural evaluation. - l Thermal.llydraulic Evaluation 1 The results of a thermal / hydraulic evaluation 'idicated that operation with 30% SGTP was acceptable I with the current hardware. Previous analyses were based on a power level of 887 MWt per steam generator and a steam pressure of 760 psia. Rese principal parameters, that is the power level and the secondary side steam pressure, are unchanged from previous analyses performed. Rus, the acceptability.of the thermal / hydraulic operating characteristics continues to be applicable for 30% j SGTP. U.itend Vibration De primary parameters affecting U-bend vibration of the tubes are the power level and the steam - i l pressure. Since the power level and the steam pressure do not change from previous analyses performed, the stability ratio of the U-bends and thus the fatigue usage are not affected. No remedial l
~l action is needed with regard to U-bend fatigue. ,
l l l
' WPFlkO2D:lDK61793 6.$ l
i 7 Structural Analvsis I
- The original structural analyses were based on a steam pressure of 790 psia. Reference 2 provided r evaluations which were based on a steam pressure of 760 psia. For the present study, the steam j pressure used was 760 psia. He structural an: tyser focused on the effects of reduced steam pressure resulting in an increased primary to secondary side pressure differential. The results indicated that, ,
although the 30% SGTP condition produces slightly increased stresses in certain component areas, the ; increases are not significant. De stress predictions are conservative due to the conservatism in the l assumed pressure differential. Fatigue analyses performed show that acceptable fatigue usage factors can be demonstrated for the 30% SGTP conditions. l t i 6.3.8 NSSS Auxiliary Equipment l r
?
The NSSS auxiliary equipment evaluation for the revised conditions was divided into three areas - -l auxiliary heat exchanger / tanks, auxiliary valves, and auxiliary pumps. f Auxiliarv fleat Fxchancer/I'anks j i
~
The regenerative heat exchanger, residual heat exchanger, seal water heat exchanger, excess letdown heat exchanger, and letdown heat exchanger were evaluated for 30% SGTP. In addition to the i i auxiliary heat exchangers, the only tanks that have transients identified are the boron injection tank l (BIT) and the safety injection accumulators. As a result of the BIT boron concentration reduction l l I program at Beaver Valley Unit 1, the original design transients of the BIT are no longer applicable. Therefore, the BIT is not impacted by the reduction in thermal design flow or increased steam generator tube plugging levels. Also, since the safety injection accumulator vessels do not have significant design transients requiring a fatigue analysis, they also are not impacted by the steam l generator tube plugging or th5 TDF reduction.
.. A review of the original design and qualification requirements for the Beaver Valley Unit I heat exchangers shower that the rcrating parameters for the regenerative heat exchangers, the letdown heat exchangers, excess setdown heat exchangers, and residual heat exchangers are bounded by the original design parameters. The seal water heat exchangers were not required to be qualified for pressure or temperature transients. De transients were not included in the design, as they were not expected to have an effect on these components. Therefore, the equipment is designed for only maximum steady WPf!802D.lDKr51793 66
i I 1 state pressures and temperatures, and the 30% SGTP will not impose any new limitations on the seal ! water heat exchangers. ; Auxiliarv Valves Evaluations of the original design and qualification requirements of the auxiliary valves at Beaver i i Valley Unit 1 indicated that the parameters associated with 30% SGTP are bounded by the original ! l design parameters. f Auxiliary Pumps j
'i The charging / safety injection pumps, residual heat removal pumps, low pressure safety injection ' l pumps, boron injecdon recirculation pumps, and boric acid transfer pumps were evaluated for 30%
SGTP. De specifications require the pumps to be qualified for pressure and temperature transients, or, if the equipment was not expected to be significantly affected by the transients, it was designed for ;
. maximum steady state pressures and temperatures only. The evaluadon indicated that the design !
qualification for the charging / safety injection pumps, residual heat removal pumps, low pressure safety : injection pumps, boron injection recirculation pumps, and boric acid transfer pumps remains bounding for the conditions of 30% SGTP. 6.4 NON 1,0CA EVAI,UATION i The current non-LOCA analysis bases for Beaver Valley Unit I assume a total RCS thermal design i flowrate (TDF) of 265,500 gpm (88,500 per loop). Beaver Valley Unit I is currently analyzed to l l allow up to 20% steam generator tube plugging. Tube plugging increases the system hydraulic l resistance of the RCS and may result in lower loop flowrates and greater reactor vessel temperature differentials (not -Tcold). ! 1 i
. This evaluation supports an increase in the steam ;;enerator tube plugging level to an average plugging level of 30%. This evaluation also considers a TDF of 261,600 gpm (87,200 gpm/ loop) and up to 5%
i h>op flow asymmetries. . A reduction in Thermal Design Flow has an adverse effect on the core thermal limits (consequently the Overtemperature and Overpower AT analysis setpoint equations) and the initial conditions assumed WPT 1tt02D.IDMi793 67 {
1 i for all of the non-LOCA transients. The core thermal limits were reviewed and revised to account for , l the reduced flow. Tne current OTAT and OPAT analysis setpoint equations were reviewed and _i
. confirmed to provide protection for the revised core limits. l t
All non-LOCA transients were examined to determine the impact of the reduced TDF. 'The current ! DNB limits were examined and determined to remain valid for the TDF reduction and no generic ; margin is needed to offset the effect of the flow reduction on DNBR. Because the TDF reduction is limited to approximately 1.5%. existing flow sensitivities were used to demonstrate that non-DNB safety criteria (e.g. peak chid temperature, RCS pressure) will also continue to be met. l Although the allocation of generic DNB margin is not required for the TDF reduction, some of the non-LOCA events required the allocation of DNB margin to offset the effects of the increased steam. j generator tube plugging or the flow asymmetry. The events which require the allocation of generic [ DNB margin are identified in the discussion of each non-LOCA event below. . I t
. Steam generator tube plugging asymmetries lead to flow asymmetries among the reactor coolant loops.
, The loop N the largest amount of tube plugging will have the lowest reactor coolant flow. Because i i of the mixing in the reactor vessel lower plenum, temperature asymmetries resulting from the flow asymmetries are minimized. The greatest concern which must be addressed is the effect on transients -l i which are sensitive to flow asymmetries. The following events were analyzed to account for up to a, j 5% kiop flow difference and a reduced TDF: , a -
. Partial Loss of Forced Reactor Coolant Flow ;
Single Reactor Coolant Pump Locked Rotor i i The following discussion addresses the impact on the non-LOCA s'icty analyses for 1) an approximate 1.5% reduction in thermal design flow 2) an average tube plugging level of 30% and 3) up to 5% ; loop flow asymmetry. Each non LOCA licensing basis event is discussed in the order in which a i
. description of it appears in the UFSAR. ;
.I i
Please note that non-LOCA evaluation references for this section are listed in Section 6.4.24. WPFIS02D:1DM51793 6-8 l
i i l 6.4.1 Uncontrolled Rod Cluster Control Assembly (RCCA) llank Withdrnwn! from a Suberitical Condition (UFSAR I4.1.1) For this Condition 11 event, rod withdrawal results in a rapid reactivity insertion anu merease in core power potentially leading to high local fuci temperatures and heat fluxes and a reduction in the !
~
minimum DN13R. The power excursion is terminated by Doppler feedback and then the transient is promptly terminated by a reactor inj on the Power Range liigh Neutron Flux - low setpoint. Due to the inherent thermal lag in the fuel pellet, heat transfer to the RCS is relatively slow. The reduction in TDF would result in a reduction in fuel-to-coolant heat transfer and the associated heat flux reduction would be a DNB benefit. The impact of steam generator tube plugging on the udnimum DNBR is negligible since transients initiated by primary side faults are insensitive to small ! changes in the secondary side operating parameters. The combined effect of the reduced TDF and increased SGTP ou DNB would be insignificant The reduction in fuel-to-coolant heat trarufer due to { the TDF rer 9a would also tend to increase peak clad temperatures (PCTs). Sensitivities show this l increase to be on the order of 1(XI*F. The Beaver Valley Unit I licensing basis analysis can accommodate this PCT penalty.
, Vic asymmetric steam generator tube plugging levels discussed in the introduction will result in flow and inlet temperature asymmetries between the RCS k> ops. Ilowever, these asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum resulting in a negligible effect on the transient. Thus, the conclusions for this event as presenteri in the UFSAR (Reference 6.4.1) remain valid.
6.4.2 Uncontrotkd Rod Cluster Control Assembly (RCCA) Ilank Withdrnwn! at Power , (UFSAR Section 14.1.2) For this Condition 11 event, tlirce initial power levels (i.e.,100%,60%, and 10%) and a range of reactivity insertion rates assuming two (ndnimum and maximum) reactivity feedback conditions are analyzed. The resulting power excursion produced by a RCCA withdrawal at power results in high l local fuel temperatures and an increase in the corc heat flux. Since the heat extraction capability of the s, cam generator lags behind the core power generation, a net increase in the moderator temperature occurs. The resulting power mismatch and increase in reactor coolant temperature can result in DNB unless the transient is terminated by either manual or automatic means. Automatic reactor protection is provided via the liigh Neutron Flux and the Overtemperature AT trip functions. The licensing-basis a wri: Iso 2p.t n/osl793 6-9 f
i analysis presented in the UFSAR ensures that fuel damage will not occur by demonstrating that the minimum DNBR remains above the safety analysis limit value. ! f'
' he reduction in RCS flow would impact the minimum DNBR. Ilowever, it has been determined that the core limits used in the calculation of the minimum DNBR are more limiting than the core limits based on the current design basis. And, since the transient conditions are not altered by the TDF reduction, the conclusion that the DNB design basis is met remains valid and no allocation of generic DNB margin is required. De impact of steam generator tube plugging on the minimum DNBR is negligible since t.ansients initiated by primary side faults are insensitive to small changes in secondary >
side operating parameters. Bus, the combined affect of the reduced TDF and increased SGTP on I DNB would be insignificant. The asymmetric steam generator tube plugging levels discussed in the introduction will result in flow -! and inlet temperature asymmetries between the RCS loops.110 waver, these asymmetries will be ! attenuated by coolant mixing in the reactor vessel lower plenum re.sulting in a negligible effect on the ( transient. Rus the conclusions for this event as presented in the UFSAR remain valid. l h 6.43 Rod Cluster Control Assemhiv Misalignment (UFSAR Section 14.13) lt
?
i This Condition 11 event is analyzed to demonstrate that following various RCCA misoperation events, j l such as dropped rod (s)/ bank or statically misaligned rods, the minimum DNBR remains above the l safety analysis limit value. De reduction in RCS flow potentially impacts the RCCA misoperation ; events by changing the initial condition assumptions used in this analysis. Reductions in the RCS ; flow and/or increases in the RCS temperature serve to reduce the margin to the DNB licensing-basis j limit. f De asymmetric steam generator tube plugging levels discussed in the introduction will result in flow and inlet temperature asymmetries between the RCS loops. Attenuauon of these asymmetries will j occur, however, as a result of coolant mixing in the reactor vessel lower plenum. Furthermore, the i no. inal RCS average temperature will continue to be controlled to the same value which was 4 prev,ously assumed in the UFSAR analysis. l I
. Allocation of the generic DNB margin ensures that the DNB licensing-basis criteria will continue to be ;
met and the conclusions in the UFSAR for this event remain valid. 1 wrriso2n;1Dc5 m 6-10 1 ^
- g. -
I I t 6.4.4 Uncontrolled Baron Dilution (UFSAR Section 14.1.4) [ i His Condition 11 event is analyzed for three modes of plant operation. De analysis indicates that sufficient shutdown margin exists, such that should a dilution event occur, there is sufficient time following the start of dilution to allow operator detection and termination of the event prior to a ! complete loss of shutdown margin. This event is analyzed for operating modes 1,2, and 6. [ t An input to the boron dilution analysis for Modes I and 2 is the RCS active volume,i.e., the total l RCS volume minus the volumes of the pressurizer, die pressurizer surge line, the dead volume of the f reactor vessel head, and plugged steam generator tubes. Reduction of the RCS active volume is , directly proportional to the reduction in operator response time for the boron dilution event described f T in the Beaver Valley UFSAR. De boron dilution event for Beaver Valley was ceanalyzed to indicate that sufficient shutdown margin f exists, such that, should a dilution event occur, there is sufficient time to allow operator action to f terminate the event prior to a complete loss of shutdown margin. Although the reduced thermal design
. flow does not adversely affect the calculations, the increased SGTP reduces the RCS volume assumed ;
in the analyses. Results of the reanalysis for the Modes 1 and 2 Boron Dilution show that the operator j has more than 15 minutes between initiation of event and loss of shutdown margin. Mode 6 operation also is unaffected since the active volume assumed comes from the reactor vessel and one RiiR loop ! (i.e., the active volume of the steam generator is not assumed in Mode 6). Therefore, the increased f tube plugging levels (and subsequent reduction in thermal design flow) will not impact the conclusions ; of the UFSAR. f i r As a result, the conclusions presented in the UFSAR for the boron dilution event in Modes 1,2, and 6 [ remain unaltered with the reduced RCS flow and the asymmetric tube plugging levels. . I 6.4.5 Partial I,oss of Forced Reactor Coolant Flow (UFSAR Section 14.1.5) l I ne Partial Loss Of forced reactor coolant Flow (PLOF) transient is a Condition 11 event which is .i analyzed under full power conditions assuming that 1 of 3 operating reactor coolant pumps coasts l
^ '
down. The reactor is promptly tripped on low reactor coolant loop flow. De current UFSAR analysis demonstrates that the minimum DNBR remains above the safety analysis limit value. The case with ; { I WPFis02D.1D/051793 6-11 . t i e
I i l l I J 3 loops operating prior to 1 loop coastdown was analyzed to incorporate the RCS flow reduction as , well as the asymmetric tube plugging scenario. I he reduced RCS thermal design How will have a negative impact on the calculated DNBR because it ; is a critical parameter in DNBR determination. Asymmetric steam generator tube plugging may { adversely impact the PLOF results from UFSAR since the reactor coolant pump could be lost in the loop with the lowest level of tube plugging (and thus highest flow). In this case, an additional reduction in RCS flow occurs since forced flow is maintained through the two loops having higher steam generator tube plugging levels (and thus lowest flow). De reduction in RCS thermal desiga ; flow may also produce an increase in the RCS moderator temperature, both of which tend to reduce the margin to the licensing-basis DNB limit for this event. The event was analyzed to determine the effects of the reduced TDF and incorporate the asynunetric scenario using the LOFTRAN (Reference 6.4.2) and FACTRAN (Reference 6.4.3) computer codes. De THINC code was used to calculate the minimum DNBR during the transient. Some of the key , r assumptions used in this analysis include: i
- a. Three reactor coolant k> ops were operating prior to the loss of flow in one loop. .
t
- b. To account for asymmetric tube plugging effects, the loss of flow occurred in the loop having the lowest level of tube plugging (i.e., the highest loop flowrate)
- c. De initial power level was 102% of the rated thermal power. ,
- d. De system pressure was assurned to be 42 psi lower than the no.ninal RCS pressure. ,
- c. The low flow reactor trip setpoint was 870k of nominal. [
he results of this analysis show that the minimum DNBR is still bounded by the complete loss of forced reactor coolant flow analysis. As a result, the increased tube plugging with reduced thermal design flow as well as the asymmetrical steam generator tube plugging levels do not alter the conclusions presented in the UFSAR for the PLOF event. Figures 6.4-1 through 6.4 3 depict the transient results for the partial loss of flow event. wPFt802D.1DAr51793 6-12
'i 5
6.4.6 Startup of an Inacthe Reactor Coolant I.oop (UFSAR Section 14.1.6) ; This Candition 11 event is analyzed assuming a maximum initial power level consistent with 2 loop operation. The startup of an inactive loop results in a reactivity insenion since the inactive loop fluid ! being injected into the core is at a lower temperature relative to the remaining fluid in the core. The analysis demonstrates that the minimum DNBR remains above the safety analysis limit value. The automatic reactor protection system terminates this transient on power range High Neutron Flux. The t licensing-basis analysis presented in the UFSAR demonstrated that the minimum DNBR was greater than the safety analysis limit value. The reduction in RCS flow would impact the minimum DNBR. Ilowever, it has been determined that ; the core limits used in the calculation of the m:nimum DNBR are more limiting than the core limits l based on the current design basis. And. since the transient conditions are not altered by the TDF reduction. the conclusion that the DNB design basis is met remains valid and no allocation of generic DNB margin is required. i liowever, asymmetric steam generator tube plugging potentially impacts these results since the mass of - cold water from the inactive loop can hypothetically come from the loop with the lowest steam i generator tube plugging level. In this situation, a larger mass of cold fluid at a higher rate enters the l core and produces a larger reactivity insertion (due to the negative moderator temperature coefficient) l i relative to the case where symmetric steam generator tube plugging was assumed. l i For conservatism, this evaluation assumes that the inactive k>op undergoing startup ccmtains the lowest percentage of tube plugging. This will produce the largest insurge of cold water into the core and ; hence the largest positive reactivity excursion. Initial RCS thermal design flow from the twc. active loops would be reduced somewhat, in comparison to the UFSAR analysis, since this flow would be based on the two loops containing the highest levels of steam generator tube plugging. During the transient, the reactor coolant flow increases due to the startup of the inactive pump, thereby l 1 producing a DNB benefit. However, a DNB penalty may be introduced as a result of a larger positive reactivity insenion in the core produced by this increased flow and redu-ed temperature. Also note > that since the reactivity feedback will be larger. the power will also rr ch the high neutron flux trip setpoint more quickly relative to the analysis documented in the UFMR. Since the increases in thermal design flow and power are competing DNB effects, the changes in minimum DNBR due to WPFlkO D.lDo$1793 6-13
t a the asymmetric tube plugging scenario are estimated to be small. The results from the analysis [ documented in the UFSAR indicate that sufficient margin exists to the safety analysis DNB limit (i.e., ! ,, to accommodate rrdnor perturbations in the minimum DNBR). I Therefore, since the asymmetric tube plugging will have an insignificant impact on the transient and the RCS thermal design flow decrease is accommodated by the allocation of generic DNB margin, the j conclusions in the UFSAR will remain valid for this event. 7 I 6.4.7 I,oss of External Electrical I,oad and/or Turbine Trin (idSAR Section 14.1.7) , i The analysis presented in the UFSAR represents a complete loss of steam load from full power without a direct reactor trip. Four cases are analyzed which are based on two different primary side pressure control strategies (automatic versus none) and two sets of core physics characteristics ; (maximum versus minimum reactivity feedback). He analysis indicates that, with the power j mismatch between the core and turbine, the primary and secondary system pressures remain below 110% of the design values and the minimum DNBR remains above the safety analysis limit value. Automatic reactor trip signals which may be generated during this event include high pressurizer : pressure and ovenemperature AT. We reduction in RCS flow can potentially impact the results of
~
this analysis with respect to the minimum 4BR calculated. The reduction in RCS flow would impact the minimum DNBR. However, the core . mats used in the calculation of the minimum DNBR for . I TDF reduction are more limiting than the core limits based on the current design basis. And, since the transient conditions are not altered by the TDF reduction, the conclusion that the DNB design basis is met remains valid and no allocation of generic DNB margin is required. In all four cases, there is substantial margin to the primary / secondary side pressure limits. Furthermore, this transient is insensitive with respect to the pressure limits, to a change in reactor l coolant flow of the magnitude specified in this evaluation. Rus, an approximate 1.5% reduction in RCS flow would not result in the violation of the licensing basis criteria following a loss ofload i event. An additional concern during the loss of load event is ensuring that the pressurizer will not fill as pressurization of the RCS results in an insurge into the pressurizer. Sensitivity analyses show that for i a change in RCS flow of 6%, the peak pressurizer volume only increases by 2 ft'. He Beaver Valley l licensing basis analysis can accommodate this small penalty in pressurizer volume. { wmsc2D:lD/051793 6-14 l l
De asymmetric steam generator tube plugging levels discussed in the introduction will result in flow and inlet tempe"ature asymmetries between the RCS loops. These asymmetries will be attenuated. however, by coolant mixing in the reactor vessel lower plenum. Also note that operability of the Overtemperature AT reactor trip function in this event is still assured under asymmetric flow and temperature conditions since each channel (1 channel / loop) is calibrated for the specific loop inlet and outlet conditions. 6.4.8 Loss of Normal Feedwater (UFSAR Section 14.1.N) The loss of normal feedwater analysis in the UFSAR presents the consequences.of a complete loss of i normal feedwater flow simultaneous to all three steam generators. He loss of AC power event (UFSAR Section 14.1.11) is similar except that the loss of offsite power also results in all three reactor , l coo!Lnt pumps (RCPs) coasting down. Rese transients are analyzed to indicate that neither the primary nor secondary sides are overpressurized, that the core is not adversely affected, and the pressurizer does not fill. i ne Loss of Normal Feedwater event is sensitive to initial steam generator mass as well as the mass in the steam generators at the time of reactor trip. An increased tube plugging level causes a slight ^ decrease (< 0.5%) in the initial steam generator mass used in the Loss of Normal Feedwater event. , Following the loss of normal feedwater, the reactor continues to operate until, due to the rapid loss of ; steam generator inventory and the continued heat transfer to the secondary side, it is tripped on a low-low steam generator level signal. The increased tube plugging level has less than a 0.5% decrease in the initial mass and the mass at time of reactor trip. De effect of reducing the RCS flow would be an increase in the heatup of the RCS during the initial phase of the transient. De increased heatup results in a decrease in the coolant density which in turn would increase the pressurizer insurge during this heatup. A sensitivity modeling reduced TDF and 30% SGTP results in an increase in pressurizer volume of ~23 ft'. De Beaver Valley licensing basis analysis contains enough margin to accommodate this increase in pressurizer volume.
~
During the long-term portion of the transient, the peak RCS temperature (and resultant peak pressurizer water volume) is reached when the heat removal capability of the auxiliarv feedwater system matches the core decay heat generation. If the assumed RCS flow reduction is due to higher . WPFl802D.lDo51m 6-15
I than anticipated loop flow resistances, the natural circulation How will be reduced by an amount proportional to the approximate 1.5'7c thermal design flow reduction. This slight reduction in natural i circulation flow at the peak RCS temperature condition would not significantly impair the heat transfer . l across the steam generator tubes, and would not result in a significant change in hot leg temperature or peak pressuriier water volume. De UFSAR Loss of Normal Feedwater analysis has enough margin f a to accommodate the reduced RCS flow and reduction in steam generator tube heat transfer area due to tube plugging. Derefore, the approximate 1.5% reduction in RCS 110w and an average steam generator tube plugging level of 30% does not cause the safety analysis acceptance criteria to be exceeded. The asymmetric steam generator tube plugging scenario will result in flow and inlet temperature asymmetries between the RCS loops.110 wever, these asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum. Rus, the primary and secondary side peak pressure licensing-basis design criteria will continue to be met, the pressurizer will not go solid and the conclusions made in the UFSAR for the loss of normal feedwater event remain valid. 6.4.9 Excessive Ileat Removal Due to Feedwater System Malfunctions (UFSAR 14.1.9) , Two cases are analyzed and described for this ANS Condition 11 event in the UFSAR. A full power case is used to .letermine the plant response to a large step increase in the feedwater flow to one steam generator; a zero power case examines a step increase in feedwater flow from zero to nominal full-load flow in one steam generator. For the full power case, the minimum DNBR is shown to remain above the safety analysis limit value. De zero power case indicates that the reactivity transient, and hence the minimum DNBR, is bounded by the rod withdrawal from subcritical event. %e reduction in RCS flow would impact the minimum DNBR.110 wever, the core limits used in the calculation of the minimum DNBR for TDF reduction are more limiting than the core limits based on the current design basis. And, since the transient conditions are not altered by the TDF reduction, the DNB design basis remains valid and no allocation of generic DNB margin is required. . t The reactivity insertion of the zero power feedwater malfunction event would decrease slightly as the ; TDF decreases and the SGTP increases because the primary to secondary heat transfer capability will
. t be slightly decreased. This results in less of a primary cooldown t.nd less of a reactivity transient. -
The maximum reactivity insertion will remain bounded by the rod withdrawal from suberitical analysis. Also, the feedwater temperature reduction transient will continue to be bounded by the wrF180 D.1D/051793 6-16 !
excessive load increase event since the steam pressure and temperature will be maintained at essentially the same values. Asymmetric steam generator tube plugging levels will result in flow and inlet temperature asymmetries between the RCS loops. However, these asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum. The reactivity insenion rate of the feedwater malfunction event, driven by the RCS cooldown, will decrease if the thermal design flow decreases and the steam generator tube plugging increases because the primary to secondary heat transfer capacity will decrease. However, as previously discussed, the change in heat transfer capability is small and these effects are minimal. Therefore, the current calculated reactivity insertion rate remains valid. The reactivity insenion rate calculated for the zero power case is bounded by the RCCA withdrawal from suberitical analysis discussed previously. Therefore, the conclusions in the UFSAR, (Reference 6.4.1) penaining to the excessive feedwater flow event, remain valid. 6.4.10 Excessive 1.oad increase Incident (UFSAR Section 14.1.10) ~ The analysis presented in the UFSAR describes the plant response to a 10% step increase in load from nominal full power conditions. Four cases are analyzed for this ANS Condition II event based on automatic versus manual rod control and minimum versus maximum reactivity feedback parameters. Reactor protection against an excessive load increase transient is provided by the power range high neutron flux, overpower AT, and otenemperature AT reactor protection system signals. Each case showed that the minimum DNBR remained above the safety analysis limit value. The reduction in primary to secondary heat transfer capability due to the increased SGTP would be a slight DNB benefit, but the reduction in RCS flow would adversely impact the minimum DNBR. However, the core limits used in the calculation of the minimum DNBR are more limiting than the core limits based on the current design basis. ' And, since the transient conditions are not altered by the TDF reduction, the DNB design basis remains valid and no allocation of generic DNB margin is required. Asymmetric steam generator tube plugging levels cause flow and inlet temperature asymmetries between the RCS Imps. Coolant mixing in the reactor vessel lower plenum, however, will act to reduce these affects. "Ihus, the DNBR will remain above the safety analysis limit value and the conclusions in the UFSAR pertaining to the excessive load increase event, remain valid. wtrito:n;inest?93 6-17
i 6.4.11 1.oss of OfTsite Power to the Station Auxiliaries (Station illackout) (UFSAR Section 14.1.11) ! The analysis presented in the UFSAR represents a complete loss of power to the plant auxiliaries (i.e., the reactor coolant pumps, condensate pumps, etc.) from full power. The loss of power results in a ; heatup and pressurization of the primary and secondary systems. The analysis indicates that adequate ; auxiliary feedwater flow is delivered to the steam generators to remove decay heat such that DNB will not occur, overpressurization of the primary and secondary systerns will not occur, and the pressurizer will not become water solid. Steam generator tube plugging potentially impacts this ANS Condition 11 event by reducing the RCS thermal design flow. Reductions in the RCS thermal design flow decrease the minimum DNBR, increase the peak RCS pressure and potentially lead to increased coolant expansion and a reduction in the margin to pressurizer filling. j , When offsite power is lost, the reactor coolant pumps coastdown and the RCS thermal design flow will eventually reduce to natural circulation flow. With asymmetric steam generator tube plugging levels, natural circulation flow rates will be slightly different between the loops; however, the change in flow resistance is expected to be proportional under full and natural circulation flow. Inlet temperature asymmetries may also exist between the RCS loops; however, coolant mixing in the
~
reactor vessel lower plenum will minimize Ods effect. The dominant driving force for natural { circulation is the density difference between the fluid in the reactor vessel downcomer and the fluid within the core barrel (in the core and upper core plenum). This driving force will act to force flow through all of the teactor coolant loops. The reduction in TDF has an insignificant impact on this interaction. i
'lhe effect of reducing the RCS flow would be an increase in the heatup of the RCS during the initial ;
phase of the transient. The increased heatup results in a decrease in the coolant density which in turn would increase the pressurizer insurge during this heatup. A sensitivity modeling reduced TDF and i 1 30% SGTP results in an increase in pressurizer volume of ~23 ft'. The Beaver Valley licensing basis l l analysis contains enough margin to accommodate this increase in pressurizer volume. l 1 The results of the complete loss of forced reactor coolant flow analysis (UFSAR section 153.4) and the loss of normal feedwater analysis (UFSAR section 15.2.8) continue to show that for a loss of all non-emergency AC power, no adverse conditions occur in the reactor core. As a result, the DNBR remains above the safety analysis limit value and the primary and secondary side peak pressure l l WPF 1802D:1D051793 6 18 l
i i licensing-basis design criteria continues to be met. Pressurizer filling also will not occur; therefore, l l the conclusions for the stadon blackout event which are documented in UFSAR remain valid. ; 6.4.12 Accidental Depressurization of the Main Steam System (UFSAR 14.1.13) ; i l
-i This ANS Condition 11 event is initiated by the full opening of a single steam dump, relief, or safety j valve from zero power conditions. De analysis indicates that the minimum DNBR remains above the !
safety analysis limit value. i i Steam generator tube plugging potentially impacts the Main Steam System depressurization event by - reducing the RCS thermal design flow assumed in the analysis. Reductions in the RCS thermal design flow potentially decrease the minimum DNBR calculated during the event. However, reduced l flowrates result in less primary to secondary heat transfer and consequently, less of a power increase. He reduction in RCS flow would impact the minimum DNBR. However, the core limits used in the calculation of the minimum DNBR are more limiting than Ine core limits based on the current design basis. Herefore, since the transient conditions are not altered by the TDF reduction, the conclusion j that the DNB design basis is met remains valid and no allocation of generic DNB margin is required. l The asymmetric steam generator tube plugging levels produce flow and temperature asymmetries between the RCS loops. This event is analyzed at no-load conditions where the temperature asymmetries would be minor. He flow asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum. De DNBR remains above the safety analysis limit value and the ; conclusions presented in the UFSAR for the Main Steam Supply depressurization event, remain valid. 6.4.13 A,ccidental Depressurization of the Reactor Coolant System (UFSAR 14.1.15) f 1 b For this ANS Condition H event, the transient is initiated by the opening of a single pressunzer relief or safety valve while the reactor is at full power. Initially, the RCS pressure drops rapidly until a i reactor trip occurs on either the pressurizer low pressure or Overtemperature AT reactor protection l i signals. At this time, the pressure decrease continues, but at a much slower rate. The analysis indicates that the minimum DNBR remains above the safety analysis limit value. wrriso:n;iomst?9) 6-19
i l The reduedon in RCS flow would impact the minimum DNBR.- However, it has been determined that f i the core limits used in the calculadon of the minimum DNBR are more limiting than the core limits , based on the current design basis. Ecrefore, since the transient conditions are not altered by the TDF l reduedon. the conclusion that the DNB design basis is met remains valid and no allocation of generic l DNB margin is required. .[ Be asymmetric steam generater tube plugging levels will produce flow and inlet temperature asymmetries between the RCS loops. Rese asymmetries, however, will be minimized by coolant : mixing in the reactor vessel lower plenum. Bus, the DNBR remains above the safety analysis limit . value, and the conclusions in the UFSAR for this event remain valid. 6.4.14 Spurious Operation of Safety Iniection System at Power (UFSAR 14.1.16) A spurious Safety injection System (S!S) signal is an ANS Condition 11 event which is assumed to be initiated at full power. De injection of highly concentrated (2000 ppm) borated water into the RCS l reduces core power, temperature and pressure until the reactor trips on low pressurizer pressure. t he RCS power and temperature reductions produce a similar reduction in pressure on the secondary
^
side of the plant. he analysis indicates that the minimum DNBR remains above the safety analysis limit value. Steam generator tube plugging potentially impacts the spurious SIS actuation event by reducing the RCS thermal design flow. Small changes to steady state RCS flow and secondary side l operadng condidons (due to SGTP) would have no significant impact on the transient behavior. De .; licensing basis analysis for Beaver Valley Unit I shows that the DNBR is never less than the initial value. De reduction in the RCS thermal design flow potentially decrease the minimum DNBR l, calculated during the event. However, the core limits used in the calculation of the minimum DNBR for TDF Reduction are more limiting than the core limits based on the current design basis. And, since the transient condidons are not altered by the TDF reduction, the conclusion that the DNB design basis is met remains valid and no alk) cation of generia DNB margin is required. , l Asymmetric steam Eencrator tube plugging levels will create flow and inlet temperature asymmetries between the RCS loops. De effect of these asymmetries, however, will be minimal due to coolant ; mixing in the reactor vessel lower plenum. De UFSAR results show that the DNBR is never less l i than the initial value. Furthermore, the effects of asymmetric tube plugging will not impact the l l behavior of this event. Therefore, the conclusions presented in the UFSAR remain valid. 1 WPrim:D.lD/051793 6-20 -j
l 6.4.15 Maior Secondary Side Pipe Rupture (UFSAR 14.2.5) , For this ANS Condition IV event, the transient is assumed to be initiated by the instantaneous double- [ ended rupture of a main steam line while at hot zero power conditions. Two cases (with and without offsite power) are considered. De analysis indicates that the minimum DNBR remains above the licensing limit value in each case. Steam generator tube plugging potentially impacts the main steam line break event by reducing the , RCS thermal design flow assumed in the analysis. Reductions in the RCS thermal design flow potentially decrease the minimum DNBR calculated for the event. His DNB penalty would be - partially offset because the lower flow would lead to reduced primary to secondary heat transfer and ; subsequently less of a power increase. Furthermore, the DNB design basis has been confirmed to be j met for the increased SGTP and reduced TDF. i De reduction in RCS flow would impact the minimum DNBR. However, it has been determined diat j the core limits used in the calculation of the minimum DNBR are more limiting than die core limits
=
based on the current design basis. Therefore, since the transient conditions are not altered by the TDF reduction, the conclusion that the DNB design basis is met remains valid and no allocation of generic. DNB margin is required. t Re asymmetric steam generator tube plugging levels discussed in the introduction create flow asymmetries between the RCS loops. These flow asymmetries, however, will be attenuated as the , coolant is mixed in the reactor vessel lower plenum. Since this analysis is initiated at hot zero power, temperature asymmetries will be minimal. Thus, the calculated minimum DNBR will stay aba* the licensing limit value, and the conclusions reported in the UFSAR for this event remain valid. 6.4.16 Feedwater System Pipe Break (UFSAR Section 14.2.5.2) t For this ANS Condition IV event, the double-ended rupture of a main feedwater pipe initially results in a cooldown of the RCS due to the heat removal of the steam generator blowdown. This cooldown period is followed by a heatup as the high levels of decay heat and the lack of inventory on the secondary side result in inadequate heat transfer. De event is analyzed to show that adequate auxiliary feedwater flow exists to remove core decay heat and stored energy following a reactor trip j from full power and that the core remains in a coolable geometry, and covered with water. Steam wPFIR02D:1D/051793 6-21
I l generator tube plugging potentially impacts the main feedwater line break event by reducing the RCS thermal design flow, and therefore, the primary to secondary heat transfer capability, assumed in the analysis. However sensitivities show that the FLB event is insensitive to reductions in RCS flows of up to 5% with respect to peak hot leg temperatures. De reduction in TDF and increase in SGTP would result in a slight decrease in the initial steam generator mass. De effect of the reduced mass is a small decrease in the minimum margin to hot leg saturation. However. the current Beaver Valley licensing basis analysis contains sufficient margin to accommodate this -20 'F penalty. For case of interpreting the transient the more restrictive criterion of no bulk boiling in the primary coolant system following a Feedwater Pipe Break prior to the time that the heat removal capacity of the steam generators, being fed auxiliary feedwater, exceeds NSSS generation has been applied. His . is determined by verifying that the RCS coolant remains subcooled. The analysis is not impacted by i small changes in nominal plant operating conditions (i.e., steam generator mass, RCS flow, and steam pressure). . RCS thermal design flow and inlet temperature variations will be produced in the loops having , asymmetric steam generator tube plugging levels; however, coolant mixing in the reactor vessel lower plenum attenuates these conditions. Rus, the conclusions presented in the UFSAR remain valid. 6.4.17 Rupture of a Control Rod Drive Mechanism Housine (UFSAR Section 14.2.6) , For this Condition IV event, a rapid reactivity insertion and increase in core power leads to high kral fuel and clad temperatures and possible fuel and/or clad damage. De RCCA ejection analysis is { analyzed at four conditions: beginning and end-of-life core physics characteristics, at hot zero power f and full power. The analysis indicates that gross fuel damage will not occur, that the core will remain in a coolable geometry, and that the RCS will remain intact. To demonstrate that these criteria are
- met, the following more restrictive, criteria are applied: ;
- 1) De average fuel pellet enthalpy at the hot spot is less than 200 cal /gm (360 Bru/lbm).
- 2) Fuel melt at the hot spot is limited to less than the innermost 10% of the fuel pellet.
- 3) Peak RCS pressure is less than that which would cause stresses to exceed the Faulted ,
Condition Stress Limits. , wPF1802D1DK'51793 6-22 . et
Re rod ejection event is characterized by a rapid power excursion terminated by Doppler feedback. The reactor is tripped on high neutron flux (Iow setting for the zero power cases, high setting for the full power cases). A reduction in RCS flow will result in a reduction in the fuel rod-to-coolant heat transfer. His may result in an increase in the calculated fuel and clad temperatures as well as the fuel stored energy during an RCCA ejection. As shown in the UFSAR, the full power cases result in the highest fuel pellet temperatures and are the most limiting with respect to criteria 1 and 2. Examination of these cases reveals that, due to the rapid power and fuel temperature rise coupled with the thermal lag in the fuel pellet itself, the time at which the maximum pellet enthalpy and fuel melt are calculated to occur is before any significant amount of heat has reached the coolant. A sensitivity analysis, which used methods consistent with WCAP-7588, Revision 1 (Reference 6.4.4), demonstrated that for a 2% reduction in thermal design flow, there was only a minor change to the maximum pellet enthalpy and fuel melt results for the full power rod ejection cases. Sufficient margin exists in the Beaver Valley analyses to absorb the differences in results from the sensitivity. Therefore, the approximate 1.5% reduction in thermal design How caused by the increased steam generator tube plugging would not cause the safety criteria to be violated for the full power cases. RCS thermal design flow and inlet temperature variations will be produced in the loops having asymmetric steam generator tube plugging levels however, coolant mixing in the
~
reactor vessel lower plenum attenuates these conditions. The zero power rod ejection cases are characterized by a sharp increase in the clad average temperature. Asymmetric considerations should be addressed for the hot zero power (HZP) rod ejection cases since these cases are performed using RCS thermal design flowrates which are based on two-out-of-three RCPs operating. Asymmetric steam generator tube plugging levels could adversely impact the corresponding UFSAR results if the loop having the highest thermal design flow (i.e., the loop with the lowest steam generator tube plugging) is not assumed to be in operation. RCS thermal design flow would be further reduced in this scenario since flow is being maintained in the two loops having more steam generator tube plugging. However, the sensitivity which addressed a 2% reduction in thermal design flow showed only a slight increase (< 1%) in the snaximum clad average temperature. There exists enough margin in the Beaver Valley analyses to absorb the differences created from the reduction in thermal design flow.' Therefore, the approximate 1.5% reduction in
~
thermal design flow caused by the increased steam generator tube plugging as well as the possible additional reduction caused by the asymmetric tube plugging would not produce a significant increase in the Beaver Valley maximum PCT. WPF1802D:lD051793 6-23
i Re analysis of the peak pressure transient for the RCCA ejection event is discussed in WCAP-7588, Rev.1. A reduction in RCS How could increase the primary side pressurization by reducing the l primary-to-secondary side heat transfer, llowever, due to the rapid nature of this event any secondary l side heat removal will lag well behind the heat addition to the primary side. Bus, an approximate l l 1.5% Cow reduction will have a minimal impact on the primary side peak pressure, liowever, in . WCAP-7588, several cases are presented which calculate the peak RCS pressure. De most detailed of these cases calculates a peak pressurizer pressure of 2600 psia. This is sufficient margin to the Faulted l Condition Stress Limits to accommodate an approximate 1.59 reduction in the RCS How. RCS ! thermal design flow and inlet temperature variations will be produced in the loops having asymmetric steam generator tube plugging levels; however, since the analysis was done at full power, coolant mixing in the reactor vessel lower plenum attenuates these conditions. , i Based upon the preceding discussions, an approximate 1.5% reduction in RCS flow and the : asymmetric tube plugging effects do not result in the violation of the beensing basis criteria following a RCCA ejection event, and the conclusions of the UFSAR remain valid. 6.4.18 Reactor Coolant Pump Shaft Seizure (l.ocked Rotor) (UFSAR Section 14.2.7) i Tius Condition IV event is analyzed under full power conditions assuming the instantaneous seizure of , i one RCP rotor using the LOFTRAN and FACTRAN computer codes. His results in a rapid RCS Dow reduction and pressure rise which may lead to DNB. De reactor is promptly tripped on a low flow signal. The analysis indicates that the maximum reactor coolant system pressure is less than the : limit value, the maximum fuel clad temperature is less than 2700*F and the amount of zirconium-water reaction is small. The lower RCS flow will result in slightly higher system pressures than those calculated in the current UFSAR analysis. He PCT analysis performed for the locked rotor event conservatively assumed that DNB occurs upon the initiation of the event. DNB significantly decreases fuel-to-clad heat transfer. : This assumption maximizes the calculated peak clad temperature and minimires the impact of a Dow reduction since fuel-to-clad heat transfer is already substantially degraded. Additionally, the asymmetric tube plugging scenario may impact the analysis because the locked rotor could occur in
~
the loop with the lowest number of plugged tubes (thus the greatest fraction of RCS total Dow). Derefore, this event was analyzed to incorporate the lower RCS flow and the asymmetric tube ! l plugging condinons. l l WPF1802D.tDe1793 6-24 l l l
i Some of the key assumptions used in this analysis include: j
- a. Three reactor coolant loops were operating prior to the pump seizure in one loop. l b, To account for asymmetric tube plugging effects, the loss of flow occurred in the loop having the lowest level of tube plugging (i.e., the highest loop flowrate)
- c. The initial power level was 102% of the rated thermal power.
- d. In the peak pressure analysis, the initial system pressure was assumed to be 42 psi greater than the nominal RCS pressure, and the initial temperature was 4.5'F greater than the nominal RCS ;
average temperature. ,
- e. The low flow reactor trip setpoint was 87% of nominal.
l The results of this analysis met the criteria stated above and the allocation of generic DNB margin ! ensures the percent rods-in-DNB is less than 18%. The conclusions of the UFSAR with respect to the locked rotor event are met for the increased steam generator tube plugging as well the 5% loop flow asymmetry. Figures 6.4-5 through 6.4-8 depict the transient response to the locked rotor event. 6.4.19 Complete Loss of Forced Reactor Coolant Flow (UFSAR 14.2.9) I This Condition III event is analyzed under full power conditions assuming that 3-out-of-3 operating reactor coolant pumps coastdown. The reactor is assumed to trip on an undervoltage signal. The analysis indicates that the minimum DNBR remains above the safety analysis limit value. Steam generator tube plugging potentially affects the complete loss of flow event by decreasing the RCS flow rate assumed in the analysis. A decrease in the RCS flow rate potentially decreases the minimum , DNBR calculated during the event. While the reduced TDF is a DNB penalty, the resulting slower I flow coastdown would be a small offsetting benefit. The flow coastdown of an underfrequency event is not impacted by the thermal design flow change because the event is driven by the RCP response to the frequency decay.
. i The asymmetric steam generator tube plugging levels discussed in the introduction cause flow and inlet temperature asymmetries between the RCS loops. RCS inlet temperature asymmetries will be attenuated by coolant mixing in the reactor vessel lower plenum and the nominal RCS average temperature will be controlled to the same value which was previously assumed in the UFSAR -
F wPF1802t11 dos 1793 6-25
- analysis. With asymmetric loop flow, the core flow basically remains unchanged relative to a uniform plugging case.
o De reduedon in RCS flow would impact the minimum DNBR. llowever, the core limits used in the calculation of the minimum DNBR are more limiting than the core limits based on the current design basis. And, since the transient conditions are not altered by the TDF reduction, the conclusion diat the DNB design basis is met remains valid and no allocation of generic DNB margin is required. Thus, the DNBR remains above the safety analysis limit value and the conclusions from the UFSAR analysis also remain valid. His event was reanalyzed, not because of the 30% SGTP or TDF reduction, but due to an issue that has been brought to the attention of the NRC regarding the circuitry associated with the underfrequency and undervoltage reactor trip functions upstream of the Westinghouse Solid State Protection System. Specifically, the cabinets containing the switchgear are not electronically separated. They are powered by an " energize to actuate" power supply, and they are not scismically qualified. Herefore, a single initiating event, such as a fire or seismic condition, could potentially result in a common mode failure of the undervoltage and underfrequency reactor trips such that they would not be available for protection. He complete loss of flow event was reanalyzed taking credit for only the low primary coolant loop flow reactor trip. De analysis also assumed the reduced thermal design flow (261,600 gpm) and 30% steam generator tube plugging. He allocation of generic DNBR margin is sufficient so that all of the applicable acceptance criteria for this event as given in the Beaver Valley UFSAR are met. 6.4.20 Sincie Rod Cluster Control Assembly (RCCA) Withdrawal at Full Power (UFSAR 14.2.10) Two cases are analyzed and presented in the UFSAR for this Condition 111 event: automatic and manual reactor control. In both cases, an increase in core power, reactor coolant temperature and hot channel factor produce a reduction in the minimum DNBR. He analysis indicates that, although it is not possible in all cases to ensure that DNB will not occur, an upper bound on the number of fuel rods experiencing DNB is less than or equal 10.5% Steam generator tube plugging potentially impacts the single RCCA withdrawal event by reducing the RCS flow which is assumed in the analysis. De reduction in RCS flow would impact the minimum DNBR. Ilowever, the core limits used in the WPFIB02D:1D051793 6 26
.I t
calculation of the minimum DNBR are more limiting than the core limits based on the current design i basis. And, since the transient conditions are not altered by the TDF reduction, the conclusion that the I DNB design basis is met remains valid and no allocation of generic DNB margin is required. Asymmetric steam generator tube plugging levels will result in flow and inlet temperature asymmetries between the RCS loops. Coolant mixing in the reactor vessel lower plenum, however, will minimize j these affects. Furthermore, the nominal RCS average temperature will continue to be controlled to l the same value previously assumed in the UFSAR. Derefore, the conclusions of the analysis discussed [ I in UFSAR remain valid for the increased tube plugging. i 6.4.21 Steam I.ine Break Mass /Enercy Release - Inside/Outside Containment l t Various steam line break cases are analyzed for the purposes of generating mass and energy release ; rates which are then applied to containment response or compartment environmental analyses. Cases i t are performed assuming various break sizes and initial power levels. Four major factors influence the [ release of mass and energy following a steam line break. Rese are steam generator fluid inventory, i protection system operation, state of the secondary fluid blowdown, and pritaary to secondary heat [. transfer. He RCS flow reduction due to the increased steam generator tube plugging levels would not l. ~ significantly affect these factors. Steam generator tube uncovery would occur slightly sooner. A lt decrease in RCS flow would tend to reduce the primary to secondary heat transfer, thereby reducing ; i the steam pr ssure and temperature during normal operation. Any reduction in the secondary side .l temperature and pressure would tend to lessen the mass and energy released during a steamline break l t event. As a result, an approximate 1.5% reduction in RCS flow would not adversely affect the steamline break mass / energy releases. Dese statements are supported by the discussion in j WCAP-10961, Revision 1 (Reference 6.4.5). -
?
I Asymmetric steam generator tube plugging levels would have the same effect described above (if the ; faulted S/G had a large amount of plugged tubes) or be bounded by the assumptions in the current l analysis (for the S/G with less than average tube plugging). l Therefore, the conclusions of the current steamline break mass / energy release calculations are l 1 ~ considered to be applicable for the reduced RCS flow and the asymmetric tube plugging scenario. j I w!T160::D.lD 2 1793 6-27 , I i l
V .l l l l 1 6.4.22 Se point Impact l The impact of asymmetrical steam generator tube plugging levels on the Beaver Valley Unit i non-LOCA accident analyses have been presented; consideration of the affect of this increased steam generator tube plugging and the asymmetry on the Overpower and Overtemperature reactor protection i functions is addressed below. l The coefficients for the setpoint equations are calculated using (among other inputs) the core thermal ; limits, RCS thermal design flow and the expected steam pressure at full power, nominal conditions. Each of these parameters changed as a result of the increase in average steam generator tube plugging l to a level of 30%. ) i lhe asymmetric steam generator plugging levels create flow and inlet temperature asymmetries j between the RCS loops. However, each channel is being used to determine the AT in individual ; coolant loops under specific loop inlet and outlet conditions. Since only one channel exists in each , loop, AT's may vary from loop-to-loop; however, the K-terms in the overtemperature AT setpoint , equation (i.e., K1, K2 and K3) and in the overpower AT setpoint equation (i.e., K4, K5 and K6) will remain constant for all three loops. Since the AT setpoints are based upon a fraction of the individual ; 1 loop AT's and the loop channels are individually calibrated based upon the loop temperatures, the i OTAT/OPAT setpoints discussed above will continue to be valid under the asymmetric conditions considered in this evaluation. 6.4.23 Non-I.OCA Conclusion ; i The impact on the non-LOCA licensing-basis analyses of plant operation with an increased average tube plugging level of 30% a reduced TDF of 87,200 gpm per loop, and 5% asymmetric loop flow ( has been examined. To suppon this evaluation, new analyses, as well as the evaluation of existing ! analyses have been performed. ! The evaluation indicates that this operation will not have a significant adverse impact upon the non- . LOCA licensing-basis analyses. In addition, the steamline break mass & energy release rates inside , and outside of containment also remain valid. The impact of asymmetric steam generator tube plugging on the overpower and overtemperature f reactor protection functions has also been considered and it has been determined that the current wPrisozoanm51793 6-28
y B U setpoints for these functions provide adequate protection to the core limit lines. Therefore, all licensing-basis criteria continue to be met and the conclusions in the UFSAR remain valid. 6.4.24 Non l.OCA References 6.4.1 Beaver Valley Unit 1 UFSAR 6.4.2 G. H. lieberle, ed., "LOFTRAN Code Description and User's Manual," WCAP-7878, Revision 5. November 1989. 6.4.3 C. liunin, P. W. Robertson, ed., *FACTRAN Code Description," WCAP-7337, Revision I-P, j December 1989. l t 6.4.4 D. II. Risher, Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized . Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1, December 1971, ) 6.4.5 J. C. Butler, D. S. Love, "Steamline Break Mass / Energy Releases for Equipment ! Environmental Qualification Outside Containment," WCAP-10961. Revision 1, October 1985. ; i f r a a i e I t
.- i f
- I i
1 i wmsominesm3 6-29 j l
\
6.5 STEAM GENERATOR tulle RUITURE (SGTR) EVALUATION l , 6.5.1 Evaluation i The Steam Generator Tube Rupture (SGTR) analysis in the Beaver Valley Unit 1 UFSAR was performed to evaluate the radiological consequences due to die event. The major factors that affect the radiological doses for an SGlR event are the amount of radioactivity assumed to be available in . ! the reactor coolant, the amount of reactor coolant transferred to die secondary side of the faulted steam generator through the ruptured tube, and the amount of steam released frorn the ruptured steam generator to the atmosphere. For the UFSAR analysis, it was assumed that the primary to secondary break flow and the steam release from the faulted steam generator would be terminated within 30 minutes after the accident. We loss of reactor coolant due to the break flow is assumed to result in reactor trip and Si actuation due to low pressurizer pressure. After reactor trip and Si actuation, the break flow rate is assumed to reach equilibrium at the RCS pressure when the incormng Si flow rate equals the outgoing break flow rate. The equilibrium break flow rate is assumed to persist until 30 minutes after the initiation of the , accident. The total primary to secondary break flow is then determined for the 30 minute period. We - amount of steam released from the faulted steam generator is calculated based on a mass and energy balance for the RCS and the steam generators for the 30 minute period. An evaluation has been completed for a reduced Thermal Design Flow (TDF) of 261,600 gpm (87.200 gpm/ loop), widi a maximum 5'7c flow asymmetry and up to 30% steam generator tube plugging to determine the impact on the UFSAR SGTR analysis. he conservative fuel failure assumption of 1% defective fuel for the Beaver Valley Unit i SGTR analysis will not change due to the reduced TDF and increased stearn generator tube plugging. De reduced TDF, assumed flow asymmetry and increased steam generator tube plugging will change the steam generator operating parameters which will affect the break flow prior to reactor trip and also the steam release from the faulted steam generator. Ilowever, the amount of radioactivity released to the . atmosphere for the Beaver Valley Unit 1 SGTR was conservatively calculated independent of the amount of steam released from the faulted steam generator, and thus, the SGTR consequences are primarily dependent upon the primary to secondary break flow. I wnno:D mosim 6-30 1
l i i i i 6.5.2 Results f . The Unit i SGTR analysis was evaluated for the reduced TDF of 261.600 gpm, a maximum flow asymmetry of 5% and the increase in steam generator tube plugging level to 30%. Tin results of the ; evaluation indicate that reduced TDF, the flow asymmetry and increased tube plugging results in a ; slight increase in the calculated break flow and consequently in the calculated radiation doses for an ! r SGTR. However, due to the conservatism in the calculated results for the SGTR reponed in the s Beaver Valley Unit 1 UFSAR, the UFSAR results remain bounding. I 6.5.3 Conclusions t The conclusions presented for the SGTR analysis in the Beaver Valley Unit 1 UFSAR remain valid for l a Thermal Design Flow of 261.600 ppm, a maximum flow asymmetry of 5% and up to 30% steam { generator tube plugging. i t I i h q
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l I l 5 t
. I i
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}
u WPrig02D;1DM1793 6-31
v i 6.'6 LOCA EVALUATION : i De following UFSAR ^3 LOCA related events were evaluated for Phase 4,3 of the BVPS.130% ; SGTP Reduced / Asymmetric TDF Program"" .f i Large Break LOCA (LBLOCA)(UFSAR Section 14.3.2.2) , Small Break LOCA (SBLOCA) (UFSAR Section 14.3.1) .; Blowdown Reactor Vessel Forces (UFSAR Section 14.3.3 & Appendix B) Post-LOCA Long-Term Cooling, Suberiticality Evaluadon (related to UFSAR Section 14.3.2) f 1 Reactor Coolant Loop LOCA Forcing Functions (DLW-80-87'C^f) s The following UFSAR'lI LOCA related events are excluded from the evaluation-I Hot Leg Switchover to Prevent Potential Boron Precipitation /Long Term SI Verification - Not W Cognizance (as discussed in detail in Reference CC). Reactor Coolant Loop Stress Reconciliation - Not W Scope (as discussed in detail in BVPS-1 l i TDF Repon'CI). ; 6.6.1 Larce Ilreak LOCA (I BLOCA). , i LBLOCA Introduction i l ne BVPS-1 LBLOCA analysis of record, which is presented in the UFSAR '1. is a B ASil Evalutuion Model33 analysis with a Peak Cladding Temperature (PCT) of 2149"F. Phase 4.3 of the BVPS-130% ; SGTP/TDF Reduction Programt83, consists of the following central elements: (1) 30% Steam Generator Tube Plugging (SGTP) in any o'r all SGs.
?
(2) Reduced Thermal Design Flow (TDF) consistent with the 10 CFR 50.92 TDF Reduction Report'C3 (analysis limit of 261.600 gpm total RCS flow). i i weriamn;inesim 6-32
?
^]
+
.l l
i 1 (3) 5% Asymmetric Loop Flow (minimum analysis loop flow of 82840 gpm = 1 0.95 x 261,600/3). - l
.. l,
'Io evaluate the effect of these desired plant configurations, a panial break spectrum LBLOCA reanalysis was performed. -
Increased SGTP results in degraded primary to secondary side heat transfer and restricted loop steam venting. His can lead to higher predicted peak cladding temperature (PCT) during a LBLOCA i transient. As indicated in the TDF Repon, the effects of RCS loop ilow are generally not numerically significant for LBLOCA analysis. However the reanalysis was performed at the bounding RCS flow configuration most appropriate and conservative for the specified reduced / asymmetric RCS flow j conditions. i A complete description of the reanalysis results is presented as a proposed UFSAR section in l a Attachment A of this Report. i i l LBLOCA Methodolocy: IM
, CDCOM,3',REFLOOD tul, BASH"! and he computer codes used for this reanalysis were SATAN LOCBART, which, as designated in Reference 1, is a combination of LOCTAN and B ARY 1 ]
The Evaluation Model (EM) methodology employed in the analysis is generally documented in 1 References I, P and Q, although additional previously referenced generic study references are presented 1 in the Attachment A UFSAR markups. In addition, the analysis has incorporated a variety of l 10 CFR 50.46 ECCS EM updates associated with the BASH EM which are discussed in detail next. l l The following 10 CFR 50.46 EM updates were incorporated in this reanalysis: Previously documented EM Updates: 1989 Evaluation Model UpdatesM J 1991 Evaluation Model UpdateslF1 as follows- . Item 2.1 Fuel Rod Model 'l I Item 3.2 Burst / Blockage (Note 1) i wPrisozo.inc5n93 6-33
I W l 1992 Evaluation Model Updates!^R' as follows: . l Spacer Grid Heat Transfer Error in BART, WCAP-lO484 Add l ^ '. l tra 1992 Evaluation Model Updates as follows: l i Improved Code I/O, Diagnostics and General Coding Maintenance Extension to NUREG-0630 Burst / Blockage Correlation! 1 l Note 1: LBLOCA Burst / Blockage DLW 91-159 Item 3.2, As designated therein, this issue must be reviewed on a case-by-case basis since a coding change was not incorporated into the model. The IFBA runs reponed herein fall into the category where the results would be affected by the issue using the algorithm developed in DLW-91-159. However. detailed sensitivity study work and investigation into the transient results concluded that the issue is inherently not applicable to the limiting IFBA transient due to the different mechanisms involved in determining the PCT. De remaining non-limiting (i.e. non-IFB A) analysis results reported in 'Results' section reflect consideration of this issue. Deir transient results fall into the DLW-91-159 'not affected' category. EM Updates not previously documented: (1) Miscellaneous minor LOCB ART Error Corrections. His includes pellet / clad ' contact and clad thinning models which were included in the ulxlated code version used in the analysis. Dese errors were deemed of negligible effect for the analysis runs of this repon. (2) Various discretionary changes to input / output format and inclusion of code diagnostics. Dese changes do not affect results. In addition, the revised LOCBART noding configurationmwas utilized, although this is considered a clarification and not a EM change. t With respect to the recent Containment / Accumulator Temperature issue"0, the analysis was conservatively performed as follows. De analysis in part supplemented the model specified in Appendix A of WCAP-8339M in response to the detailed investigation reported for this issue. In general, the reanalysis was WPF1602D:IDAr51793 6-34
i performed in a more conservative manner than prior analyses. De containment / accumulator-input values utilized in the re-analysis were provided by DLCom based on BVPS-1 operating conditions and/or Technical Specifications. Hree inputs are associated with this issue (a) Accumulator Temperature. (b) Containment Pressure and (c) Containment Temperature. As designated in the UFSAR Markups, the high nominal accumulator temperature was 93*F and the containment temperature was taken as a nominal 95"F. The lowest contaimnent pressure , g was taken as 8.9 psia as constrained by the Tech Specs!"3 , Of these three input specifications, accumulator temperature is the most significant and the 93*F is a more conservative value than utilized in prior BVPS-1 analyses reported in the UFSAR. As stated in the Reference R,S study..Wl had previously typically set the accumulator ; I temperature at the value determined for containment temperature. Ecrefore, the analysis was performed more conservatively than the prior methodology would demand. This is not considered an evaluation model application change at this time. 9 The containment pressure specified is a more conservative value than necessarily committed to R in WCAP-8339, and is thus interpreted to not be an EM change. Its discussion is included herein only because of the relevance to the Reference R,S study. The containment nominal temperature model is consistent with the Reference R.S report which concluded that this parameter had only a small effect for BASH analysis, with the direction of-conservatism being analysis dependent. His containment temperature input specification dms supplements the lowest operational temperature model specified by WCAP-8339. in the previous 1987 Upflow Conversion BASH full spectrum analysis reported in UFSAR Section 14.3.2, the following cases were analyzed: Cd=0.4,0.6 and 0.8 minimum Si cases, and Cd=0.4 maximum Si case which was the limiting case for that plant configuration. The 20% SGF1 reanalysis of 1991 reanalyzed both Cd=0.4 cases with the result being reported in UFSAR 14.3.2.2 as _j the Minimum Si case being limiting. For the Phase 4.3 30% SGTP reanalysis, the Cd=0.4 and Cd=0.6 Minimum Si cases were reanalyzed. Re analysis results are reported in detail in the UFSAR Markup (attachment A), with the Cd=0.4 case being limiting. De thermal-hydraulic transient results of the Cd=0.4 Minimum Si case, in comparison with the prior UFSAR Max and Min Si transients, clearly indicated that the Cd=0.4 Maximum S1 - l WPrit(02D:lD051793 6-35
- l l
i s I would be non-limiting to a greater degree than reported in the 207c SGTP results. Therefore, this case did not require reanalysis. Instead, the Cd=0.6 case (next most limiting discharge coefficient case from the full spectnam Upflow Conversion UFSAR results) was repeated and remained non-limiting ' . with respect to the Cd=0.4 case, f
- i A detailed fuel study was performed to consider the various fuel types intended for inclusion in the l Cy 10 core design (presumes that Region 1&9 assemblies from the Cy 9 Core DesignMFl g,jgg g
[ discharged, that Region 12 will be introduced as Vantage 511 (V5H) w/o IFMs, and that another ; Region 1 Standard (STD) assembly from the spent fuel pit will be re-introduced): Region 10,11,12: l V5H w/o IFMs non-IFBA 275 psig backfill: Base Case used to determine limiting ; Cd configuration I i V5H w/o IFMs IFBA (100 psig & 200 psig backfills): Sensitivity Case for Cd=0.4 MIN SI ^ Region 1: STD non-IFBA 500 psig backfill: Sensitivity Case for Cd=0.4 MIN SI l The determination of the limiting Cd case was made based on analysis results using the 275 psig , V5H fuel Base Case. From that determination, the remaining fuel sensitivity cases were all performed using the limiting Cd=0.4 transient. The sensitivity analysis results showed that the limiting fuel type , was the 100 psig V5H w/o IFMs IFBA. Of panicular interest to this study was the incorporation of a j reduced, STD fuel specific LOCA assembly peaking factor (P-bar-HA). The P-bar-HA was selected based on the prior work performed for Phase 4.5 which is documented in the Cy 10 P-bar-HA , l Study'1 The Cy 10 P-bar-HA Study used a P-bar-HA value of 1.20 (provided by W Commercial Nuclear Fuel Division) and concluded for Cy 10 operation based on the 20% SGTP UFSAR analysis, [ that the STD fuel would remain non-limiting with respect to the V5H w/o IFMs that populates the , majority of the core Based on the large PCT margin seen in this study, the Phase 4.3 LDLOCA .! analysis utilized an increased P-bar-HA value of 1.25 for the STD fua.1 in order to allow for additional - future core design flexibility. : The V5H w/o IFMs fuel retains the P-bar-HA value of 1.46 designated ; WPF1402D:1DO51793 6-36 l
4 - I .5, . in the Cy 10 P bar-IIA Study. P-bar-IIA is not a Tech Spec /COLR value and therefore no liccasing action is required. The P-bar-II A values will be incloded as part of the Reload Safety Analysis 5 Checklist (RSAC) review performed each cycle as part of the Reload Safety Evaluation (RSE) program which adheres to the requirements of the E Reload Topical *l. LHl OCA Input Parameters Specific to the LOCA Analyses, a detailed E-DLCo interaction scquence to establish key input parameters was undertaken which cone;uded with the establishment of LOCA Reanalysis Parameters ff " Significantly included in this interaction was recalculation of S1 Performance and establishing Containment / Accumulator Temperature modelling and values (as discussed in ; Methodology).
'Ihe LDLOCA analysis was performed consistent with the DLCo specified LOCA ParametersM with the following clarification: .l In Parameter I.A. DLCo specified use of Technical Specifications (Tech Spec)
Amendment 165. This was specified prior to the TDF Report!G which included a Tech Spec TDF change. As stated in the Introduction, the analysis was performed at the bounding configuration consistent with the TDF Report, and thus the analysis is consistent with both the j I current Tech Spec"1 and that instituted by the TDF ReportIG, I It is highlighted that the reanalysis is consistent with both the current Tech Spec"1 and the BVPS-1 , Setpoint Methodology Document IU Significant key input parameters to the LDLOCA analysis are presented in Tables 14.3-2a and 14.3-3a l of the UFSAR Markup. The inputs associated with the Containment / Accumulator Temperature issue are presented in Table 14.3-3a and were discussed in detail in the Metimdology section. I The reanalysis was performed based on the DLCo Si Performance *3 In Reference U, DLCo identifics , that the revised Si performance has been generated accounting for the Issues designated in
- t DLW-91-218 1^G, which in turn encompastes the prior recent SI Issues of DLW-89-848 ^M and- j DLW 90-729 "1. ~Ihe complete list of issues included in DLW-91218 and therefore incorporated into j the analysis are:
wPFtis02D IDA151793 6-37
ECCS Runout Margin (DLW 91-218) Miniflow Operadon (1.E. Bulletin No 80-18 Note 1) RCP Seal Injection / Tech Spec Verificadon (DLW-89-848) Branch Line & Total System Resistance (DLW-89-848) Suction Boost (DLW-91-218 Attachment 3) Flow Measurement Uncertainty (DLW-90-729) Note 1: St Performance most conservatively assumes C11G/SI miniflow valve r~uains OPEN throughout transient; supporting a permanent resolution to the miniflow auto-isolation issue. Li1LOCA Results
- A concise presentation of the transient results for the 275 psig V511,100 psig V511 IFB A, and 500 psig STD cases is presented in the following table. 'nie : 00 psig V511 IFB A fuel was not specifically analyzed but was evaluated to be bounded by the results of the other V5H fuel types. 'Ihe 100 psig IFBA fuel was limiting with an analysis PCT of 2144*F. More detailed results including Tables and Figures is presented in the UFSAR Attachment.
Tnble of Analysis Results Cr.se Cd =0.4 Cd=0.6 V51I
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V511 IFBA V5H non-IFBA STD non-IFBA PCT 2144 2(M9 1891 1902 Local Zire 13.7 5.9 3.8 3.2 Avg. Zire El% <1% <1% <1%
- A new Cy 10 PCT Summary Table is enclosed (see Section 6.6.5) and has been prepared consistent with the new W 10 CFR 50.46 Reporting Topical *. DLCo retairis final cognizance of 10 CFR 50.46
, Reporting considerations. The prior Cy 10 PCT Table was documented in the Cy 10 P-bar-IIA StudyNIwhich is hereby superseded. In addition, the prior Cy 9 PCT Tables generated for the 1992 EM Update!'"' and the BASIIER licat Link Error ^4, I which were generated consistent with the wn isominuim 6-38
format of the 50.46 Topical, wu also considered in the generation of the new Cy 10 PCT Table. The !
- Table includes the following PCT changes which have been applied against the analysis results: j (1) A l'F PCT penalty has been assigned for the Seismic /LOCA Tube Crush issue. His issue was originally described in detail as item 3.3 of ta .991 EM Report C As pan of 1 the program,,W calculated a reduced BVPS-1 Tube Crush factor of 0.23% based on a j comparison of BVPS-1 configuration with a reference analysis. his represents significant margin in comparison to the generic 5% Tube Crush factor reflected in the .;
1991 EM Report. His is considered a safety evaluation of a less conservative plant i specific application of an analysis input term associated with a previously reponed EM ; Change. and not a 50.46 Reportable change. This becomes item A.2 on the new PCT sheet. , (2) A 2*F PCT penalty has been assigned for the increased RCS Tavg Instrument ; Uncertainties'^3 t (increased to 4.5'F). This PCT penalty was originally generated in the l RCS Tavg Uncertainty Evaluationi ^ and remcins applicabic to the reanalysis. This is considered a safety evaluation of an increase in uncertainties associated with an analysis . { input term, and not a 50.46 Reportable change. The becomes item A.1 on the new PCT ! sheet i (3) A 25'F PCT benefit has been assigned for the WREFLOOD Structural Heat Model- [ Error instituted in the 1992 EM Report'^"3, which remains applicable to the reanalysis. This becomes item B.1 on the new PCT sheet. l l t The cumulative LBLOCA PCT is thus 2144'F + l'F + 2'F -25'F = 2122'F. ; Other pertinent points associard with the new LBLOCA PCT Table. ; (a) Structuring consistent with the 50.46 Topical W . (b) De LBLOCA Power Shape issue (new PCT sheet item E.1) is a Cycle specific item
" that will be addressed during the Cy 10 core design and discussed in the Cy 10 RSE. ;
wmso:n loosim 6-39 I
i r h (c) De new PCT Table is applicahic to the intended Cy 10 fuel configuration and is not l f pertinent to Cy 9 operation. i
= ;
(d) The line item for fuel sensitivity studies performed in the P-bar-IIA Study are removed , since they have been superseded by the various fuel studies performed directly herein. j (c) ET-NRC-92-3770 Rod Internal Pressure (RIP) Issue (item E.2 on new PCT sheet). The issue has two aspects, as designated in the letter to the NRC. De RIP Uncertainty , aspect was briefly discussed in the Cy 10 P-bar IIA Study and appeared as item F.3 in , its PCT sheet. This aspect remains directly incorporated into the reanalysis results (PCT Sheet item 'A*) as designated herein. De Low Backfill Fuel issue is also directly -{ addressed by the reanalysis through direct analysis of this (IFBA) fuel type. In particular, the 100 psig backfill pressure fuel is limiting for this reanalysis and is the , PCT reponed as item 'A' of the new PCT sheet. ; (f) LOCBART pellet / clad contact and clad thinning models, and the discretionary 1/0 and i diagnostic changes (discussed in Methodology section). These are items D.1 & D.2 on i the new PCT sheet. f 4 (g) De recently transmitted Large Break LOCA BASIIER IIcat Link Errol *H present in the
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FSAR 20% SGTP Analysis is not present in the 30% SGTP Reanalysis. The ICT ! i Sheets in the B ASHER Heat Link Error transmittal included a PCT penalty for the error, and also a compensatory PCT benefit based on Cy 9 specific PharilA. Neither of these ! two items is pertinent to the reanalysis and thus they do not appear on the new PCT , sheet. ! LBLOCA Conclusion - P Conformance with the 10 CFR 50.46(b)(1,2,3) limit! vas maintained for LBLOCA. This analysis will l
. r be used as the basis for the Cy 10 RSE once DLCo confirms that it is considered to be their licensing basis analysis for Cy 10, as per the requirements stipulated in the Reload TopicalN. ; +
i i wPrino:D.IDet793 6-40 , l s
I i 6.6.2 Small Hreak LOCA (Sill OCA) { i l r SBLOC A Introduction The BVPS-1 Small Break Loss-Of-Coolant Accident (SBLOCA) analysis of record, vhich is presented , in the UFSAR 3^3, is a NOTRUMP Evaluation Model*"' analysis with a Peak Cladding Temperature (PCT) of 1802*F, PCT penalties to the analysis have been assigned as most recently documented in the 1992 Beaver Valley 10 CFR 50A6 Repon!** which indicates a cumulative PCT of 2197*F. l Phase 4.3 of the BVPS-130% SGTTVTDF Reduction ProgramSI, consists of the following central ; I clements: (1) 30% Steam Generator Tube Plugging (SGTP) in any or all Steam Generators. r (2) Reduced Thermal Design Flow (TDF) consistent with the 10 CFR 50.92 TDF Reduction , ReponlG (analysis limit of 261,600 gpm total RCS flow). , (3) 5% Asymmetric Loop Flow (minimum analysis loop flow of 82840 gpm = 0.95 x 261,600/3) j
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The 30% SGTPTTDF Reduction Program also addresses the various safety injection performance issues (references U and AC), incorporates numerous safety evaluations, and regains margin from the current . l SBLOCA PCT of 2197'F. ,
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The SBLOCA reanalysis investigates a full break spectrum. A description of the reanalysis results is presented as a proposed UFSAR section in Attachment A of this report. SBLOCA Methodolony - i I I The computer codes used for the reanalysis were NOTRUMP SG and LOCTA-IV!M. The Evaluation Model (EM) methodology employed in the analysis is generally documented in references BB and BD. i in addition, the analysis has incorporated 10 CFR 50.46 ECCS EM updates associated with the UOTRUMP EM. EM updates incorporated in this reanalysis: WPF1802D:1DA15I793 6-41 i l
l, i Previously documented EM Updates: ; l t i 1989 Evaluation Model Updates" -(
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1991 Evaluation Model Updates'" as follows: 1
~i Fuel Rod Initial Condition Inconsistency !
NOTRUMP Solution Convergence Reliability SBLOCA Creep Model , SBLOCA Rod Internal Pressure Assumption
-[
5 1992 Evaluation Model Updates'^"3 as follows: t Auxiliary Feedwater Flow Table { S.G. Secondary Side Modeling Enhancements i t Basis Change for Hot Assembly Rod Gap Pressure i EM Updates not previously documented: , i (1) Auxiliary Feedwater (AFW) Actuation on Low Pressurizer Pressure (LPP) Safety i Injection (SI). .! i The NOTRUMP EM sai assumes that AFW delivery actuates on the combination'of , Reactor Trip coincident with Loss-of-Offsite Power (LOOP). DLCo and 3' have.. j previously interacted (reference B A) on this issue and DLCo has instructed (reference T) j W to assume AFW actuation on the safety injection signal. Through manipulation of ) code inputs the analysis was performed accordingly. .; l (2) Various Discretionary Changes to Input / Output Format and inclusion of Code ; Diagnostics. These changes do not affect the results. j I
.l WPF1802D:1DK61793 6-42 i!
l
.l I
4 Summarv of Reanalysis Methodolony 1 i l In the previous 1987 Upflow Conversion NOTRUMP full spectmm analysis reponed in the UFSAR section 14.3.2. the 2 inch,3 inch, and 4 inch break sizes were analyzed. For the Phase 4.3 30% SGTP i reanalysis, a full spectrum of break sizes was reanalyzed. The full break spectmm included equivalent break diameters of 1.5,2,3,4, and 6 inches. A 6 inch break site was investigated because this break size is larger than the liigh llead Safety injection (111151) pipe diameter, The IlliSI spilling line ; 1 backpressure assumpdon varies with break size. For break sites smaller than the 1111S1 pipe diameter, i 1111S1 is assumed to spill to RCS pressure. For break sires larger than the 1111 51 pipe diameter,111151 l is assumed to spill to containment pressure (reference BB). The analysis results are reported in detail i in the UFSAR Markup of attachment A. j t The full break spectmm was completed at beginning of life (BOL) fuel rod condidons. 'Dtis analysis showed that the 3 inch break has the highest BOL PCT, but this break has a PCT which is only slightly greater than the PCT from the 2 inch break. The PCTs of the 1.5 inch,4 inch, and 6 inch t breaks were shown to be much less severe. Since the 3 inch break has the highest PCT at BOL, an evaluadon was performed to address the most limiting time in life. This evaluation accounts for burst j and blockage effects caused by increased rod pressure during burnup conditions. l 1 0 f For the 6 inch break analyzed, the standard intact loop seal clearing assumption was removed. The j typical loop seal assumption used in the SBLOCA analyses allows only the broken loop to clear and ; vent steam during the small break LOCA transient. Section 5-3-1 (pages 5-44 through 5-46) of j t reference BB, contairts a discussion of the loop seal modeling and of the restriction applied in i NOTRUMP. This section discusses the phenomena of a threshold break size for which larger breaks-will realistically vent steam through more than one loop seal, The typical use of the loop seal restriction in NOTRUMP ensures conservative behavior for break sizes below the threshold break size. Reference BB (Section 5-2-5); Table 5-2 2 shows results from a SBLOCA break spectrum and l l indicates that both the intact and broken loops vent steam for the 4 and 5 inch breaks. Reference BD , (Section 3.3.4) contains a 6 inch break sensitivity that showed both the intact and broken loops would clear and therefore the restriction could be removed for the 6 inch break. Based upon the information ; i found in references BB and BD. the removal of the intact loop seal restriction for the 6 inch case in .l this analysis is considered appropriate. l l l wPriso2n:Inm51793 6-43 l 5 t h
c t SBLOCA is not significantly affected by the mechanical geometry differences between Standard and ; VANTAGE SH (V5H) without IFMs fuel (reference BF), largely because grids are not currently ; modelled in the NOTRUMP EM. Modeling the V511 w/o IFMs fuel is appropriate to address the ; current Cycle 9 and planned Cycle 10 and beyond fuel.
~ i SBLOCA Input Parameters: ,
Specific to the LOCA Analyses, a detailed W-DLCo interaction sequence to establish key input , i parameters was undenaken which concluded with the establishment of LOCA Reanalysis Parameters M Included in this interaction were: < k (1) Recalculation of Si performance (2) Isolation of Main Feedwater l (3) Actuation of Auxiliary Feedwater Pumps (4) Recalculation of Steam Generator Safety Valve Flow The SBLOCA analysis was performed consistent with the DLCo specified LOCA Parameters" with the following clarifications: ;
. i i
In Parameter I.A. DLCo specified the use of Technical Specifications Amendment 165. 'Ihis '!. was specified prior to the Thermal Design Flow (TDF) ReportlO which included a Technical , Specificadon TDF change. As stated in the introduction, the analysis was performed at the TDF consistent with the TDF Report, and thus the analysis is acceptable with respect to the change instituted by the TDF Report'G. : l The reanalysis is consistent with the BVPS-1 Setpoint Methodology DocumenF and the Technical :
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- Specifications"I, except for the assuraed AFW actuation on LPP SI. As noted by DLCo in the LOCA }
parameter letteF, actuation of AFW off from LPP SI will be implemented as a technical specification j change. , The reanalysis was performed based on the DLCo S1 Performance M . In Reference U, DLCo identifies : that the revised SI performance has been generated accounting for the Issues designated in 3 DLW-91-218*G.' which in turn encompasses the prior recent SI Issues of DLW-89-84d*l and WPFIE02DJDK251793 6-44 j l I
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f DLW-90-729 ^El. nie complete list of issues included in DLW-91-218 and therefore incorporated into the analysis are: ECCS Runout Margin (DLW-91-218) Miniflow Operation (I.E. Bulletin No 80-18 Note 1) RCP Seal Injectionfrech Spec Verification (DLW-89-848) Branch Line & Total System Resistance (DLW-89-848) Suction Boost (DLW-91-218 Attachment 3) Flow Measurement Uncenainty (DLW-90-729) Note 1: SI Performance most conservatively assumes CHG/SI miniflow valve remains OPEN throughout transient; supporting a permanent resolution to the minillow auto-isolation issue.
'Ihe significant input parameters to the SBLOCA analysis are presented in Table 14.3-2b of the UFSAR Markup.
SBLOCA Results A concise presentation of the transient results for the 1.5 inch,2 inch,3 inch,4 inch, and 6 inch breaks is presented in the following table. More detailed results of the 15 inch,2 inch. 3 inch, and 4 inch breaks are included in the UFSAR Attachment. 'S 9 wrriso:o:iom1793 6-45
.. .- . . _ ~ - -
-i
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SMALL BREAK LOCA REANALYSIS RESULTS t 30% STEAM GENERATOR TUBE PLUGGING i i
-' BOL 500 MWD /MTU i
Break Site (Inches) 1.5 2 3 4 6 : Peak Cladding Temperature (*F) 782 1748 1750 1275 774 I A new PCT summary Table is enclosed (see Section 6.6.5) and has been prepared consistent with the i new Westinghouse 10 CFR 50.46 Reporting Topicalm, DLCo has already received the 199210 CFR 50.46 Report ^"1. which adhered to the new Reporting Topical. DLCo retains final j cognizance of 10 CFR 50.46 Reporting considerations. W considers the PCT summary sheet j l associated with this analysis to supersede the UFSAR analysis, and will discontinue PCT tracking against the UFSAR analysis. .
~
The new cumulative SBLOCA PCT for this analysis is 1766*F. !
.i Re most recent SBLOCA PCT Table provided in the 199210 CFR 50.46 Report l^"I was reviewed to ,
1 ensure that specific PCT assessments were incorporated into the reanalysis. Specific items of this i Small Break PCT Margin Utilization Table are discussed in detail below as warranted. (a) The prior permanent ECCS Model Assessments of 37*F (Item B), were all addressed by ) using the current SBLOCA methodology in the SBLOCA analysis. l (b) The 10 CFR 50.59 and 50.92 PCT assessments (Table A) were incorporated by using a ? 24.5'F RCS tediperature uncertainty and the Auxiliary Feedwater Flow rate specified in. reference T. i 1 Items E.1 and E.2 of the prior PCT Table total to 214*F of margin utilization. These i (c) items were revised to reflect the 15'F applicable to a 1751*F beginning-of-life PCT. Also, the LOCTA II.A. Pressure aspect of these line items has been directly. incorporated by the updated SBLOCA Evaluation Model. wPFl*02D IDM1793 ' 6-46 !
I (d) Item F.1 was addressed by using auxiliary feedwater actuation off from the low < pressurizer safety injection signal. (c) The SBLOCA analysis uses a P-bar-IIA of 1.46 which returns the Cycle 9 specific use of margin in item F.2. De combined LOCA plus Safe Shutdown Earthquake (SSE) Scismic Loads will result in a small amount of Steam Generator Tube Collapse. his SG Tube Collapse issue, reported to DLCo in , reference F, has been addressed through an evaluation yielding a l'F permanent PCT assessment. l The SG Tube Collapse l'F permanent assessment, plus the 15*F temporary time in life assessment yield a 1766*F SBLOCA PCT. ;
.i SBLOCA Conclusion Conformance with the 10 CFR 50.46(b) (1), (2), and (3) limits is maintained for the SBLOCA. His analysis will be used as the basis for the Cyc!c 10 RSE once DLCo confirms that it is considered to be their licensing basis analysis for Cycle 10, as per the requirements in the Reload TopicalN . a 6.6.3 Illowdown Reactor Vessel and 1,000 Forces ne Reactor Vessel LOCA forces conclusions are currently presented in WCAP-ll556'C83 Blowdown forces are typically limiting immediately after the break, and are influenced primarily by Design Tcold.
Design Teold decreases for the 5% Asymmetric RCS Flow and 30% SGTP condition and LOCA > I forces increase, liowever, the increase is accommodated within the available break size margin. The LOCA Forcing Functions assumed in WCAP-Il556 are based on a 120sq in break area. However, the actual predicted break opening is 65sq in. The increase in Forcing Functions due to the Tcold decrease are offset by taking credit for break opening area down to the actual 65sq in. The overall structural integrity conclusion of WCAP-Il556 remains valid. It is noted that this Phase 4.3 . Evaluation actually incorporates directly the previous evaluations performed for the TDF Report 3C1 and the RCS Tavg Uncertainty increase EvaluationA. i i
}V records indicate that a variety of Blowdown Loop Forcing Function (LFF) and integrated Reactor l Coolant Loop (RCL) Loads / Displacement evaluations were provided by }V, to DLCo in the past. This ;
WI'F1800D 1Do51793 6 47 ,
i IG historical sequence is reflected upon in detail in the TDF Report In response to the request for DLCo coordination in section 3.5.1 of the Program *, DLCo indicated that the desired evaluation for the BVPS-1 SGTP Program is to address the LFF provided in 1980. In 1980, W provided DLCo the Reactor Pressure Vessel Outlet Nozzle (RPVON) break LFF via DLW-80-87" His case assumed a 144in2 opening area at the direction of SWEC. W functional group internal records indicate that shortly thereafter, the complete 11 break spectrum analysis was completed. Acrefore, W will directly comment on the DLW-80-87 RPVON LFF and the remaining 10 breaks that DLCo may have received information for in 1980. In both of these cases, SWEC and/or DLCo would have cognizance of the complete Loop Stress Reconciliation. l l LFF are primarily influenced by RCS temperature, break size and break opening time. The RPVON break is governed by RCS Bot, which increases as a result of the TDF Reduction, but could increase due to the 309 SGTP/5% RCS flow asymmetry aspects. Herefore the RPVON LFF of DLW-80-87" increases by 2.0% for the RPVON break. For the remaining 10 break locations, a bounding 2.7% increase in LFF is imposed by the complete Phase 4.3 & Phase 4.2 scope (30% SGTP, Reduced TDF consistent with the TDF RepottlG, and 5% Flow Asymmetry) which causes a decrease in RCS Teold. It is clarified that this 2.7% increase is with respect to the original analysis results, and not an additional 2.7% above and beyond the increase of .35% evaluated for the Phase 4.2 TDF Report. (Alternately viewed as .35% for Phase 4.2 and an incremental 2.35% for Phase 4.3 by itself). Some of the breaks in the 11 break spectrum were run assuming full double-ended main reactor coolant piping areas. DLCo might be able to offset the LFF increases of the Phase 4.3 program for these cases with break size margin if (for example) Leak-before-Break has been approved for the RCS Loop piping. 6.6.4 Post-l.OCA Lonn-Term Cooline, Suberiticality Evaluation The Westinghouse position for satisfying the requirements of 10 CFR 50.46(b)(5) 'Long Term Cooling
- is defined in WCAP-8339"M, WCAP-8472A, and Technical Bulletin NSID-TB-86-OSICN.
The )V, commitment is that the reactor will remain shutdown by borated ECCS water alone after a LOCA. Since credit for the control rods is not taken for a LBLOCA, the borated ECCS water
~
provided by the accumulators and the RWST must have a concentration that, when mixed with other sources of borated and non-borated water, will result in the reactor core remaining suberitical assuming all control rods out. De asymmetric RCS flow aspect of Phase 4.3 does not influence the calculation wPFl*02D:1Dc51793 6-48
to any degree. The 30% SGTP aspect of the program can influence RCS mass directly and also by vinue of influence on RCS Initial Temperature conditions. The temperature aspect is negligible for ; this program, since no change in RCS Tavg is mposed. The reduction in RCS volume / mass j associated with increased SGTP to 30% is an unquantified benefit to the calculation since the RCS is a j
~
dilution source. Therefore, the existing calculation remains bounding. This conclusion is checked by , . i l W on a cycle by cycle basis at the time of the RSE. most recently the Cy 9 RSE*D. i 6.6.S Small and Larce Hreak LOCA PCT Tables , t PCT Sheets in the WCAP-13451AFormat are contained on the next 3 pages. 6.6.6 Section 6.6 LOCA References r A. BVPS-1 UFSAR Rev 10,1992 l. B. Technical Descriptions for DLCO P.O. D-109601 (SGTP Program), consisting of: B.l. DLW-92-147, 'DLCo P.O. D-109601 W PG-51657;' Simmers @ to Heck (DLCo). 9/28/92. (TD for Ul/U2 SGTP Program). 1 B.2. DLW-92-296, 'BVPS Unit 1 Large Break LOCA Analysis Scope Clarification, for the SG'IP Program (Unit 1);' Simmers (W) to Halliday (DLCo), 11/17/92. ; i C. DLW-92-378, 'DLC BVPS Unit 1 RCS Reduced Thermal Design Flow Report;' Simmers @ l to Siegel (DLCo), 12/17/92. ! D. NS-NRC-89-3463, '10CFR50.46 Annual Notification for 1989 of Modification in the E ECCS Evaluation Models,' Johnson @ to Murley (USNRC),10/5/89. l l
- i E. DLW-90-Sil,'DLCo BV Units 1&2 Reporting of ECCS Evaluation Model Revisions'; )
i Steinruetz @ to Noonan (DLCo) 1/2/90. O F. DLW-91-159. 'DLCo BVPS Units 1&2 ECCS Evaluation Model Changes,' Steinmetz @ to Noonan, (DLCo). 6/20/9 L WPF1802D;1D1051793 6-49
')
l-
==ms._.=- _ ====4W10ReaLPeakfla d.Ternperature .Marginytili1? tio n. _.
Prenous PCT Inlormation: DLW-H-2021992 EM Updates t Plant Name: Beaver Valley Unit i Eval. Model: NOTRUMP Fuel: l~t t' w o IFMs Utility Name: Duquesne Light Company FO = 2.40 Fall = 1.62 S( i l P = 10 P - ha r - II A = 1.46 Ref e rence * (' lad Temperature Notes A. AN Al YSIS of RECORD i1992 309 SGTP Program) l PCT = l'50 'F l
- 1 St i lube (' rush DI.W-vi- 159 #33 Plant Specific 1 APCT= 1 'F 2 it PRIOR PERMANENT ECCS MODEL ASSESSMENTS
- 1. None APCT= 0 'F :
C 10 CFR 50.59 to CFR 50.92 EVALUATIONS Table A APCT= 0 "F D. lu9110 CFR 5()46 MODEL ASSESSMENTS - NONE APCT= 0 'F tPermanent Assessment of PCT Margm) E. I EMPORARY ECrS MODEL ISSUES"
- 1. Limiting Time in Life in SHLOCA (Hase 1751'F) 8 APCT= 15 "F F. OTilER MARGIN Al. LOCATIONS
- 1. None APCT= 0 *F ,
LICENSING BASIS PCF + MARGIN ALLOCATIONS PCT = 1766 *F
-
- References for the Peak Clad Temperature Margin Utilization summary can be found in Table B.
" It is recommended that these temporary PCT allocations which address current LOCA model issues not be considered with respect to 10 CFR 50.46 reporting requirements.
Notes: 1 Analysis 'A' incorporated prior EM changes as follows: , (1989) DLW-90-511 (Ref 3) 1 (1990) DLW-91-049 (Ref 4) (1991) DLW-91-159 (Ref 7) (1992) DLW-93-202 (Ref 8) 2 Not 10 CFR 50.46 Reportable v f l 1 WPF1802D;lDC50st 6 50 l i i a
. . _ _ -_._Large Breakycak Clad Temperature Margin z _ Utilization _
Preuous PCT Information: DLW-41-2021492 EM Update tassociated with a prior analysisj and DI W-43-214 HVPS- 1 HAS11 Error 10CFR50.46 Notification (associated with same prior analysts > Plant Name: lleaver Valley Unit 1 LM: HASil . Fuel: 17tl7 V511 w o 11 Ms A STD Utility Name: Duquesne Light Company FO = 2.40 Fall = 1.62 St iTP= W , Pharil A = 1.46; Limiting Break Case: ('d=0.4 Min Sl; V411 w o IFMS !Fil A Applicability commences with () 10 Re f e re nce
- Clad Temperature Notes A. ANAL BIS OF RECORD (lV42 30G S(iTP Program) 1 PCT = 2144 T 2 I R('S Iavg Uncertaintv (DLW 42- 140) Evaluation 1.5 APCT= 2T 1
- 2. St i l ube Crush DLW-41-159 *1.3: Plant Specific 1 APCT= 1T 1
- 11. PRIOR PERMANLNT EC('S MODEL ASSESSMENTS
- 1. WRFFI OOD Structural Metailleat Model lA APCT= -25 T C 10 CI'R 40.54 to Cl R 50.92 EVALUAT10NS Table A JPCT= 0 'F SONE D.144110 CFR 50.46 MODEL ASSESSMENTS (Permanent Assessment of P(7 Margin)
- 1. LOCHART Models: Pellet Clad
Contact:
Clad Thinning 1 J PCT = 0 'F 6
- 2. Discretionary Code l'O ' Diagnostics ! Maintenance ! JPCT= 0 "F 6 E. TEMPORARY ECCS MODEL ISSUES" I. Large Break Power Shape APCT= TBD'F 1
- 2. Rod laternal Pressure [ET-NRC-92-3770 (Ref 13)l Issue ! JPCT= 0 *F 5 F. OTilER MARGIN ALLOCATIONS 1.NONE APCT= 0T 1.lCENSING HASIS PCT + MARGIN ALLOCATIONS PCT = 2122 'F
- References for the Peak Clad Temperature Margin Utilization summaries can be found in Table B.
" It is recommended that these temporary PCT allocations which addess current LOCA modelissues not be considered with respect to 10 CFR 50.46 reporting requirements.
Notes:
- 1. Power Shape issue [DLW-90-676 (Ref 9) DLW-91-159 (Ref 7) #3.1) to be addressed during Cy 10 RSE.
- 2. Analysis 'A' incorporated prior EM changes as follows:
(1989) DLW-90-Sil (Ref 3) (1990) DLW-91-049 (Ref 4)
-(1991) DLW-41-159 (Ref 7) #2.1 Fuel Rod Models only. . (1992) DLW-93-202 (Ref 8) Item HART Grid Errors [WCAP- 10484 Add I (Ref 11)] only.
ET-NRC-92-3746 (Ret 12) NUREG-000 Burst / Blockage Extension
- 3. A.! & A.2 NOT50.46 Reportable.
. 4. deleted
- 5. E.2 locorporated into Analysis 'A': includes 'BOL RIP Uncertainty' PCTSheet item from DLW-93-202 (Ref 8).
- 6. D I & D.2 Incorporated into Analysis 'A'. The model change is described in Reference 1.
t WPF1800D:t omwn 6-51
, y i
. m- .m m . . m = -~= Ja b.le 3210 CFRJ049 /10.CFR40J(Evaluations . .
Plant Name: 11 caver Valley Unit I Company: Duquesne Light Company Reference (' lad Temperature Notes if nt ALL HREAK Errs SAFETY EVALUAT10NS--NONE TOTAL 10 CFR 50.59 /10 CFR 50.92 SMAl.L HREAK ASSESSMENTS APCr= 0 *F .,
- 11. LARtiE HREAK ECUS SAFETY EVALUATIONS -- NONE l
TOTAL 10 CFR 50.59 /10 CFR 50.92 LARGE BREAK ASSESSMENTS APCT= 0 *F f Notes: 1
}
Table 11 - Refer.ences __
- l. 'TlilS LETTER' for 30ri SOTP Program Phase 4.3 !
- 2. deleted !
. i
- 3. DLW-90-511,'DLCo BV Units I&2 Reporting of ECCS Evaluation Model Revisions', ;
Steinmetz (W) to Noonan (DLCo).1.2.90. .
- 4. DLW-91-049. 'DLCo BV Units l&2 ECCS Evaluation Model Reporting Update', ;
- Steinmetz (W) to Noonan (DLCo),3/5/91. .l S. DLW-93-071. 'DLCo BVPS Units 1&2 RCS Tavg and Flow Uncertainty, Simmers (W) to Siegel(DLCO). I 27 93 l
- b. NDIMNE:6312. 'DLCo 11V Unit 1 increased SGTP and Reduced TT)F Program,'
flalliday (DLCo) to Simmers (W), 10/12/92 (LOCA Parameters). '
- 7. DLW 150,'DLCo BV Units I&2 ECCS Evaluation Model Changes',
Steinmetz (W) :o Noonco (DLCo),6/20,91.
- 3. DLW-93-202,'DLCo BV Units l&210CFR50.46 Notification and Reporting Information', ;
Hall (W) to Tonet (DLCo).1/29/93. ;
- 9. DLW-91-009. 'DLCo BVPS Unit i Reduced Auxiliary Feedwater Pump Capacity SECL', i Steinmetz (W) to llalliday (DLCo),1/17S 1.
- 10. deleted ~ -l t 1. WCAP- 10484-P Addendum 1,' Spacer Grid lleat Transfer Effects During Reflood,' Shimeck,12S2.
- 12. ET-NRC 3746, Liparuto (W) to NRC Documeat Control Desk,' Extension of NUREG-0630...l 9/16/92.
i
- 13. ET-NRC-92-3770, Liparulo (W) to NRC DCD,' Interim Report . Pursuant to 10 CFR 21.21(a)(b),' 11/13N2.
)
i WPFito:D.!DO50r43 6 52 l 1 i
G. ET-NRC-92-3746, ' Extension of NUREG-0630 Fuel Rod Burst Strain and Assembly Blockage : Models to High Fuel Rod Burst Temperatures,' Liparulo (W) to NRC Document Control Desk, 9/16/92. f'
- 11. BVPS-1 Technical Specifications through Amendment 165 i
- 1. WCAP-lO266-P-A Rev. 2 WCAP-11524-NP-A Rev 2, "Ihe 1981 Version of the f Westinghouse ECCS Evaluation Model Using the basil Code', Kabadi, J. N., et al., March 1987; including Addendum l-A ' Power Shape Sensitivity Studies
- 12/87 and Addendum 2-A
- basil Methodology improvements and Reliability Enhancements' 5/88. .
t J. NS-NRC-90-3524, ' Additional Information on use of BASH Code,' Johnson QV) to Jones ! (NRC RSB NRR),7/20/90. t K. WCAP-8302 P, WCAP-8306-NP, SATAN-VI Program: ' Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant,' Bordelon, F. M., et al., June 1974.
. i r
L. WCAP-8327, WCAP-8326, ' Containment Pressure Analysis Code (COCO),' Bordelon, F. M., and E. T. Murphy, June 1974. M. WCAP-8170-P, WCAP-8171 NP, ' Calculational Model for Core Reflooding after a ; Loss-of-Coolant Accident MReflood Code),' Kelly, R. D., et al., June 1974. N. WCAP-8301-P, WCAP-8305-NP, 'LOCTA-IV Program: Loss-of-Coolant Transient Analysis,' Bordelon, F. M., et al., June 1974. O. WCAP-9561-P-A, WGAP-9695-NP-A, 'BART-1 A: A Computer Code for the Best Estimate Analyzed Reflood Transients,' Young, M., et al., ~1984; including Addendum 3 Rev 1 (7/86). , P. WCAP-9220-P-A, WCAP-9221-NP-A, ' Westinghouse ECCS Evaluation Model, 1981 Version,' Rev.1 Eicheidinger, C.,1982. Q. WCAP-8339-NP, ' Westinghouse ECCS Evaluation Model-Summary,' Bordelun, F. M., II. W. Massie, and T. A. Zordan, July 1974. ! WPTIkO D:lDM1793 ! 6-53 i
i l R. ET-NRC-92-3699, 'Results of Technical Evaluation of Containment Initial Temperature f Assumptions for Large Break Loss of Coolant Accident Analysis', Liparuto @ to NRC ! Document Control Desk,6/1/92. ; S. DLW-92-095, 'DLCo BV Units 1&2 Containment Initial Temperature Assumption for Large Break Loss of Coolant Accident Analysis;' Steinmetz @ to Noonan (DLCo) 6/4/92. , T. NDlMNE:6312, llalliday (DLCo) to Simmers @, DLCo BV Unit 1 Increased SGTP and Reduced TDF Program, 10/12/92. (LOCA Parameters) l t U. NDIMNE:6340, Halliday (DLCo) to Simmers @, DLCo BV Unit 1 Increased SGTP and Reduced TDF Program, 11/12/92. (ECCS Performance /PCWG Footnote Confirmation). f I V. DLW-91-155, 'DLCo BVPS Units 1&2 Steam Generator Tube Plugging Analysis Program Engineering and Licensing Report'; Steinmetz @ to Halliday (DLCo),6/5/91 as supplemented by DLW-91-249,'DLCo BVPS Units 1&2 Revision 1 to WCAP-12966 Steam l Generator Tube Plugging Analysis I'rogram Engineering and Licensing Report'; Steinmetz @ f to lialliday (DLCo). 12/12/91.
?
W. DLW-92-410, 'DLC BVPS Unit 1 Large Break LOCA P-bar-HA Sensitivity Study Report;' Simmers @ to Siegel (DLCo), 12/28/92. i i X. WCAP-9272-P-A/WCAP-9273-NP-A, 'W Reload Safety Evaluation Methodology,' S. L. j Davidson and W. R. Kramer (Ed.),7/85. l 1 Y. WCAP-11419-P/WCAP-11420-NP, '.W_ Setpoint Methodology For Protection Systems Beaver Valley Unit 1,' W. He Moomau,10/87. .; l Z. WCAP-13451-NP, *W Methodology For Implementation of 10CFR50.46 Reporting,' J. S. Ivey
& M. Y. Young,10/92. l l
AA. DLW-92-140,'DLCo BVPS Unit 1 RCS Flow Uncertainty Calculation;' Simmers @ to Halliday (DLCo),9/23/92. wPFiso:D:1D/051793 6-54 i i
j AB. Not used.
, AC. DLW-91-218, 'DLCo BVPS Units 1&2 ECCS Pump Runout Limit Issues'; Steinmeti @ to Halliday (DLCo), 9/27/91.
L i j-AD. DLW 89-848,'DLCo BV Unit 1.W ECCS Flow inconsistencies'; Steinmetz @ to Noonan (DLCo),12/4/89. l AE. DLW-90-729,'DLCo BV Units 1&2 ECCS Flow Measurement Potential Discrepancies'; i Steinmetz @ to Noonan (DLCo),9/25/90. , 1 AF. 91DL*-G-0012. 'DLCo BVPS Unit 1 Cycle 9 RSE'; McKenzie @ to Ireland (DLCo), 3/26/91. AG. WCAP-1(484 Addendum 1; ' Spacer Grid Heat Transfer Effects During Reflood,' D. J. Shimeck,12/92. Ali. DLW-93-202, 'DLCo BV Units 1&210 CFR 50.46 Notification and Reporting Information/
~
Hall @ to Tonet (DLCo),1/29/93. A1. DLW-93-219, 'DLCo BVPS Unit 1 10 CFR 50.46 Evaluation of Large Break LOCA basil Metal Heat Link Error,' Simmers @ to Tonet (DLCo) 3/15/93. AJ. DLW-93-071. 'DLCo BVPS Units 1&2 RCS Tavg and Flow Uncertainty;' Simmers @ to Siegel (DLCo),1/27/93. BA DLW-92-238, 'DLCo BV Unit 1 Small Break LOCA Evaluation for Auxiliary Feedwater Pump Actuation,' Simmers @ to Halliday (DLCo), 10/27/1992. B B. WCAP-10054-P-A/WCAP-10081-NP, ' Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,' Lee, N., Rupprecht, S. D., Tauche, W. D., & Schwarz, W. R, 8/85, w111so2D:1D/051793 6-55
i BC. WCAP-10079-P-A/WCAP-10080-NP, 'NOTRUMP, A Nodal Transient Small Break and ~ General Network Code,' Meyer, P. E., 8/85. t t
.o B D. WCAP-lll45-P-A/WCAP-11372-NP, ' Westinghouse Small Break LOCA ECCS Evaluation i Model Generic Study With the NOTRUMP Code,' Rupprecht, S. D., Osterrieder, R. A., Wills, l . t M. E., Willis, J. M.,10/86.
B E. ET NRC-91-3647, ' Interim Repon of Evaluation of a Deviation or Failure to Comply Pursuant ' to 10 CFR 21.21(a)(2),' Westinghouse to NRC Document Control Desk, 12/20/91. j r B F. 89DL*-G-0033,'DLC BVPS Unit 1 Plant Safety Evaluation Report ' Mckenzie @ to Ireland l (DLCo), 4/24/89. [ l CA. DLW-80-87, 'DLCo BVPS Unit 11.nop Thrust for RPV Outlet Nozzle,' Noon @ to l VanWassen (DLCo),7/22/80 ! t CB. WCAP-11556-P, ' Reactor Pressure Vessel and Internals System Evaluation for the Beaver ; 4 Valley Unit i Upflow Conversion,' R. R. Laubham j CC. 89DL*-G-0067, 'DLCo BVPS Unit I Cycle 8 RSE', McKenzie @ to Nass (DLCo),8/18/89 l CD. WCAP-8472. ' Westinghouse ECCS Evaluation Model: Supplementary Information', Bordelon, i F. M., et. al.,4/75. !
-l:
CE. DLW-86-670, 'DLCo BV Unit 1 Technical Bulletin NSID-TB-86-08 Post-LOCA Long-Term , Cooling: Boron Requirements'; Triggiani @ to Lacey (DLCo),12/1/86. . CF. DLW-93-071,'DLCo BVPS Units 1&2 RC.3 Tavg and Flow Uncertainty;' Simmers @ to i Siegel (DLCo),1/27/93. j
.i o
I h WPFikO2D;1D051793 6-$6
6.7 ASSESSMENT OF UNREVIEWED SAFETY QUESTIONS Operation of Beave ' Valley Unit I with a steam generator tube plugging level of up to 309, a TDF reduced to 261,600 gpm and 5% loop flow asymmetry has been evaluated using the guidance of NSAC-125 and does not represent an unreviewed safety question based on the following justification.
- 1. Will the probability of an accident previously evaluated in the FSAR be increased?
No. The probability of these events previously analyzed in the UFSAR will not increase due to the increased tube plugging. An assessment of the NSSS primary and auxiliary conoonents, including the reactor pressure vessel system, reactor coolant pump, steam generator, pressurizer, Control Rod Drive Mechanisms, and Capped Latch Housing, concluded that the integrity of the components will be unaffected by the increase in steam generator tube plugging level to 30%. Also, evaluations of the NSSS Fluid Systems and Balance of Plant Systems concluded that the 30% SGTP level will not adversely impact the adequacy of the systems. Steam generator tube plugging level
+
does not adversely impact steam generator tube integrity. Because the changes do not involve a significant impact on the heat transfer capabilitics of the steam generators, the
~
change will not result in any additional challenges to the plant equipment. -Therefore, the probability of a SGTR event previously evaluated in the UFSAR will not be increased. Also, since the reactor coolant pressure boundary integrity and system function are not adversely impacted, the probability of occurrence of an accident evaluated in the UFSAR will be no greater than the original design basis of the plant.
- 2. Will the consequences of an a'cident previously evaluated in the FSAR be increased?
No. Since safety design requirements continue to be met and the integrity of the reactor - coolant system pressure boundary is not challenged, the assumptions employed in the . calculation of the offsite radiological doses remain valid. Therefore, the consequences of the accidents considered in the Beaver Valley licensing basis remain unchanged.
- 3. May the possibility of an accident which is different than any already evaluated in the FSAR be increased?
WPFis02D;1DA)51793 6-57
No. All aspects of the proposed steam generator tube plugging have been evaluated, and no new or different accidents or failure modes have been identified for any system or component important to safety. Nor has any new limiting single failure been identified. Because the change does not affect the integrity of the steam generator or any other equipment,it can be determined that an accident which is different than any already evaluated in the FSAR will not be created as a result of this change.
- 4. Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
No. De proposed tube plugging increase has identified no new single failures. De evaluations concluded that no additional challenges to equipment imponant to safety have been identified. %e change in the steam generator tube plugging level is not expected to challcoge the integrity of the equipment assumed to be operable. The small change in the vessel / core inlet temperature has a negligible effect on the analyses of the pressure boundary compcments. An assessment of the NSSS primary and auxiliary components concluded that the integrity of the components will be unaffected by the increase in steam generator tube plugging level to 30%. As stated previously, since
~
equipment and system design criteria continue to be met, it is judged that the probability of a malfunction of equipment important to safety will not be increased.
- 5. Will the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?
No. The change in the steam generator tube plugging level will not challenge the integrity of the equipment assumed to be operable. All safety analysis acceptance criteria continue - to be met. Since safety design requirements continue to be met and the integrity of the reactor coolant system pressure boundary is not challenged, the assumptions employed in the calculation of the offsite radiological doses remain valid. Derefore, the consequences of the accidents considered in the Beaver Valley licensing basis remain unchanged.
- 6. May the possibility of a malfunction of equipment imponant to safety different than any already evaluated in the FSAR be created?
~ WI11so:D.1DO51793 - 6-58
g . . _ _ 4' l r
?
i t No. Since safety design requirements continue to be met and the integrity of the reactor f coolant system pressure boundary is not challenged, no new failure modes are created. All safety-related equiprnent is expected to perform as designed. The ability of any j safety-related equipment to perform its intended function is not affected.
- 7. Will the margin of safety as defined in the bases to any technical specification be reduced? l i
i No. The evaluation of the accident analyses considered in the licensing basis indicated that . l the analysis acceptance criteria will continue to be met with the increased level of steam generator tube plugging. All acceptance criteria continue to me met: i
= DNBR > 1.3 ;
- PCT < 2200*F (LOCA) f t
- System Pressure < 110% of design i
= PCT < 2700*F (non-LOCA) l
= RCS pressure boundary is not breached
. Stresses are within acceptable limits :
= Containment pressure < design :
t 4. - The margin of safety with respect to primary pressure boundary is provided, in part, by the safety factors included in the ASME Code. Since the components remam m l 6 compliance with the codes and standards in effect when Beaver Valley Unit I was .l originally licensed, the margin of safety is not reduced by the increased steam generator i tube plugging level. l l Therefore the margin to safety determined for the Unit I licensing basis safety analyses . remains unchanged. 4
6.8 CONCLUSION
-l Based upon the information provided herein, it can be concluded that the increase in the level of steam generator tube plugging (to a maximum of 30%), with a maximum loop flow asymmetry of 5%. does ,
not constitute an unreviewed safety question, provided the overall thermal design flow of 261,600 gpm is maintained, wrriso2D:tD451793 6-59
3 i r a 7.0 NUCIIAR FUEL EVALUATION An evaluation of the effects of the increased steam generator tube plugging to either the 30% or the TDF limit on the fuel design was performed with respect to the nuclear design, the thermal-hydraulic j design and fuel rod performance. 7.1 CORE DESIGN ; De core design evaluation indicated that the steam generator tube plugging level results in no impacts to the core design. i For a discussion of the P-bar-IIA margin issue, please refer to Reference 9. 7.2 TilERMAL.flYDRAULIC DESIGN ,
.t As the Thermal Design Flow was reduced to 261,600 gpm through the steam generator tube plugging -
design activity, a review of the design bases indicated the continuing applicability of all assumptions and design parameters, resalting in the continued acceptability of existing core limits, axial offset .I
~
limits and limiting values for Departure from Nuclear Boiling Ratio when considering the increased .fr tube plugging level. 7.3 FUEL ROD PERFORMANCE Fuel performance evaluations are performed for each fuel region to indicate that the fuel rod design criteria will be satisfied for all fuel in the core under the specified operating conditions. Evaluations r of the effect of the proposed design parameters on meeting the fuel rod design criteria were performed ! i for Beaver Valley Unit 1 Cycle 10 design, which is the first anticipated occurrence of such conditions. Based on the evaluations, the effect of increased steam generator tube plugging on meeting the fuel l rod design criteria will be documented. "Dtis will be confirmed for all future fuel regions as part of. , the cycle specific reload safety evaluation process. i i WPfikO2D:lDO3170) 7.] ;
r r
- 8.0 NSSS/ItAI.ANCE OF Pl; ANT INTERFACE EVAI,UATION i The interface between the NSSS and the Balance of Plant systems and components has been reviewed to assess the effects of increased allowable tube plugging level to the 30% or the revised TDF limit, and considering loop-to-loop flow asymmetry. This evaluation considered the effects of the increased j tube plugging on the interface requirements for the following Balance of Plant systems: !
I ne Main Steam System l De Condensate and Feedwater System ! ne Auxiliary Feedwater System ; he Steam Generator Blowdown and Sampling System i t The review utilized the power capability parameters as presented in Secdon 2.1. It is significant to note that the parameters which affect Balance of Plant systems either do not change, or change in a ! favorable direction. De full power steam and feedwater mass flows decrease slightly. Reactor power ; remains the same (2652 MWt). The final feedwater temperature remains at 437.5'F and the full power
- steam pressure remains at 760 psia. De zero load steam generator temperature remains at 547'F, and i
this shows that the zero load steam generator pressure will remain at the present 1020 psia value. This 1 evaluation indicated that the evaluation performed as pan of 1990 effort (20% SGTP) apply to the j 30% SGTP, reduced TDF conditions. Please refer to WCAP-12966. Section 8.0 for a discussion of , i the NSSS/ Balance of Plant Interface Evaluation. j
'l
[
-[
Il
'Ia
+-
,j l
i
'i e
WPF1802DilD051793 8-1 .l l l
9.0 REFERENCES
- 1. WCAP-11591," Beaver Valley Unit 1 10 Percent Steam Generator Tube Plugging Licensing Report " Y.A. Jen, et.al., September 1987
- 2. WCAP-12966, " Beaver Valley Power Station Units 1 and 2 20 Percent Steam Generator Tube Plugging Analysis Program Engineering and Licensing Report," R. C. Surman, Revision 1 November 1991
)
- 3. letter DLW-92-378, " Beaver Valley Power Station Unit 1 Reactor Coolant System Reduced j Thermal Design I' low Repon," G. D. Simmers @ to 11. M. Siegel (DLC), December 17, 1992
- 4. letter DLW-92-089, " Beaver Valley Unit 2 Reactor Coolant System Reduced Thermal Design Flow Report," J. N. Steinmetz @ to K. E. Italliday (DLC), May 29,1992
- 5. letter ND3NSM:5925. " Technical Specification Change Nos. I A-208/2A-74," T. P. Noonan (DLC) to J. D. Sieber (DLC), February 18,1993
- 6. letter DLW-92-140, " Beaver Valley Power Station Unit 1 RCS Flow Uncenainty Calculation,"
G. D. Simmers @ to K. E.11alliday (DLC), September 23,1992
- 7. IBVT 1.6.1, RCS Flow Measurement, Issue 2, Revision 9, Data following 8R
- 8. facsimile transmission, "BVPS 1 Steam Generator Tube Plugging Status," R. A.11 ruby (DLC) to T. J. Gerlowski @, April 8,1992
- 9. letter DLW-92-410. " Beaver Valley Power Station Unit 1 Large Break LOCA P-bar-IIA Sensitivity Study Report,", G. D. Simmers @ to }{. M. Siegel (DLC), December 28,1992
- 10. letter DLW 93-ll3, " Reactor Coolant System Reduced Thermal Design Flow Report,",
G. D. Simmers @ to IL M. Siegel (DLC), February 11,1993 WPI ikO2DJDM1793 9-1
- - _ _ _ _ _ _ - _ _ - _ _ - - _ _ _ _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ - _ _ _ - _ = _ - _ - _ _ _ - - _ _ _ - _ _ - _ - _ _ -
I 11.- W. T. 'Ihomson and M. V. Barton, "The Response of Mechanical Systems to Random Excitation", Journal of Applied Mechanics (TRANS ASMEL June,1957, pp. 248-251. f i
- 12. Combustion Enj;ineering, Inc. Report CENC-Il83, " Analytical Repon for Duquesne Light !
t Company Beaver Valley Power Station Unit No.1 Reactor Vessel," July 1972 i
- 13. Combustion Engineenng, Inc. Report CENC-1183-Al, " Addendum I to Analytical Report for Duquesne Light Company Beaver Valley Power Station Unit No.1 Reactor Vessel," t November 1973 I i
- 14. Combustion Engineering, Inc. Report CENC-1183-A2, " Addendum 2 to Analytical Report for ;
Duquesne Light Company Beaver Valley Power Station Unit No.1 Reactor Vessel," Octobe 1974 I
- 15. Westinghouse Report MED-PCE-10710, " Evaluation of the Reactor Vessel Effects of 20 l Percent Steam Generator Tube Plugging - Addendum to Combustion Engineering Report :
r Number CENC-1183 for Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel," July 1991 l
- 16. Westinghouse Equipment Specification 676413. Revision I for General Reactor Vessel and - !
Addendum Equipment Specification 678801, Revision 2 for Beaver Valley Power Station - Unit No.1 Reactor Vessel :
- 17. WCAP-9404 " Study of Reactor Vessel Upper Head Fluid Temperature," R. H. McFetridge )
and D. C. Garner, February 1978 18.- letter DLW-93-071, " Beaver Valley Power Station Units 1 and 2. Reactor Coolant System T,,, and Flow Uncenainty," G. D. Simmers QV) to H. M. Siegel (DLC), January 27,1993 l o
- 19. letter DLW-92-147, Duquesne Light Company P. O. D-109601," G. D. Simmers (W) to F. J. Heck (DLC), September 28,1992 (original offer letter for the integrated Unit I and
~" Unit 2 analyses and evaluations to support increased levels of steam generator tube plugging and reduced thermal design flowTate) WPF1802D:lDel793 9-2 l l
~. .., .
7-. . i NOTE: References used in the NON-LOCA and LOCA sections of Section 6 are included in those sections. ; t i
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I APPENDIX A : i
, 30 PERCENT STEAM GENERATOR 1UBE PLUGGING ANALYSIS PROGR AM l UFSAR UPDATE [
FOR ; BEAVER VALLEY POWER STATIONS -; UNIT 1 i
.i LOCA ,.
non-LOCA (Complete Loss of Flow) non-LOCA (Partial Loss of Flow) j non-LOCA (Locked Rotor) ; i f t
?
i t f 4
-{
g. l
.t -;, ' 5 l
l winsaminosim A-1 .j i 1 I
+
a l 1 i
-1 I
I LOCA' UFSAR Page Changes ; P l Pages Ficures ! 14 3-1 14 3-2 through -7 l 14 3-2 143-9A through -9C l 14 3-3 143-10A through -10C -; 14 3-4 143-11A through -11C ! 14 3-5 143-120 through -12U 14 3-6 14 3-131 through -13L l 14 3-7 143-14K through -14Q i 14 3-8 143-24D 14 3-9 i 14 3-10 l 14 3-11 l Tables
-t . 143-1c ,
143-Id i 143-le 143-1f , 143-2a 143-2b 143-3a l 143-4b 143-4c f 6 i I
BV?S-1-LPCATED FSAR Rev. ;0 (1/92; l 14.3 LOSS OF COCLANT ACO:OENT 14.3 presents the analyses and evaluations Of 5ecti:n Less-of-C0olant Accidents (LCCA). Section 14.3.1 discusses tne 1:ss of reactor ecolant from small ruptured pipes or fron cracks
;n large pipes which actuates the E=ergency Core Cooling System.
This type of fault is classified as a Condition III occurrence, sy definitien, Condition III occurrences are faults which may i occur very infrequently during the life of the plant. They will be accer.nodated with the failure of only a small fraction of the fuel rods although sufficient fuel da= age might occur to preclude resumption of the operation for a considerable outage time. The release of radioactivity will not be sufficient to interrupt er restrict public use of those areas beyond the exclusion radius. A Condition III fault will not, by itself, generate a condition IV fault or result in a consequential loss of function of the Reacter
* -r'er .e' qu C e T q Coo _lan ( 5 s y containmen*
4.. e reak R ctor C lant Sy. em P1 Rup ure yte'shs vn o ring he L Table '4.3-1
'{ ,
N.l Section 14.3.2 discusses the Major Reactor Coolant System Pipe Ruptures up to and including the Double Ended Rupture of the largest pipe in the reactor coolant' system. This type of accident has been classified as a Condition IV occurrence. Condition IV occurrences are faults which are not expweted to take place, but pestulated because their consequences would include the are potential for the release of significant amounts of radioactive material. These are the most drastic occurrences which must be designed against and represent limiting design cases. Condition IV faults are not to cause a fission product release to the environment resulting in an undue risk to public health and safety in excess of guideline values of 10CTR Part 100. A single Condition IV fault is not to cause a consequential loss of
, required functions of systems needed to cope with the fault including those of the E=argency Core Cooling System and the containment. It fied in 10CFR50.46 and 10CFR50AppendixK.ggybeenanalyzedass pur4h/ 5 p e C 4r* 2.
Section 14.3.2.2 discusses -1 :---- - reana s,Ts formed for increased Steam Gener 'Tucu % ging. "' results presented in 14.3.2.2 supersede those presented in 14.3.2. The previous full spectrum analysis presented in 14.3.2 is being retained for historical perspective. Section 14.3.3 prasents an analysis of the core and reactor internals integrity following the LOCA. The analysis of thyroid and whole body doses resulting from the LOCA appears in Section 14.3.5. Core activities which form a basis for these calculations are presented in Appendix 14B. Sections 5 and 6 also include discussions of the systems ,
~
contributing to limiting of radioactivity releases from the containment during a LOCA. 14.3-1
i r SVPS-1-CPCATED TSAR Rev. 13 ,;f3;;
;4,3,1 irss Of Reacter Ceclant frem Small Muetured Pires er Fre Cracks in Larce Pires Which Actuates E?ercerry re , C:oline System
{ Identificatien ef causes and Accident Descrirtien : t t
- A less of ecolant accident is defined as a rupture of the Reacter ,
c clant System piping or of any line connected to the system. See ; Section 4 for a = ore detailed description of the less of reacter - coolant accident boundary limits. Ruptures of small cross section ! will cause expulsion of the coolant at a rate which can to ! accommodated by the charging pumps which could maintain an l operational water level in the pressurizer permitting the operater to execute an orderly shutdown. The coolant which would be released to the containment contains the fission products existing in it. < The maximum break size for which the normal makeup system can !
=aintain the pressurizar level is obtained by comparing the l calculated flow from the Reactor Coolant System through the i postulated break against the charging pump makeup flow at normal -
Reactor Coolant Systas pressure, i.e., 2,250 psia. A makeup flow i rate from one centrifugal charging pump is typically adequate to ; sustain pressurizer level at 2,250 psia for a break through a ; 0.375 inch diameter hole. This break results in a loss of ! ,- approximately 17.5 lb per second. l Should a larger break occur, depressurization of the Reactor Coolant System causes fluid to flow to the Reactor Coolant Systes
. from the pressurizar resulting in a pressure and level decrease in ,
the pressurizar. Reactor trip occurs when the pressurizer low pressure trip setpoint is reached. The Safety Injection System is actuated when the appropriate setpoint is reached. The ! consequences of the accident are limited in two ways:
- 1. Reactor. trip and borated water injection complement void formation in causing rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay.
- 2. Injection of borated water ensures sufficient floodin of l' e to prevent _axcess_ive clad temperatures.
tha G 7:171; g # *\ Ion ,. 1 Lid C B i-M etie e' 0=425 'en
;;t
" ;1 p. N ;i:::ipt\;" M.eeftru cirrr !
yiTQ \'" g- t*-3::="'at *" OM 5-la-- :t h tion D nirient i, '
,' 8 ?ie me1 any maine ib pri-- te n= actor ~ toolane my=* r Ae-=ura s raucMe i
. r-hp r- \ ::1 ;t mq_. . e.f ay - i 7"*y ==- d". ;.z1
.e m a.m.. - D is-assure' a . (4.;. C : Pr au c 1snt syee gen mum;ie.. %h##s '
Before the break occurs, the plant is in an equilibrium condition, i.e., the heat generated in the core is being removed via the I t i 14.3-2
37?S-1-CPCATED TSAR Rev. ;; .; 32. sec=ndary system. Ouring blewdown, heat frc= decay, het internals and na vessel centinues to be transferred to the React:r C: lant System. The heat transfer between the Reacter Coolant System and ~ the secondary system =ay be in either direction depending en :ne relative temperatures. In the case of continued heat addition to the secondary system pressure increases and steam dump may eccur. Make-up to the secondary side is aut0matically provided by the
, aux 11iarv f tafdtalsr pu=ps. _
,w ,-- -
/ , ,
fCV tri; P%.%Ciritf)'(Pr%fs Wh ?;e01 e c gme,1 ; qu , ' The :::c cr eigns. incident _;i** f e f f e it(_p_cuer4 steps mor=al feedwat C flow cy closing the main f eedwater TIRiii isolation valves and initiates emergency feedwater flew by starting auxiliary feedwater pumps. The secondary flow aids in the reduction of Reactor C olant Systa= pressure. When the RCS depressurizes to 600 psia, the accu =ulators begin to inject water into the reactor coolant loops. The reactor coolant pu=ps are assumed to be tripped at the initiation of the accident and effects of pump coastdown are included in the blowdown analyses. Analysis of Effects and consecuences Method of Analysis Tor breaks less than 1.0 ft 2, the NOTRUMP I1) digital co=puter code is employed to calculate the transient depressurization of the Reactor Coolant System as well as to describe the mass and enthalpy of flow through the break. Small Break LOCA Evaluatien Model For loss-of-coolant accidents to small breaks less than 1
, square foot, the NOTRUMP Igug) computer code 'is used to calculate the transient depressurization of the RCS as well as to describe the mass and enthalpy of flow through the break. The NOTRUMP computer code is a state-of-the-art one-dimensional general network code consisting of a number of advanced features.
Among these features are the calculation of thermal non-equilibrium in all fluid volumes, flow regine-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes and regime-dependent heat transfer correlations. The NOTRUMP small break LOCA emergency core cooling system (ECC valuation model was developed to determine the RCS response to basis small 4 "~ break LOCAs and to address the NRC conc ressed in NURIG-0611, " Generic Evaluation of Teodwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants". In NOTRUMP, the RCS is nodalized into volumes interconnected by flowpaths. The broken loop is modeled explicitly, with the intact
- loops lumped into a second loop. The transient behavior of the ! i e
14.3-3 i l l
)
SVPS-1-tJPCATED TSAR Rev. ;; :p 32; systam is mass, energy, determined and :::entum fr:s the geverning conservation equattens :f applied throughout the system. detailsd description of A 1 and 2. ne FCTRUMP code is provided in References i use of i The
- na representation NOTRUMP of in the analysis involves, among other things, i with the reactor core as heated control volumes '
neight ancalculation. associated bubble rise model to permit a transient mixture The multinode capability of enables the program system an components. explicit and detailed spatial representation of various : of the behavior of the In particular, it Icop seal during enables a proper calculation accident. a loss-of-coolant Safety injection flow rate to the ! function of the system pressure is Reactor Coolant System as a i The Safe used as part of the input. ; the RCS, ,as ; A; Injection seconds System after_ the (sis) was assumed to be delivering to
- signa 1. L _y,
; - . generation of a safety injection Qggy q Tor these SI injection flanalyses, N' the nyntas delivery considers J u lad f:;i C f i: fig-.. ~',2--0. :: 1 f _ :::en 1
/_ i
' 7........ ...._ fi; = ::,. .....:e inj ;;.i:n fle" f rer *i_ 5I --
;- ; c., 2.
r ::
,. 1r f __- r:
d- ' - d g__d:d ;:-t frr de N esel i- ".. M:neJ'D7 second delay incluaes time requires for
' startiup and Ioadine of the safety iniection pumps onto the i
g ey _ buses. A _ ;f f:= ef _-
- "::f : ;- , 01:~_ i: -2 g -. ...... eG - m ; ~ eine: 't i- e.'.;t:0f Md pr::s; Z'= b i n :r - - ' .. n3 i' W Also
;2-- t i;.e &* 1^= - -- f *h r '
--d
^
i minimum Safeguards Emergency i. ore _..;ir.t cooling eenei'Sys N capability and operability has been assumed in these analyses. PegI IV clad temperature analyses are performed with the LOCTA v code which uses the RCS pressure, fuel rod power history, steam flow past the uncovered part of the core and mixture height cry calculated by the ode. I Ranults
- +
" # ' g -+ Reactor Coolant System Pipe Breaks j
[Mc r+ a , p-_ - N -"1*= ef $ 18 8 M 7 he^^t -i;; i=
- e= v . z.. .
c 1--;
.===_.
r 2; i: 2 M 2 di n: er:: r * +: t-ni - 12 3 "_ . _ - During of the earlier part of the small break transient, the effect the break ! flow is not strong enough to overcome the flow
, maintained by the reactor coolant pumps through the core as they are coasting down following reactor trip. Therefore, upward flow through the core is maintained.
the fuel rod and clad The resultant heat transfer cools to very near the coolant temperatures as long as the core remains covered by a two phase mixture. The maximum hot spot clad temperature transient is Dyeet
- F, ' 21- calculated during the
- f i.., m ":r: If fr i f: :M :ti" -
!;25D 14.3-4
S*7ps-1-Cp;ATID T5AR 71 e v , ; . , 3. ; j a 21: = : d -FM : -- - - - +. The peak clad temperature transien
.s snown in Ticure I Q for !
E ;; ;th *t:
. ;- -_ i p : = 01 ^.the 6 .}worst p gj break si:e d n m The* steam fi:V rate l . 'f :r tne worst breax is sn51.u vu i m re 14.3-5. *'han the mixture belcw the top of the core, the steam flew ce=puted in level drops '
Peterence 1 provides ecoling to the upper portion of the cere.
- ne heat transfer =ceffielents for this phase of the transient are
. given in Figure 14.3-6.
w=rst creak is snown in Figure 14.3-7. The het spot fluid temperature !ce tne i Tne =cre power (dimensionless) transient following the accident ! (relative to reactor scram ti=e 's shown in Figure 14.3-8. i 4.'7 The reacter shutdcwn time trip signal ti=e
& seconds) is_ecual to the reacter (2.0 seconds) plus 5semei.3 seconds for red ,
insertion. During this rod insertion period the reactor is conservatively assu=ed to operate at rated power. ; Cenelusiens g alyses presented in this section show rti' 'nf d
- g the Emergency Core Cooling System,., 4 K;O.;r _it r m ;t::: s'.i previ;is s sufficient core flooding to Q::
' " - ,1:rl;t d p : 2 -1 C j h,er W nig d :: _ :! i)isrequired protection limits of 10CTR50.46 q ence, adequate afforded by the Emergency Core Cooling System in the ,
. event of small break loss-of-coolant accident.
(b)(I),(2), an (1), ' Additional armak sires h M ~ - -# ! Additional break sizes were analyzed. Figure (14.3-9A,and 14.3-98,a M h present the RCS pressure transient for the jand 4Jin d breaks ! respectively and Figures 14.3-10A fand 14.3-10STyresents the volume history (mixture ht ots for 4eehpraaks.
. temperatures to 4 ase are less than the Thepeak peak clad clad f p ,i tem erature of the 3 Tnch break. The peak clad temperat3ra %
;- qMFE __
/Neases are given in Figures $3-11A and 14.3-115,ad N,3-hc.
f'or 14.3.2 Maier Rameter coolant Svetan Pine Ruetures (Lena of coelant Accident) HJ-\Sh! The analysis specified by 10CTR50.46(5), " Acceptance i Cri aria for Energency Core Cooling Systems for Light Water Power Reactors", is' presented in this section. 'The resu
- the loss-of-coolaat accident analysis are shown in Table . nd show compliance with the Acceptance Criteria. The cal techniques used are in complianen with Appendix K of 10CTR50 and are described in References . Reference 59 describes the incorporation of -
630 nurst and blockage into the 1981 Evaluation Model. c,tg' u.( ry pa A,aA -- spe.c+m e Section 14.3.2.2 discusses aanal f erformed for increased Steam Generator ~hbu M Iugging. n.a results presented in 14.3.2.2 supersede those presented in 14.3.2.1. 1 8 14.3-5
3'<*?S-1-UFOATED FSAR Rev. *. f;, 3; The coundary considered for less-of-coolant accidents as relatec Oc c:nnecting piping is defined in Sect:0n 4. Should a major break occur, depressurl:ation of the React:r
- lant System results in a pressure decrease in the pressurl:er.
React:r trip occurs wnen the pressurizer low pressure tr:p se: point :s reached. A Safety Injection System signal is actuated wnen the appropriate setpoint is reached. These counter:easures w:;l ;;=:: ne consequences of the accident in two ways:
- 1. Reacter trip and borated water injection c==plement veid formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.
- 2. Injection of berated water provides heat transfer from core and prevents excessive clad te=peratures.
At the beginning of the blowdown phase, the entire Reactor Coolant j System centains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling. After the break develops, the time to departure from is calculated, consistent with Appendix K of nucleat9 10CTR50t 5
) boiling
. Thereafter, the core heat transfer is based on l
I local conditions with transition boiling and forced convection to ) as the major heat transfer mechanisms. During the refill steam I period rod-to-rod radiation is the only heat transfer mechanism. I When the Reactor Coolant System pressure falls below 600 psia the ) accumulators begin to inject borated water. The conservative i assumption is made that injected accumulator water bypasses the l core and goes out through the break until the termination of bypass. This conservatism is again consistent with Appendix K of 10CFR50. 14.3.2.1 Thermal Analysis Westinchouse Performance criteria for Emercency Core Ceolina System The reactor is designed to withstand thermal effects caused by a 1 less-of-coolant accident including the double-ended severance of the largest Reactor Coolant System pipe. The reactor core and internals together with the Emergency Core Cooling System are designed so that the reactor can be safely shutdown and the essential heat transfer geometry of the core preserved following the accident. The Emergency Core Cooling System, even when operating during the injection mode with the most severe pingle active failure, is designed to meet the Acceptance Criteria (53, aO
\v 5 e 14.3-6 Ad 1
4 . -6
i BVPS-;-UFOATED TSAR Rev. 10 :; ;;,
;4.3.2.1,1 Meched of Thermal Analysis
- Descriptions of the various aspects of the LOCA analysis are provided in references 6, 8 and 13. These documents describe the ma]cr phenomena odeled, the interfacas among the co=puter codes and features of the codes which serve to maintain compliance witn the acceptance criteria of 10CFR50.46. The analysis of a large-break LCCA transient is divided into three pnases: Blowdown, Refill and Reflood. A series of computer codes has been developed to analyze the transient based on the specific pnenemena which govegg each phase. During the blowdown portion, the SATAN-VI code l is used to calculate the RCS pressure, enthalpy, density = ass and energy flows in the primary system, as well as the heat transfer between the primary and secondary system. At tne end of the blowdown, informationonthy9ptateof the system is transferred to the WREFLOOD code which performs the calculation of the refill period to bottom of the core (BOC) recovery ti=e. Once the vessel has refilled to the bottom of the the reflood portion of the transient begins. 7gJe is used to calculate the thermal-hydraulic The BASH code simulation of the RCS for the reflood phase. Information concerning thecoreboundaryconditionsisgpkenfrom , all of the above codes and input to the LOCBART code I for the purpose of calculating the core fuel rod thermal response for the entire transient. From the boundary conditions, LOCBART computes
- the fluid conditions and heat transfer coefficient for the full
, length of the fuel rod by employing mechanistic models appropriate to the actual fuel and heat transfer regimes. Conservation assumptions ensure that the fuel rods modeled in the calculation represent the hottest rods in the entire core.
*gg)
I containment pressureanalysisisperformedwiththegg{0 with the WREFLOOD code code The
, which is interactive .
transient computed by the COCO code is then input to the BASH code p7gysurefor the purpose of supplying a backpressure at the break plane while computing the reflood transient. The ' containment parausters used in the COCO code to deter =ine the ECCS backpressure are presented in Table 14.3-3. 14.3.2.1.2 Results of Large Break Spectrum Calculations of cold leg double-ended guillotine pipe. breaks are 4 performed over a range of Moody discharge coefficients (C p) to identify the case which produces the highest peak clad
. temperature. For this analysis calculations were performed for discharge coefficients of 0.4, 0.6 and 0.8. This spectrum of breaks was performed assuming the availability of only minimum safety injection flow capacity (Minimum Safeguards), in accordance - with the single failure criteria of 10CFR50, Appendix K. A break discharge coefficient of 0.4 was found to result in the highest peak clad ta=perature.
O 14.3-7 v
. ~ . _ _ . . . _ _ __ _ __
l
-l S'.*PS-1-CPDATED FS AR Rev. 10 (ic32; Consistent with the =ethodology described in reference 31, an additional calculation is p;rformed for the worst break size. n this calculation, ter=ed Maxi =um Safeguards, no failures of the safety injection syste=s are assu=ed to occur. This case was l found to result in the limiting peak clad te=perature of 1918'F, !
wnich is acceptably below the 2200'F limit of 10CFR50.46. The j , results of these calculations are su==arized in Tables 14.3-1 (a & : b). Tacle 14.3-2 contains sc=e key plant parameters input to the I analyses. l
.i Tables 14.3-4 and 14.3-4a present the (MAX SI case) blowdown and :
reflood = ass and energy releases to contain=ent, and the broken loop accu =ulater mass and energy release rate to containment, respectively. Figures 14.3-12 through 14.3-23 show transient plots of i=portant parameters from the code calculations. For each break calculation, transients of the following para =eters are presented. . For the blevdown portion of the transient: I i A. RCS pressure, B. Core inlet and outlet flovrates, , C. Cold leg accumulator delivery rate, D. Core pressure drop, , E. Break mass flovrate,
- F. Break energy discharge rate, and G. Normalized core power. l For the reflood portion of the transient: ,
A. Core and Downconer liquid levels, B. The core inlet fluid velocity, as input to the rod .
. thermal analysis code, ,
C. The accumulator flovrates, and D. Pumped Safety Injection flovrates. j From the fuel rod thermal analysis, at the peak temperature location: ; A. Fluid mass flux, ! B. Rod heat transfer coefficient, E C. The clad peak temperature transient, D. The temperature transient at the burst elevation, and . E. Fluid temperature.
~
The containment pressure transient and the containment vall i condensing heat transfer coefficient are presented in Figures ; 14.3-24 (a & b), for the most limiting break size. The results of several less-of-coolant accident studies are i reported in Reference 37. These results are for conditions which
~
are not limiting in nature and hence are reported on a generic , basis, i 0 14.3-8 F 4 ()<s"g n .
i 3'IPS-1-CPOATIO TSAR Rev. 10 ; 3; , The : lad temperature analysis is based en a total peaking facter of 2.40. The n:t spot .etal water reaction reached is 3.30 l . percent, which is well below the embrittlement limit of 17 per:ent, as required by 13CFR50.46. In addition, the total core t retal-water reaction is less than 0.3 percent for all breaks as
- pared with the 1 percent criterien of ICCFR50.46. l, A general less-of-coolant accident sensitivity study for a 3-icep ,
plant with 17 x 17 fuel is presented in Appendix 14E. These . results are for cenditions which are not limiting in nature and ! hence are reported on a generic basis. ; 14.3.2.1.3 Conclusions - Thermal Analysis For breaks up to and including the double-ended severance of a reacter coolant pipe, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFR50.46. That is: ; i
- 1. The calculated peak fuel element clad temperature provides margin to the require =ent of 2,200*F, based on an F g value of 2.40.
- 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
- 3. The clad temperature transient is terminated at a time ,
when the core geometry is still amenable to cooling. The cladding oxidation limits of 17 percent are not exceeded during or after quenching.
- 4. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the ,
long-lived radioactivit emaining in the core. s e 14.3.2.2 7 tl'nd -
. / Reana vsMfs for Increased Steam Generator l Tui5e P1Fadrnd ' --
The full spectrum analysis presented previously in sections 14.3.2 [' 4p through 14.3.2.1 was for a Steam Generator Tube P1"aninn (SGTP) g 1 level of .10% as reported in Table 14.3-2,_,/1 mit'.g b a i g, f analy d to -a sv le 1o 20% d pre nte n t . .2 The imi gb ak ana is ap fo .ed m d im Sa uar c as f th i .in >
'C = .4 he mana sis res .s e p sen d Ta e' ,
.3-1 hnw enwp anc Wi the 10CF 50. ce anga) pfcri* ia- The 'n the BASH Evaluation !
. t ,odel described;preanalysis in Reference was performed 8, a u eme ed h, supp jo'Io $ riyr visions documented in Reference ad 4ke ekr44 oj ge 4,,, y,
' )
14 '
- hod of Thermal Analysi
. 2 * */* M P
. Th A analysis w performed using the methodology and computer co 4 in References 4, 6, 7, 8, 9, 10, 13 and 59. The k
14.3-9 1
i l l 7-m' L; & SVPS-1-CPDATED FSAE Rev. 13 ,.,32; l f C
/ general description of and interfaces between the various codes is T
/ provide f in Section 14.3.2.;.; and is applicable to tne .
j j reanalys)fs , ,,,.
+ y e >';
The cd5 in=ent parameters utilized in the COCO (Reference 10) M
- ode are provided in Table 14.3-Ja and are applied consistent with _
6; 'hgg {J the methodc. logy described in Appendix A of Reference
. , tt.
Table 14 '. 3 - 2 a contains other key plant para =eters input to :ne reanalystsy F1-M 3 -1C and -1D cepict the SI performance utilized /in th greanaly f14 mi 4 and taximum Safeguards cases respectively. g 10% S C'T P G ASEtT ; N- s' e
^ 14.3.2.2.2 Results of Larged, k Reanalys%s it. 3 2. 2. 2. l 2 0% h f P O'~ + t., / , r '
The previous analysis s res d in a liniting Moody discharge ecoefficient (Cg) of 0.4 w ich was analyzed for both Minimum and Maxi =um Safeguards C D =0.4 cases. Based on the close proxi=ity of the minimum and maximum Safeguards PCT results reported in ! Table 14.3-la, the reanalysis was performed for both C D =0.4 mini =um and maximum Safeguards cases. Tha limiting case in the reanalysis was the mini =um Safeguards case with PCT =2149'F, which continues to demonstrate margin to the 10CFR50.46 limit of 2200*F. The results of both the maximum and minimum Safeguards cases are summarized in Tables 14.3-Ic and 14.3-1d. The non-limiting C D=0.6 and CD =0.8 break sizes were not reanalyzed because the limiting break size is not expected to shift for the following reasons:
- 1. There is significant PCT margin between the limiting and non-limiting C cases in the prior analysis (PCTs of 1911'F; 1759'F,D1667'T respectively).
- 2. The degree and number of input chainges with respect to the previous full spectrum analysist reported in Section 14.3.2 are expected to affect the various break sizes relatively equally.
- 3. The BASH EM is employed by both analynes.
- 4. The Cn=0.4 case was also limitinct in the prior FSAR analysis using the 1978 H EM (WCAP-9220-P-Rev 0).
Tables 14.3-4b and 14.3-4c present the minimum SI case blowdown and reflood mass and energy releases to containment, and the broken loop accumulator and SI mass and energy release rate to containment, respectively. Figures 14.3-12H to N, 14. 3-13 E to H, and 14.3-23F to J document the blowdown, reflood and clad performance transient for the maximum safeguards case respectively. Figure 14.3-24C documer ts the containment backpressure transient for the limiting minimum Safeguards case.
, ne l'f. 3, 2. 2. 2 . 2 30 7. AT P ktw<'7 > 5 14.3-10 b LWWF
- n f
ff SVPS-1-iJP::ATED F Rev. 70 ( 1/ M ,
/ ,*,-L 4 / 3r. t g-T P ;
( ~ &n ; 14.3.2.2.3 Ocnclusion - Thermal Analysis-of La- _reak ReanalysKs, ~
/ ,
- For breaks up to and including the doddle-ended severance of a reactor coolar.: pipe, the ECOS will meet the Acceptance Criteria as presented in 10CFR50.46. That is: ,
. 1. The calculated =axi=um peak fuel element clad temperature provides margin to the requirement of 2200*F.
l
- 2. The total amount of hydrogen generated fro = the che=ical !
reaction of the cladding with steam and water did not j exceed .01 times the hypothetical a=ount generated if all ' the metal in the cylindrical cladding were to react.
- 3. The clad temperature transient is terminated at a time when the core gec=etry is still amenable to cooling. The oxidation limit of 17% of the cladding thickness is no exceeded during or after quenching.
- 4. The core te=perature is reduced and decay heat is removed I for an extended period of ti=e, as required by the long-lived radioactivity re=aining in the core.
f, , 4' 14 . .' . 3 Core and Internals Intecrity Analysis I 14.3.3.1 Internals Evaluation
.n~\
.g3f Che forces exerted on the reactor internals and the core following P m LOCA are computed by employing the BLODWN-2 digital computer y gphkU\/ program developed for space-time dependent analysis of multilocp PWR plants.
N k W 14.3.3.2 Design Criteria
,(pg, Following a LOCA, the basic requirement is that the station shall be shut down in an orderly manner and cooled down so that fuel cladding temperature is kept within specified limits. This implies that the defor=ation of the reactor internals must be kept sufficiently small so that the core geometry remains substantially intact to allow core cooling and insertion of a sufficient nu=ber of control rods.
After the break, the reduction in water density greatly reduces the reactivity of the core, thus making the core suberitical, causing reactor shutdown independent of the control rods. In other words, the core is subcritical whether or not the rods are tripped. (The subsequent refilling of the core by the ECCS uses borated water. to maintain the core in a Therefore, suberitical state.) insertion of most of the control rods further ensures the ability to shut the reactor down and keep it in a safe shutdown condition.
~
Maximum allowable deflection limitations are established for those regions of the internals that are critical for reactor shutdown. 14.3-11
.-M.epws ,,4 l
BVPS-1-UPDATED FSAR Rev. 10 ,1, 3 : ; l i Peferences to Secticn 14.3 . i e = ; S.D., Tauche, W.D., Schwarz, W.R., t Lee, H., Rupprecht, 1.
" Westinghouse Snall Break ECOS Evaluation Model Using the - NOTRUMP Code," WCAP-10054-P-A, August 1985.
P.E., "NOTRUMP, A Nodal Transient Small Break and ;
- 2. Meyer, !
General Network Code," WCAP-10079-P-A, August 1985. .:
- 3. J. M. Hellman, " Fuel Densification Experi= ental Results and l Application", WCAP-8219, Westinghouse Model for Reactor Electric Corporation (October 1973). ;
et al., '"LOCTA-IV Program: Loss-of-Coolan*
- 4. F. M. Bordelen, '
inghouse Electr:.c Transient Analysis", -
-~~ 3 Corporation (June 1974). w c49-93 a t -P wch Pf Jor-Np
- for :
- 5. " Acceptance Criteria for __e a -
10CTR50.46 and i Light Water Cooled Nuclear Power Reactors" Federal Register, Volume 39, Number 3, Appendix K of 10CFR50. Westinghouse Electric Corporation (January 4, 1974). l
- 6. F. M. Bordelon, H. W. Massie, and T. A. Zordan, "Westinghousa ECCS Evaluation Model Summary" WCAP-8339 Westinghouse Electric l Corporation (July 1974). ,
Program: Comprehensive F. M. Bordelon, et al., " SATAN-VI 7. Space-Time Dependent Analysis of Loss-of-Coolant", WCAP-8302 !
- n-Pro version),
(Proprietary Version) Westinghous ctri rporation (June _1972)prietary -
. j j gp .:. 2
" 19 '
o t
.. ho
. W -1026 24 t
- CCS t , de ng B , _
culat Model f a flooding i Ty;ee4 9. R. D. Kelly, et al., ,W -8170 4 after a Loss-of-Coolant Accident (WREFLO
- Code)
WCAP-8171 (Non-P ristary ? sion), ! b (Proprietary Version),
\ j Westinghouse Electric Corporation (June 19 ).
G
- ontainment . essure l
- 10. F. M. Bordelon, and E. . hy, WCAP- 7 (P stary ersion), i Analysis Code (COCO WCAP-8326 (Non-Prop a ry Version , Westi e Electric l 1
Corporation (June 197 1 I (USNRC),
- 11. Johnson, W. J. (Was in house) letter to Murley, T. E.
NS-NRC-89-3463, Octobe 5, 1989.
^
I i
- 12. Deleted by Revision O. l
~
- 13. F. M. Bordelen, et al., " Westinghouse ECCS Evaluation Information", WCAP-8471 (Proprietary Model-Supplementary Version), WCAP-8472 (Non-Proprietary. Version), Westinghouse Electric Corporation (January 1975).
- 14.3-51 1
1
3'<*PS-1-UPDATED FSAR Rev. ;3 .;,3:= l Feferences to Sectien 14.3- Icent'd) I l
1
- 14. S. Fabric, Cc puter Program WHAM for Calculation of Pressure, l Velocity and Force Transient in a Liquid-Filled Piping !
Network", Report No. 67-49-R, Kaiser Engineers (November
. 1967).
- 15. "LOCT*C - A Cc=puter Code to Determine the- Pressure and Temperature Response of Dry Containment to a Loss-of-Coolant i Accident", Stone and Webster Engineering Corporation, !
Proprietary Report, (September 1971).
- 16. W. H. McAdams, Heat Transmission. Third Edition, 1954.
- 17. "LOCTVS - A Computer Code to Determine the Pressure and Temperature Response of Pressure Suppression Containments to a Loss-of-Coolant Accident", SWND-2, Supplement No. 1, Stone &
Webster Engineering Corporation, March 1973. r
- 18. A. H. Shapiro, "The Dynamics and Therredynaries of -
coveressible Fluid Flow", Vol. 1, The Ronald Press Company, l New York (1953). -
- 19. G. B. Wallis, One-Dimensional Two-Phase Flew, McGraw Hill Book 1 Co., New York (1969).
{;
- 20. Deleted by Revision 2
- 21. Deleted by Revision 2 i
- 22. Deleted by Revision 2 ;
, 23. Deleted by Revision 2 ;
r
- 24. Deleted by Revision 2 j :
- 25. Deleted by Revision 2 t
- 26. Deleted by Revision 2 l
l
- 27. Deleted by Revision 2 !
- 28. Deleted by Revision 2 ,
- 29. Deleted by Revision 2
~
- 30. Deleted by Revision 0.
7
- 31. Rah . P. (Westinghouse) letter to Tedesco, R. L. (USNRC), l NS 2538, December 1981.
~
6fk * !
- 32. Deleted by Revision 0.
l
- 33. Deleted by Revision O.
' Y 14.3-52 O t
-, ,_ __ - - - - - _ ~
.~ . .. . . . .. - . .
-, . .~ . - -.
f n' BVPS-1-CPDATED FSAR -Rev.L;f r;,3:; ;
'l i
Peferences to Section 14.3 (Cent'd) j
- 49. Deleted by Revision 2 !
- 50. Deleted by Revision 0, ,
l
- 51. Deleted by Revision O. j
- 52. Deleted by Revision O. ;
from C. N. Dunn (Duquesne Light Company) to R. W.'Reid I
- 53. Letter (Nuclear Regulatory Commission),
Subject:
proposed permanent' ! modifications to correct NPSH modifications. (November 17, l 1977). ;
- 54. Letter from C. . N. Dunn (Duquesne Light Company) to'.A. !
Schwencer (Nuclear Regulatory Commission),
Subject:
completed ! request for additional information (August 3, 1978). ;
- 55. Deleted by Revision 0. ;
- 56. Deleted by Revision 0.
~ . 57. T. Kreith,.Princieles of Meat Transfer, International Textbook.
Company (1966).
- 58. Deleted by Revision.7.
.f I
- 59. " Westinghouse ECCS Evaluation WCAP-9220-P-A, Revision 1, Februa RModel- -
1981- Version", j 1982. i L- , j
- 60. Deleted by Revision 7.
N 4 e4-P-- 9221-N P.- A j-
- 61. Deleted by Revision 7. ,
- 62. Deleted by Revision-7.
- 63. Deleted by Revision 7.
- 64. Deleted by Revision 7.
. .I
- 65. Deleted by Revision 7. f
- 66. Safety Evaluation in the Matter of Virginia Electric Power Company,. Surry Power Stations- Units-1 and 2, Docket Numbers; j
.50-280 _and- 50-281, pp. ~
57-58, ' Atomic Energy. Commission
-(February- 23,- 1972). j.
- 67. NUREG-0772, " Technical Basee' for Estimating. Fission Product .;
-i Behavior'During LWR Accidents",-Appendix ~E, June 1981.
- 68. . DLC Calculation ERS-SFL-83-016- d
- 69. DLC Calculation'ERS-5TL-83-017- , L j 14.3-54 b
l 37PS-1-L*PCATED TSAR Rev. 13 ';;;;; , Peferences to Section 14.3 reent'd1 '
-^
_s_
/ c M%cc, > cf MG tc em. m l
- 70. OLC Calculaticn ERS-STL-84-008 g [eg
' m' w ,' 3H,3,1 Mb UEt d g ,,
- j,g 3,
- 71. CLC Letter NDISLC:1037 -
'O'd --
#5.
7 esign Change Package 372, Removal of Safety Injection T Signal from Centrifugal Charging Pumps L, Valves e Miniflow Isola tion / !
- 73. Cortined BV-1 -
Bases Accidents BV-2 Control Room Habitability Due to Design (except LOCA) at BV-1, Calculation Identification Number 12241/14410.39, UR(B) -4 56 dated 4-2 4 -8 7. ;
- 74. Doses to the BV-1 Control Room Due to a LOCA at BV-1, i Calculation Identification Number- 14110.39, 5-11-87. UR(B)-457 dated
- 75. "American National Standard for Decay Heat !
Water Reactors", ANSI /ANS - Power in Light 5.1 - 1979, American Nuclear Society, August 1979.
- 76. D. C. Slaughterback, for condensing Steam in "A Review'of Heat Transfer coefficients a Containment Building Tollowing a .(
Loss of Cooland Accident", . Interim Task Report, ; 4.2.2.1, Idaho Nuclear Corporation, January 1970. Subtask ;
- 77. "PWR TLECHT Tinal Report", WCAP - 7665, Westinghouse Electric !
Corporation, April 1971. Yr '
' efere es do not ar in th text .e are a p .
e BVPS-1 C5 Re ys , Westingh Letter D -79 3 date April 25, 79, an are pr purpose d for in ation i P w Lo v.e r < G.'
~
't i
14.3-55 h
i
,i l '
INSERT A: ;f =.
'Panial break spectrum reanalyses were performed for 20% SGTP and then for 30% 'l SGTP. These reanalyses are presented in section 14.3.2.2. For the 20% SGTP reanalysis.
the limiting Co=0.4 break was reanalyzed for the minimum and maximum safeguards l cases. The 30% SGTP analysis analyzed the Co=0.4 and Co=0.6 minimum safeguards l cases. Both the 20% and 30% SGTP reanalyses results are presented in Table 14.3-Ic l and show compliance with the 10 CFR 50.46 acceptance criteria.' j INSERT B: ,
'The 30% SGTP reanalysis was performed using the BASH EM described in reference 8, as supplemented by various EM changes listed in section 14.3.2.2.1.
INSERT C: i
'The 30% SGTP analysis was performed using the methodology and computer codes ;
described in References 4,6,7,8,9,10,13,59,78 (1989 W EM Update), item 2.1 (Fuel i Rod Model) of Reference 80 (1991 W EM. Update), Reference 84 (1992.W_ EM Update) Item ' Spacer Grid Heat Transfer Error in BART' (Reference 83), the NUREG-0630 Burst / Blockage Correlation Extension (Reference 85), various discretionary changes to input / output format and inclusion of code diagnostics that do not affect results, and several negligible LOCBART Error corrections associated with pellet / clad contact and ~ clad thinning models. In addition, the Reference 80 (1991 W EM. Update) Item 3.2 Burst / Blockage application change was incorporated into the analysis as discussed.in detail below. The Reference 80 Burst / Blockage issue must be reviewed on a case-by-case basis since a coding change was not incorporated into the n.odel. The IFBA runs reponed herein fall into the category where the results would be affected by the issue using the algorithm developed in Reference 80. However, detailed sensitivity study work and investigation into the transient results concluded that the issue is inherently not applicable to. the limiting IFBA transient due to the different mechanisms involved in determining the PCT. The remaining non-limiting (i.e. non-IFBA) analysis results reponed herein reflect
-consideration of this issue. The non-limiting transient results fall into the Reference 80 -
'not affected' category.
- a.
INSERT D;
'as supplemented by the use of nominal containment temperature for the 30% SGTP analysis only. Nominal containment temperature use is consistent with recent sensitivity studies which have indicated a very weak sensitivity to this parameter with the direction of sensitivity being analysis dependent. In addition, the 30% SGTP analysis utilized a more conservative value (of 8.9 psia) for containment initial pressure than would necessanly be required per the Reference 6 model.'
INSERT E:
'The SI performance in the 30% SGTP reanalysis was supplied by DLCo Reference 82, from which Table V.B.3 (I HHSI pump + 1 LHS1 pump) was utilized.'
1 INSERT F:
'The Co=0.4 Minimum Safeguards limiting case from the 20% SGTP analysis was I reanalyzed at the 30% SGTP configuration. Reviewing the transient thermal hydraulic results of this case (in comparison to the Maximum and Minimum Safeguards cases of the original full spectrum BASH analysis of Section 14.3.2 and the 20% SGTP BASH reanalysis of 14.3.2.2.2.1), it was determined that the Maximum Si case for the 30%
SGTP plant configuration would be even more non-limiting than previously observed in the 20% SGTP analysis. Therefore, by virtue of this engineering evaluation, the Maximum SI case wa: not analyzed. Several LOCBART clad heatup calculations were performed to account for the variety of fuel types under consideration for Cy 10 and future operation. LOCBART runs were made for 17x17 Vantage 5H (V5H) w/o IFMs
- (IFBA & non-IFBA), and Standard (STD) (non-IFBA). The limiting result was for V5H w/o IFMs IFBA fuel, with PCT =2144*F. The fuel cladding transient results are presented l in Table 14.3-Ic and the sequence of events is presented in Table 14.3-1d for all the fuel types.
The degree and number of input changes with respect to the previous full spectrum analysis reponed in section 14.3.2 are not expected to cause a shift in the limiting break coefficient. However, to quantify the amount of PCT margin to the next most limiting break coefficient, a Co=0.6 Minimum Safeguards analysis was performed for the V5H w/o IFMs non-IFBA fuel. This analysis showed 147'F margin to its C =0.4 o counterpart. These 30% SGTP PCT results supersede those presented in 14.3.2.2.2.1 for the 20% SGTP analysis, which is retained for historical perspective only. All of the 30% SGTP reanalysis results continue to demonstrate margin to the 10 CFR 50.46 limit of 2200'F. Tables 14.3 4b and 14.3-4c present the minimum SI Co=0.4 case blowdown and reflood mass and energy releases to containment, and the broken loop accumulator and Si mass and energy release rate to containment, respectively. Figures 14.3-120 to -12U and 14.3-
- 131 to -13L document the blowdown and reflood hydraulic transient respectively for the Cd=0.4 MIN Si case.
Figures 14.3-14K to -140 document the clad performance transient for the limiting IFBA fuel type PCT transient. Figures 14.3-14P and -14Q present the Clad Temperature transient for the non-limiting V5H non-IFBA and STD non-IFBA fuel types. '
INSERT G:
- 78. NS-NRC-90-3524, ' Additional Information on use of BASH Code,' Johnson (3') to Jones (NRC RSB NRR),7/20/90.
79 not used.
- 80. DLW-91-159, 'DLCo BVPS Units 1&2 ECCS Evaluation Model Changes," Steinmetz (%')
to Noonan, (DLCo), 6/20/91.
- 81. ND1MNE:6312, Halliday (DLCo) to Simmers (W), DLCo BV Unit 1 increased SGTP and Reduced TDF Program, 10/12/92. (LOCA Parameters).
- 82. ND1MNE:6340. Halliday (DLCo) to Simmers (_%), DLCo BV Unit 1 Increased SGTP and Reduced TDF Program, 11/12/92. (ECCS Performance).
- 83. WCAP-10484 P Addendum 1; ' Spacer Grid Heat Transfer Effects During ReDood.' D. J.
Shimeck, 12/92. (Transmittal to NRC DCD: ET-NRC-92-3787,12/22/92, ' Noti 6 cation of Changes to the W Large Break LOCA ECCS Evaluation Model'.)
- 84. DLW-93-202, 'DLCo BV Units 1&210 CFR 50.46 Notification and Reporting Information,'
Hall (.%9 to Tonet (DLCo),1/29/93.
- 85. ET-NRC-92-3746, ' Extension of NUREG-0630 Fuel Rod Burst Strain and Assembly Blockage Models to High Fuel Rod Burst Temperatures,' Liparuto (3.) to NRC Document Control Desk,9/16/92.
Insert H
- 8. WCAP-10266-P-A Rev. 2, WCAP-11524-NP-A Rev 2, 'The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code', Kabadi, J. N., et al., March 1987; including Addendum 1-A ' Power Shape Sensitivity Studies' 12/87 and Addendum 2-A
' BASH Methdology Improvements and Reliability Enhancements
- 5/88.
es
i l l Note: there is no Insert I. t t i Insen J:
*A summation of the plant parameters used in the SBLOCA analysis are provided in Table 14.3-?b.
The 30Fo SGTP analysis was performed using the methodology and computer codes desenbed in References 1,2,4,11, and 80. The methodology updates in reference 84, i which were used in the 30?o SGTP analysis include: Auxiliary Feedwater Flow Table f S.G. Secondary Side Modeling Enhancements Basis Change for Hot Assembly Rod Gap Pressure Modifications were also made to the actuation of Auxiliary Feedwater (AFW). The l NOTRUMP EMm assumes that the AFW actuates on reactor trip coincident with loss of offsite power. The full break spectrum analysis conservatively actuates AFW on the Low- ! Pressurizer Pressure SI signal.' : l Insert K: : l
'A full spectrum of breaks was analyzed at the beginning oflife fuel rod conditions.- This i spectrum included break sizes of 1.5 inches,2 inches,3 inches, and 4 inches in diameter. l The limiting break size was shown to be the 3 inch diameter break, but by only 2*F over t
the 2 inch diameter break. A summary of the results can be found in Table 14.3-le and Table 14.3-If.' ; e e f
i i SVPS-1-UPDATED FSAR Rev. 10 (1/92) TABLE 14.3-1c
- 1.ARGE BREAK LOCA FUEL CLADDING DATA 20% STEAM GENERATOR TUBE PLUGGING ,
Break 0.4 Mini =um SI 0.4 Maximu= SI Parareter 1 Peak Clad Temperature (F) 2149 2089 ' Elevation (ft) 8.75 8.75 Maximum Zr/H2O Reaction (%) 8.75% 7% Elevation (it) 8.75 8.75 Total 2./H2O Reaction (%) <1% <1% Hot Rod Burst Time (s) 39.3 39.3 Elevation (ft) 6.00 6.00 , [.'
, LARGE BREAK LOCA 'A 20 FUEL CLADDING DATA 'e ,
s pt STEAM GENERATOR TUBE PLUGGING e
~
O.4 %.mSr 9,g m.m, s y_ Parareter W N0"-5rsA {gy ,,3pgg Peak Clad Temperature (F) Wy,gg707T W 1901 Elevation (ft) pp Maximum Zr/H2O Reaction (%) Elevation (ft) W S', T W 37 : 7,2 f . /.J f-Total Zr/H2O Reaction (%) T lYo _ < l 0/, Hot Rod Burst Time (s) N SO Elevation (ft) 6.0 f.f j
'f ' , } -Break 0*Y NIlk1*** S S Q , y % ;,. m .i t b 4 V5 y SF8A gy Parareter Peak Clad Temperature (F) M 7, 5' Elevation (!t) 32[
Maximum Zr/H2O Reaction (%) M # 2' 7 78 Elevation (ft) 7,'g f {
, Total Zr/H2O Reaction (%) (jp 4/ fo i
Hot Red Burst Time (s) TI 36' 3 Elevation ~(ft) [, o (, fy
l
)
3'.*?S- 1-CFOATED F5A?.' .
?.e v , ;0 ,1,3 : I
-I rr-str u.,--
( LA.GE 9 PEAK OCA ! TIME SEOUENCE OF EVENTS i 20% STEAM OENEPATOR TTEE PEUOG!NG ! i
.o ;
3reak Less C: efficient: 0.4 Minimum SI 0.4 Maximum S: i Event ! Break Opens 0 0
- Reactor Trip .38 .38 l i
Safety Injection Signal 2.01 2.01' j I Intact Accumulator l Injection Begins 13.9 13.9 j Pumped SI Begins 29.01 '29.01 ! I End of Bypass 31.87 31.87 {
.. End of Blowdown '31.87 =31,87 ~
Bottom of Core Recovery 45.7 45.1 l Accumulators Empty 54.1 54.1 i PCT Occurs 239 237 '; i 30% STEAM GENEPATOR TUBE PEUGGING l 0 l' (p m .~, a . l Break Loss Coefficient: 0.4 Minimum SI 0.g Mmmeusm. SI- y i l Event i ' d Break opens 0 0 ; Reactor Trip ,yo ,fD Safety Injection Signal 2.I8 - l* Y i Intact Accumulator g *7 - 10. 4 I Injection Begins , Pumped SI Begins 24*Ig 1R b I End of' Bypass I* ' End of Blowdown 3/ h . 2 8. YI
. Bottom of Core Recovery *50 3007 Accumulators Empty $ 3.[pf , 48, PCT Occurs ,21Y- -. Wh', 34) I%: N" j l
,n1 / AV u.s-IF64) _
Table 14.3-le i SMALL BREAK LOCA TIME SEQUENCE OF EVENTS '
- 30% STEAM GENERATOR TUBE PLUGGING Break Size 1.5 " 2" 3" 4" i
Break Opens 0 0 0 0 1 Reactor Trip Signal 55.3 28.8 12.1 7.1 Safety Injection Signal 65.6 38.4 20.5 15.0 Accumulator Injection N/A N/A i168 615 PCT Occurs 10997 3485 1271 701 Transient Termination 17000 8000 5000 4000 i a i e
i.
' Table 14.3-If SMALL BREAK LOCA FUEL CLADDING DATA
304 STEAM GENERATOR TLTE PLUGGING 1.5 " 2" 3" 4" Break Size 782 1748 1750 1275 Peak Clad Temperature (*F)
I1.0 12.0 11.75 11.0 Elevation (ft)
.03 4.9 1.8 .09 Maximum Zr/H2O Reaction (4) 11.25 12.0 11.75 11.0 Elevation (ft)
Total Zr/H2O Reaction (%) < l .0 < l .0 < l.0 < l .0 N/A N/A N/A N/A Hot Rod Burst Time (sec) Elevation (ft) N/A N/A N/A N/A l 1 F
~
1 u i l
-l l
l 3
SVPS-1-UPDATED FSAR Rev. 10 (1/92) TABLE 14.3-2a 2 PLANT PAFAMETEPS USED IN L\RGE BREAK LocA 2 0% e STEAM GENEPATOR TUBE PLUGGING l App 3c#o / [ Core Power 102% of 2652 Mvt
. ** 1 core Peaking Fac* r _
2. of 12.49 kw/ft Baffle-Barre 1 Configuration Upflow Steam Generator Tube Plugging ,QOt,30 % (Uniform: Up to 20%,30% /
/ in any or all) /
RCS Initial Conditions zo% 5pyr 3o'LSoTP
=. ntW Vessel Flowrate 28CSI l lbm/sec h I W J,<.
Core Flowrate 2gc3/ 'lbm/s W p4n, Vessel Inlet Temperatur 54 .7*F raf f. (, *p _ Vessel Outlet Temperature 609.5'T 2 997 (;, g,iop j
. Pressure 2280 psia p 2.1 80 PDA Cold Leg Accumulator Conditions &
. %, ,, s Cover Pressure ater Volume / Accumulator 600 psig 1035 ft (Note 1) / W"oM i
'Is M Water Temperature 90*F 9); 'f 3' (
Fuel Analyzed 17xl7__S M ARD'(N (2 0Jh
#/
w Nde 5 _(3E* 5WP)
- Notes:
- 1. 1035 ft 3/ loop accumulator water nominal volume (7740 gal); is measured from check valve closest to A the RCS and does not include the undeliverable volume below.the elevation of the outlet piping.
- 2. Analysis of 17x17 STANDARD bounds 17x17 VANTAGE SM withou 99 R&< 0
y, a g a,,,
], y Q sl] STAWbM9 ("w- Z F6A ' h l]*II YMVS'E
.. $y 4. d I vP As ( I-F64 w2 **uJ F 84) A f-
, L OC h rl f h.stembly Pes kivy YucA* p , 3 ct 5"1 $
- 10%5&R 1 of 1 1.42. Vcud 2%5MT sH l'f Oc' $ 0'0 W
i Table 14.3-2b
~
PLANT PARAMETERS USED IN SMALL BREAK LOCA 30% STEAM GENERATOR TUBE PLUGGING 2 Core Power 102% of 2652 MWt Total Core Peaking Factor (FQ) 2.40 , t Hot Channel Factor (FDH) 1.62 LOCA Hot Assembly Peaking Factor 1.46 (P-bar-HA) i Peak Linear Power 102% of 12.5 kw/ft Baffle-Barrel Configuration Upflow 4 Steam Generator Tube Plugging 30% (Up to 30% in any or all) Minimum Auxiliary Feedwater Flow Rate 300 gpm/SG
. RCS Initial Conditions 7 Vessel Flowrate 26114 lbm/sec (248,520 gpm)
Vessel Inlet Temp. 546*F Vessel Outlet Temp. 618'F-Max. RCS Pressure 2280 psia Cold leg Accumulator Minimum Cover Pressure 600 psia Water Volume / Accumulator 978 ft' (1) Max. Water Temperature 105'F Fuel Analyzed W 17x17 STD/ Vantage SH w/o IFMs
. (1) The accumulator is modelled as 978 ft) nominal water volume (measured to tank outlet neglecting unusable water volume) without crediting the average 57 ft'/ loop volume between the -
tank outlet and the first RCS check valve.
3VPS-1-UFOATED FSAR Rev. 10 (1/92) - TABLE 14.2-?a , CON *AINMENT.3ATA USED IN '_APGE BREAK LOCA (VWr e.
~
.i 2 0YiSTEAMENERATOR TUBE PLUGGING n{ cp A L, yj p j
/
y 'n .' 4 NET FPEE VOLUME 90 3,9I 1,890,000 ft YW l Ve m -l INITIAL CONDITIONS . s Pressure 9.5 s i a ( MN#) ' W'" *
) ,
Te=perature 90' (le%)'* Qs# h*(30'fe)
- Humidity 99%
RWST Temperature 45'F t Outside Temperature -20*F : g ypy,y gyg3y $ 9 .S Y W o n v a l ( A cruce8/) r Number of Pumps Operating 2 Runout Flow Rate (each) 2200 gpm per pump . Actuation Time 55 seconds l, TAN COOLERS NONE . STRUC"VRAL HEAT SINKS (1) I t
. 1 Wall Thickness Paint surface' Area !
Number Material (ft) M (sc'fti , i
. 1 Concrets' O.5 4 6,972 1 2 Concrete 1.0 4 77,446 {
t 3 Concrete 1.5 4 36,848 i I 4 Concrete 2.0 4 17,010 ' 5 Concrete 3.0 4 8,632 , I 6 Carbon Steel 0.03125 5 18,635 l Concrete 4.5 i 7 Carbon Steel- 0.03125 5 33,094 - Concrete 4.5 l 8 Carbon Steel 0.04167 5 26,775- 1 i Concrete 2.5 j 9 Concrete 2.0 4 13,388 Carbon Steel 0.03125 Concrete 10.0 1 of 2 i l l 4
/
gi#l> SVPS-1-CFDATED FSAR Rev. 12 (;/92) M, h TABLE 14.3-4b
' *%'DObH AND REFf OOD MASS AND ENEPGY M LEASE P.ATES TO CONTAINMENT LARGE PPEAK 'CCA 20%) STEAM GENEPATOR TUBE PLUGGING (Oces not include Br: Ken Lecp ECCS spill (Table 14.3-4c)
=
.W^m jCD = 0.4 minimum SI %
30% 56 7 f kcc.veIV T;d l u, m -
.l l
Time (sec) Mass Flevrate (lbs/sec) Energy Flcwrate -l Million BTU /sec
.2 51750 27.56 2 44146 24.01 5 24864 14.11 10 13514 8.87 15 6933 5.13 !
20 3686 3.02 l 25 3944 1.60 30 2965 .742 38 0. O.
, 46 4.36 .0057 49 23.5 .0304 59 176 .142 69 226 .152
, 89 250 .152 109 259 .148 159 294 .141 209 333 .144 279 345 .138 0:9 MWm G Jo% STP Ree-4N
%e Dec) hau Phwr-k lik/s.c) 8seny'f"k"h"0'"h 29.3 f
2r SzC 98 g nsw zy t3e 23.9.9 g 13. 9 7 gy 17930 g 57 G SS1 1,87
- vs' 2.a 3a ZW)b 0.487 0 0 48 2A' . 029 T TT so? , / /1.
Ii 17f 1 og 1 ,11T G1 pi .13 4, 87 3 >o ./O7-
i m F.e v . 10 ,1. 3 2 ) 3*.*FS-1-CFOATID FSAR
*/' Vgo , 1 T131E 14. -4c ~
BPOFIN iCOP ACCL*MT*LATOR AND SI MASS AND ENERGY PEiEASE RATES DUFING bio *4DC'a*N '
- _AFGE BREAM LOCA 2 0 V7STEA GENEP.ATOR T'_*B E PLUGGING e
C = 0.4 minimus SI 56f7 &c'~o (20 'f-
- n. - a ,m- - - -
Ti=e (sec) Mass F1 curate (lbm/sec) Energy Flowrate BTU /sec 0 4239 252731 2 3516 209633 4 3074 183254 7 2641 157477 10 2344 139752 14 2061 122911 18 1855 110612 22 1698 101221 26 1576 93984 28 1525 90943
-- 29 0 0 31 168 2184 rfy
[ ,T o.1 p, . n S2 WTP R w a% Sis j Jo*/a l NEU *A Smf(< e) ane(Scc) . [ O 7 4240 3517 asyn 230al03 I % S3 'l l y so ?S* 2 (, Y3 165 '/ 7f' -
,f -
23yg 146876 l fy .2 0 6 f /21267 i g (, o ll6 Y2 9 - 1 18 -
/7o 3 /oCG .5 3 _
- ' T h3 1 of 1 U/ 08 ;
2& is3A W 9' 2 V 13 o O n.
REV. i I l 2400. ; 2208. i 2888. .
?
1888. , f= 1688. , t
- a ,
x 1488. j t C :
- s
$$ 1288.
0
% \ i 1888. -!
l 988. 3 [ 688.
\ 1
% n s-w i 488. y ~gg j
.t 2M ft
- 1. 500. ISOS. 1580. 2800. 2580. 5808. 5588. 4888. 4588. 5880.
TIME (SEC) .) l l FIGURE 14.3-2 l i RCS DEPRESSURIZATION TRANSIENT (3 Inch) BEAVER VALLEY POWER STATION UNIT NO. i i UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging ;
REV ! t
- o 54 ,
t , 52.I, L l 30.
. i
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~ #
,,*j sj l v i 3 t
- 10. \ /.g.
l' 14 ; i 12 B. 500. 1000. 1500. 2000. 2500. 5000. 5500. 4000. 4500, 5000. i TIME (SEC) F l FIGURE 14.3 CORE MIXTURE HEIGHT (? Inch) BEAVER VALLEY POWER STATION UhTT NO.1 ! UPDATED FINAL SAFETY ANALYSIS REPORT ' 30% Steam Generator Tube Plugging
~
c- .
REV d
' 1
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! : l i i i 5
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- 6:C I i 1400. .
I <
.e 5
'200 A
h
?
s-W 1000. r-A 1 0 800. l l 600. i 400. 2500. 3000. O. 500. 1000. 1500. 2000. . l TIME (SEC) j 1
% l 1
FIGURE 14.3-4 -l CLADDING TEMPERATURE TRANSIENT (3 Inch i l BEAVER VALLEY POWER STATION UNIT NO. 1 UPDATED FINAL SAFETY ANALYSIS REPORT . l 30% Steam Generator Tube Plugging
REV 1 1 1 l d . 2220. ; ; i . i l 1, >
, l ,
! l 4 2220. ,
,- t W
e 1800. ; e S i l I w 1520. , i' 9 1400. .! e j . E i
< 1200.
m l 1 '
~
O l Q 1000. , x w Z - W 900. D
*E I 600. ?
H .l m b 400. E_ 200. ,
^
0 500. 1000. 1500. 2000. 2500. 5000. 5500. 4000, 4500. 5000.
- 3. ;
TIME (SEC) 4 om f i i FIGURE 14.3-5 l STEAM FLOW (3 Inch) l BEAVER VALLEY POWER STATION UNIT NO. I ! UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging )
REV ! 1 i l
- 3 .- ;
- i ;
+ t r4 ,
i l
+ ! ! l l x
E ' l l l l ; P
~
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< 93
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i I h l E I : 6 i i_ i i i l l , u 1- t g { w I l g w o 102 1 U
- i i l C i i i ,
E i A '
< l L ./ N I :
x ~ f-m
- 10 1
- 0. 500- 1000 t500. 2000. 2500. 3000. ;
TIME (SEC) ! FIGURE 14.3-6 i ROD HEAT TRANSFER COEFFICIENT (3 Inch) BEAVER VALLEY POWER STATION UNIT NO. I UPDATED FINAL SAFETY ANALYSIS REPORT I 30% Steam Generator Tube Plugging j i
i REV
?
.. 5:C , , I l 'l i , s:c. . i I i i 14 0. gi e 5 s A H 12:o. !
< I l x *e > = .c W
c-1000. , e ' 9 2 300. . r i 600' l.d i 400. 3000; O. 500. 1000. 1500, 2000. 2500. - 5 i TDdE (SEC) - { f s FIGURE 14.3-7 HOT SPOT FLUID TEMPERATURE (3 inch)- BEAVER VALLEY POWER STATION UNIT NO. l , UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging . l
- r. L
?- I REV !
?
2400. : i 4 i i 2200. ! 2000. 2 E a.
~
1908. m x o w
, s 1698.
O x 1498. 1-1298. N m s-1898 y,. 9888. 12000. 16888. 4333 - TIME (SEC) FIGURE 14.3-9A ~ RCS DEPRESSURIZATION TRANSIENT (l.5 Inch) BEAVER VALLEY POWER STATION UNIT NO. I UPDATED FINAL SAFETY ANALYSIS REPORT' 30% Steam Generator Tube Plugging
REV i 2402. 2200. 2280. ; _ 1000. w 1600. x a
- n h 1488.
c.
'A E i 12 8. '
5 5 s ; 1.. . r
- 9. 3 s= . . 1. 2. s..s..7. . ..
l TIME (SEC) L I FIGURE 14.3-9B l RCS DEPRESSURIZATION TRANSIENT (2 Inch) BEAVER VALLEY POWER STATION UNTT NO. 1 UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging
7-l i REV 1 i
- 2488.
2208. ; 2888. 1888. 8 ,
$ 1688.
?
t W 1488. x
=
. $w 1288. i
% i v>
. M 1888.
t 688. , t 688. 488. L ' V 't r V N ! 200
- 3. 580. 1000. 1500. 2000. 2500. 5808. 5588. 4880. i TIME (SEC) i i
i FIGURE 143-9C l RCS DEPRESSURIZATION TRANSIENT (4 Inch) ; BEAVER VALLEY POWER STATION UNTT NO. I UPDATED FINAL SAFETY ANALYSIS REPORT , 30% Steam Generator Tube Plugging -
3 i l RE\ ; i . Ea. . i
! i !
{ t l i
,j q I
ij ! , 1
-i ;
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.~
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; h. ' ' = ac . ! > 49.
wm I A t Js ' w '< \ 1 x .., 4 PE FC [g j
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, w C _
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t 1 - >
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/ )
1 , 1 20 16000.
- 2. 4000. 9000. 12000. j i
]
FIGURE 14.3-10A CORE MIXTURE HEIGHT (1.5 Inch) i l
. BEAVER VALLEY P.OWER S. TATION UNIT NO. 1 UPDATED FINAL SAFETY ANALYSIS REPORT -
30% Steam Generator Tube Plugging ! i
REV. l
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- i i t
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t i
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a f- 26.: -
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2m i i I 1 f Dx i i- O
- XG 24, 46 w.
x-Ph 22. I l V~ .. t 20- 7 s
- 19. 7
/
- 2. 1000. 2000, 5000. 4000. 5000, 6000. 7000. 8000.
TIME (SEC)
~
FIGURE 14.3-10B , CORE MIXHTRE HEIGHT (2 Inch) i BEAVER VALLEY POWER STATION UNTT SO. ! UPDATED FINAL SAFETY ANALYSIS REPORT -! 30% Steam Generator Tube Plugging :
REV .' f F 54 ' ; a i > I l 1
- 32. ,, j i
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- 52. j,, ,
i
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- 29. ,
C ' *( ; { 25.' )' l .
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\ - s . i za CO 1 \/ g \j N N U D, 18.
16. 14 12 2000. 2500. 5000. 5500. -4000. B. 500. 1000. 1500. , TIME (SEC) j i i l .~ i FIGURE 143-10C CORE MIXTURE HEIGHT (4 Inch) ' BEAVER VALLEY POWER STATION UNIT NO. 'l -j UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging i
i ~i
.. 7 i
RE\'- ! f
,e ; ..- t 2 ,, - ,
i I 7_- _ AA l, iy l 750.t ; I - 7'* E ;
,7 700, ,
E i 2 675. .
. x i i h
b ' I .!
,j 650. ;
E- l O !' !
$ 625. '
u 600. , 575. . ~ l i t i , i r ' I 550. -- l i 525. 16E+05 0 4000. 8000. .12E+05 l
. TIME (SEC)
I l FIGURE 14.3-11A j CLADDING TEMPERATURE 11UuNSIENT (l.5 i Inch). BEAVER VALLEY POWER STATION UNIT NO. 1 UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging - ,
. .i
~,
i i i i RE\. ; J i
'5:0 .
i
. I 1600.
j j 1400. .
.?
m
- x ,
? 1200.
i 2 l m ' S j p 1000. C a e U S00.. l 600.
/
400. 5000. 1500. 2000. 2500. 3000 3500. '4000. 4500. ; H TIME (SEC) i I
,i I
1 l .. l FIGURE 143-1IB CLADDING TEMPERATURE TRANSIENT (2 Inch) BEAVER VALLEY POWER STATION UNIT NO. I UPDATED FINAL SAFETY ANALYSIS REPORT. 30% Steam Generator Tube Plugging .
a
.I I
REV. i 1
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I 1100. l
- f. .'t!
I ? e-1000. w r t x
?
r 900. i
'j f A 'A-t g
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= G'
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e I l .; c
.a
-*' u 700.. i i 600. ,
\ l i
[ 500. s ./ y , 6
-i 400. 2500.
O. ~500. 1000. 1500. 2000. f TIME (SEC) 'i i
.]
I i FIGURE 14.3-llc CLADDING TEMPERATURE TRANSIENT (4 Inch) BEAVER VALLEY POWER STATION UNIT NO;~ I ; UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging : ;
h REV f l 2$00. , , i 2250 2000.
/ $ 1750. ->
w
=
- tu z 1500. .
D 1250. ( N c \ > 0 M 1000.
\
N , 750. 1 4 500. .
\
250. l O. x _ O. 5. 10. 15. 20. 25. 30. -35. TIME (SEC) FIGURE 14.3-12 0 -l ' REACTOR COOLANT SYSTEM PRESSURE - DECLG (CD = 0.4, MIN SI) I BEAVER VALLEY POWER STATION UNIT NO. ; l UPDATED FINAL ~ SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging
i REi t b "000. , , t . i - OUT _ET 6000. E il
'I C in y
4000. 'I
= 11 M Il
~
\l 2000.
41
< is t x .
s 3 11 , I s 0 - ~ _ 3 z g , v_ __ + W C -2000.
- In let
-4000.
-6000.
-6000. 25. 30. 35.
O. 5. 10. 15. 20. TIME (SEC) FIGURE 14.3-12P CORE FLOWRATE - DECLG (CD = 0.4, MIN SI) BEAVER VALLEY POWER STATION UNTT NO. 1 UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging
RE\ 5000- r 4500 h i
~2 @ 40C0.
O M OO . z Q 3000. 1 2500. N 2000. 7 D U N 1500. 1000
/ .
500. O. O. 5. 10. 15. 20. 25. 30. 35. TIME (SEC) FIGURE 14.3-12Q ACCUMULATOR FLOW DURING BLOWDOWN - DECLG (Co = 0.4 MIN SI) , BEAVER VALLEY POWER STATION UNIT NO. 1 l UPDATED FINAL SAFETY ANALYSIS REPORT I 30% Steam Generator Tube Plugging
REV ; r 10
, t l- -i' 5
~
ww - p N'Nb
= -5 h i 5 41 c -10 : = \
1 O )
-15 9
w
.?
m -20
- x C
U
-25
-30 ,
-35
~
0 5 10 15 20 25 40 35 TIME (SEC) i i FIGURE 14.3-12R ! CORE PRESSURE DROP - DECLG (CD = 0.4 MIN SI) BEAVER VALLEY POWER STATION . UNIT NO. I ; UPDATED FINAL SAFETY ANALYSIS REPORT' i 30% Steam Generator Tube Pluggmg ; 1 I
RE\ P .5E+05
.Si+05 .~
M
~
d
~,, 4E+05 3
O, I i 2 .3E*05 N
.2E+05 1E+05
\ ,
x_- -
.4 0.
\ )
O. 5. 10. 15. 20. 25. 30. 35. ! i TIME (SEC) l i
~
FIGURE 14.3-12S BREAK. FLOW DURING BLOWDOWN - l, DECLG (CD = 0.4, MIN SI) j BEAVER VALLEY POWER STATION UNTT NO. I ! UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging l l
e RE\ , e 1 . +
, ,3E C8 l
t
.25E-C8 ,
t L
- N, ,
C ,2E-CB E
~
i t y 15E-CS -g t Z
= .
M
$ . *E-C8
.. z N ,
~
{ i
<t.07 x N s -
l
- 0. P
,5E-07 0, 5. 10. 15. 20. 25. 40. 35. ;
TIME (SEC) l
-l i
i i FIGURE 143-12T BREAK ENERGY DURING BLOWDOWN - DECLG (CD = 0.4, MIN SI) ! BEAVER VALLEY POWER STATION UNIT NO. 1 UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging i
i, i i R E'- ! l t i l P i
.9 i
.8' {
i 7 x .S - iis n w
^
.5 m !
C ; e 4 !
.3 j 4
l 2 6 l
)
l i ! I w i l
- 0. - ,
0 5 to 15 20 25 20 35 l i TIME (SEC) ) i i FIGURE 14.3-12U ; CORE FOWER - DECLG (CD = 0.4. MIN SI) BEAVER VALLEY POWER STATION UNIT NO. I UPDATED FINAL SAFETY ANALYSIS REPORT- l 30% Steam Generator Tube Plugging - l l 1 i
I R E', i j w
.i 9
0 22 ,
> I
- v. }
20 .......,,........-.........r, . . . , , , , , , , , , , , -
. a.ai. kf I
E 16 -
}. .
,y.
DOWNC0f1ER g 7 16 l ! f '
.* 14 ,
E l l g 12 , 5 10-- -- ' " ' - - ~ - J CORE SATJRATION j l s 175 200 0 25 50 75 100 125 150 TIME AFTER BOC (SEC) ; i FIGURE 14.3-131 CORE AND DOWNCOMER LIQUID LEVELS ! DURING REFLOOD - l DECLG (Co = 0.4, MIN SI) ' BEAVER VALLEY POWER STATION UNIT NO.1 UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging 9'
RE\ I i I 9' , F ! 5 z 5 .
= i
>. 7
$ l x
3 6 5 i l I I
. 4 3
2 1 O O 20 40 60 80 100 120 140 160 180 200 TIME AFTER BOC (SEC)
-- FIGURE 14.3-13J CORE INLET FLUID VELOCITY FOR ROD THERMAL ANALYSIS DECLG (Co = 0.4, MIN SI)
BEAVER VALLEY POWER STATION UNIT NO. I UPDATED FINAL SAFETY ANALYSIS REPORTL 30% Steam Generator Tube Plugging
REV.
.e Q
- 3
- i O
b '4000. O d y 3500. g e 3000. N z . x O 2500. c
~* 2000.
o d [ 1500. 52 1000. 2 500. D U
. u D.0. 25. 50.
< 75. 100. 125. 150. 175. 200.
TIME AFTER BOC (SEC) i
.i l
- i l
FIGURE 14.313K ACCUMLATOR FLOW DURING REFLOOD'- DECLG (Co = 0.4. MIN SI) BEAVER VALLEY POWER STA110N UNIT NO.. I UPDATED FINAL SAFETY ANALYSIS REPORT - 30% Steam Generator Tube Plugging -
r ,i RE\ r _
?
9 318 0 316 E I' 3 j JY M' " " b ,34 _ ! 12 e > x 310 - g b 1 p 3oS , g
& 306 m V
304 302 O 25 50 75 100 125 150 1'/ 5 200 ! i TIME AFER BOC (SEC) t l
.i
?
i l -. 1
- FIGURE 14.3-13L SI FLOW. DURING REFLOOD -
DECLG (CD = 0.4, MIN SI) BEAVER VALLEY POWER STATION UNIT NO. I ; UPDATED FINAL SAFETY ANALYSIS REPORT l 30% Steam Generator Tube Plugging i l
^
'\
REi' r ll 10 6 R u
?! 0 u
; kQ 1 Tv me i
=
d Y 2 s _10 c g . F x 3 -20 m
. . m
- s
-30
-40 -
L 0 25 50 75 100 125-150-175 200 225 250 TIME (SEC) I FIGURE 14.3-14K
" ' MASS FLUX AT111E PEAK ROD TEMPERATURE ELEVATION (V5H IFBA)
DECLG (Co = 0.4, MIN SI)
-BEAVER VALLEY POWER STATION UNIT NO.;l UPDATED FINAL SAFETY- ANALYSIS REPORT 30% Steam Generator Tube Plugging-
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=
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I I l . f I I b, A l O
,l M o
. = W 8 10 1 1.
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a t 2 100 0 25 50 75 100 125 150 175 200 225 250 i f TIME (SEC) 1 a ' i
. FIGURE 14.3-14L ROD HEAT TRANSFER COEFFICIENT ATl PEAK ROD 'EMPERATURE ELEVATION l (V5H IFBA) DECLG (CD = 0.4. MIN SI) I!
BEAVER VALLEY POWER STATION UNTT ,
. UPDATED FINAL SAFETY ANALYSIS REPO 30% Steam Generator Tube Plugging i
I
i i RE\ ; l
',,"n '-
, l
., l
/m :
2000-f a l i t , I
= 1800.
i M '! i i l ;
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- 1
- E !
a
,- u 1400. i 2
' O.w< *
. o [ ;
2 1200. a i W P 1000.
^
'O . 50. 100. .
150. 200. 250. TIME (SEC) i FIGURE 14.3-14M FUEL ROD PEAK CLAD TEMPERATURE - (V5H IFBA) DECLG (CD =.0.4, MIN SI) BEAVER VALLEY POWER STATION UNIT NO. I UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging
I RE\ l
'303
, i 180] '
/ >
G ) ; i t r a 1700. i E n . g 1600. ' t
= ,
It0 j
~
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- \
t\ 1400*
. y (
% i e
e 1300. / S
.- a \
r a . \ . a
< 1200.
t 1100. . 1000. :
- 0. 50. 100. 150. 200. 250. i a
TD4E (SEC)
- I i
FIGURE 14.3-14N CLAD TEMPERATURE AT THE BURST NODE j (V5H IFB A) DECLG (Co = 0.4, MIN SI) BEAVER VALLEY POWER STATION UNIT NO. 1 l UPDATED FINAL SAFETY ANALYSIS REPORT : 30% Steam Generator Tube Pluggmg : l 1
I REV ! l 0 1803. ; N - 1603.
; l -
&d \
S 1400. M%N[I s '
. 1200. ; ,
;: o E
= 1000.
x
= , J E
E c 800. l 2
.9 a
600. 2 W 400. . 200. O. 50. 100. 150. 200. 250. , TIME (SEC) . l
- . FIGURE 14.3-14 0 - '
FLUID TEMPERATURE AT THE PEAK ROD TEMPERATURE ELEVATION , (V5H IFB A) DECLG (Co = 0.4. MIN SI) . BEAVER VALLEY POWER STATION UNTT NO. ! l . UPDATED FINAL SAFETY ANALYSIS REPORT r 30% Steam Generator Tube Plugging y r- -t
REV 2203 . i . 2000.
; g -
' m' f '
- s W 1800. 1 A i 5
=
2 1600. } . p n 1
= J' ?
w 1400. ;
. .a L J c
[ 2 1200. ai , a a 1000. : J
^
800. .
- 0. 50. 100. 150. 200. 250. .
TD4E (SEC) i l i FIGURE 14.3-14P FUEL ROD PEAK CLAD TEMPERATURE - (V5H NON-IFBA) DECLG (Co = 0.4, MIN SI) , BF WER VALLEY POWER STATION UNIT NO 1 UPDATED FINAL SAFETY ANALYSIS REPORT . 30% Steam Generator Tube Plugging . !
y \ r : RE\ l
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i
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# ;(
' 3 i i
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) iv i
i t c 1700. ( E E 1600. / ! l b f h 1500. y ! E a 1400. - U i
- e
. 5 1300. i e > l a f 3 1200. l g -
a l
=- 1100. .
1000. l 250.
~
000'O . 50. 100. 150. 200. ; TIME (SEC) l r
~
i
- FIGURE 14.3-14Q FUEL ROD PEAK CLAD TEMPERATURE -
(S'ID NON-IFBA) DECLG (CD = 0.4. MIN SI) BEAVER VALLEY POWER STATION USTr NO. 1 UPDATED FINAL SAFETY ANALYSIS REPORT 30% Steam Generator Tube Plugging
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'I non-LOCA UFSAR- Page Changes (Complete Loss of Flow) ..
t f, ages Ficures 14.2-46 14.2-31 through -33 - 14.2-47 Table 14.2-5 . f a 9 s P
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BVPS-1-UPDATED FSAR Rev. 8 (1/90) enrichment than In the e single pin or pellet has a : the nominal va consequences in of reduced DNBR and . increased fuel an temperat. 1 be limited to the incorrectly loaded pin or p Tuel assembly leading s evented by administrative procedures implemente ing core 1 In the unlikely event that a loading e occurs, analyses in action confirm that resulting pow ribution effects will eith sadily detected by the i moveable detector system or will ca sufficiently small ation to be acceptable within the uncert s allowed ! be ominal and design power shapes. I comnlete Less of Forced Reactor Coolant Flow 14.2.9 14.2.9.1 Accident Description A complete loss of forced reactor coolant flow may result from a i simultaneous loss of electrical supplies to all reactor coolant : I pumps. If the reactor is at power at the time of the accident, the immediate effect of loss-of-coolant flow is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor were not tripped promptly.
, l This event is classified as an ANS Condition III incident. The following provida necessary protection against a loss-of-coolant flow accident: , 1. Undervoltage or underfrequency on reactor coolant pump power supply busses
- 2. Low reactor coolant loop flow.
- ADb inSEhtT L -
The reactor trip on reactor coolant pump bus undervoltage is provided to protect against conditions which can cause a loss of voltage to all reactor coolant pumps, i.e., station blackout.
- ADOIMEtT 3 -
The reactor trip on reactor coolant pump underfrequency is provided to open the reactor coolant pump breaker and trip the reactor for an underfrequency condition, resulting from frequency disturbances on i the major power grid. The trip disengages the reactor coolant pumps from the power grid so that the pump kinetic energy is available for full coastdown. If the maximum grid frequency decay rate is less than approximately 5 Hz/sec., this trip function will protect the core from underfrequency events without requiring tripping of the RCP breakers. Reference 17 provides analyses of grid frequency disturbances and the resulting nuclear stem supply system protection requirements which are generally applicable. The reactor trip on low primary coolant loop flow is provided to protect against loss of flow conditions which affect only one reactor coolant- loop. It also serves as a backup to the undervoltage and underfrequency trips. This function is generated by two out of three low flow signals per reactor coolant loop. Above approximately 30 percent power (Permissive 8), low flow in any loop will actuate a 14.2-46
BVPS-1-UPDATED FSAR Rev. 8 (1/90) reactor trip. Between approximately 10 percent power and 30 percent . power (Permissive 7 and Permissive 8), low flow in any two loops will actuate a reactor trip. Nor=al power for the reactor coolant pumps is supplied through busses from a transformer connected to the generator. Each pump is on a separate bus. When generator trip occurs, the busses are i automatically transferred to a transformer supplied from external ' power lines, and the pumps will continue to supply coolant flow to the core. Following any turbine trip, where there are no electrical faults which require tripping the generator from the network, the generator remains connected to the network for approximately 30 seconds. The reactor coolant pumps remain connected to the generator thus ensuring full flow for 30 seconds after the reactor trip before any transfer is made. 14.2.9.2 Method of Analysis Tyrs AMLys& HAS SCCD turkraHtED n nouco UP T* 30% sT5Wr' Gract4Tck ToeE PLMCWG . The complete loss of flow transient has been analyzed for a loss of i all three reactor coolant pumps with three loops in operation. --ADO lusaT3 , The method of analysis and the assumptions made regarding initial operating conditions and reactivity coefficients are identica1'to those discussed in Section 14. Figures 14.2-30 through 14.2-33 show the transient response for the loss of power to all reactor coolant pumps with three loops in i operation. The reactor is assumed to be tripped on ep 6 e i signal. Figure 14.2-33 shows that the minimum DNBR i always greater than the safety analysis limit. A wat4c" * :oowT"xPFlow + Since DNB does not occur, the ability of the prima coolant to - remove heat from the fuel rod is not greatly reduce. Thus, the average fuel and clad temperatures do not increase significantly above their respective initial values. The calculated sequence of events for the case analyzed is shown in Table 14.2-5. The reactor coolant pumps will continue to coast down and natural circulation flow will eventually be established. With the reactor tripped, a stable plant condition will eventually be attained. Normal plant shutdown may then proceed. 14.2.9.3 Conclusions The analysis performed has demonstrated that for the complete loss of
. forced reactor coolant flow, the DNBR does not decrease below the limit value at any time during the transient. Thus, the DNB design basis as described in Section 3 is met. There is no fuel clad damage or release of fission products to the Reactor Coolant System. L 14.2-47
3 i INSERT 1 ;
- 3. Overpower AT Reactor Trip Function.
t
- t INSERT 2 g A
However, due to the possibility of a common mode failure in the cabinets containing circuitry the
- Undervoltage and Underfrequency reactor trip functions may not be available. The Undenoltage and i Underfrequency trip functions are blocked below approximately 10% power (Permissive P-7). If the Undervoltage and Underfrequency reactor trips are not available, the low reactor coolant flow trip is the next available reactor trip and would become the primary reactor trip function. Reactor protection '
system diversity is provided by the overpower AT reactor trip function. l INSERT 3 ' The analysis assumes a common mode failure which results in the unavailability of the Undenoltage and Underfrequency reactor trips. Thus, the low reactor coolant loop flow reactor trip is the primary reactor trip function for the complete loss of flow analysis. J e 1 l I l i
BVPS-1-UPDATED TSAR Rev. 8 (1/90) TABLE 14.2-5 TIME SEOUENCE OF EVENTS FOR CONDITION TTT EVENTS Accident Event Time (see) Complete Loss of Forced Reactor Coolant Flow (3 Loops Operating, 3 RCPs Coascing Down) Coastdown Begins 0.0 Low !s4 cmc cousuT VLCW Mr "-fcrc.1tz;e. Trip Point Reached -Gre-/.6 Rods Begin to Drop 4,5-Q.6 Minimum DNBR Occurs 3.A d. b m M e e 1 of 1
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- FOR THREE LOOPS IN OPERATION, THREE PUMPS COATING DOWN, COMPLETE LOSS OF FLOW BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT i i
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, PRESSURIZER PRESSURE TRANSIENT-FOR THREE LOOPS IN OPERATION, 1 THREE LOOPS COASTING DOWN, COMPLETE LOSS OF FLOW BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT l
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3 TIME (SEC) .. t FIGURE 14.2-31 , NUCLEAR POWER TRANSIENT AND. PRESSURIZER PRESSURE TRANSIENT FOR THREE LOOPS .IN OPERATION,- ; THREE LOOPS COASTING DOWN, 1 COMPLETE LOSS OF FLOW ! BEAVER VALLEY POWER STATION UNIT UPDATED FINAL SAFETY ANALYSIS REPORT
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FIGURE 14.2-32 i AVERAGE AND HOT CHANNEL HEAT i FLUX TRANSIENTS FOR THREE LOOPS IN OPERATION, l THREE PUMPS COASTING DOWN, COMPLETE LOSS OF FLOW BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT i
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FIGURE 14.2-32 AVERAGE ANO HOT CHANNEL HEAT FLUX TRANSIENTS FOR THREE LOOPS- IN OPERATION, THREE PUMPS COASTING OOWN, COMPLETE LOSS OF FLOW BEAVER VALLEY POWER STATION UNIT 1 UPOATED FINAL SAFETY _ ANALYSIS RE? ORT
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. FIGURE 14.2-33 DNBR VS TIME FOR THREE LOOPS IN OPERATION, THREE PUMPS COASTING DOWN, COMPLETE LOSS OF FLOW BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT !
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~
BEAVER VALLEY. POWER STATION UNITJ1--
. UPDATED FINAL SAFETY ANALYSIS ~ REPORT
h i i l i non-LOCA UFSAR Page Changes (Partial Loss of Flow) ! Paces Ficures 14D-3' 14.1-13 through -16 ;
-i 14D-14 14D-15 .. .
14D-16 ; 14D-17 .j 14D-46 Table . 14.1-2
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.. power capab1_ . This higher the power value is designated as ESF design rat This power output ,
includes the the. power gener y the reactor coolant pumps. - Note: The thermal power value for each transient analyzed l are given in Table 14D ; Where initial power ope g condit are assumed in accident i analyses, the "guaran Nuclear Steam ly System thermal power output" plus allowa or errors in steady te power determination demonstration of adequac the containment and is assumed. ESF are con , the " Engineered Safety F res design rating" plus allowan or error is assumed. ; 14D.3.2 Initial conditions For accident evaluation, the initial conditions are obtained by adding maximum steady state errors to rated values. The following steady state errors are considere4-T-- ADD ImEPJ F -
- - 1. Core power 2 percent allowance for calori- i metric error i
- 2. Average RCS tamparature 14*F allowance for deadbank and
- measurement error
- 3. Pressurizar pressure 130 psi allowance for steady state fluctuations and measurement ;
error Initial values for core power, average RCS temperature and pressurizar pressure are selected to minimize the initial DNBR unless; otherwise, stated in the sections describing specific accidents. The outer surface of the fuel rod at the hot spot state operates at a operation at temperature of approximately 660*F for steady of nucleate rated power throughout core life due to the onset boiling. Initially (beginning of life), this_ temperature is that of the cladding metal outer surface. During operation over the life of the core, the buildup of oxides and crud on the fuel rod surface causes the cladding surface temperature to increase. Allowance is made in the fuel center melt evaluation for this temperature rise. , basis limits DNB, adequate heat l Since the thermal-hydraulic design j transfer is provided between the fuel cladding the reactor coolant so that the core thermal output is not limited by considerations of the l Figure 3.4-4, shows the axial variation of i cladding temperature. average cladding temperature for a typical rod (17 x 17 fuel l assembly) both at beginning and and of life. End of life is after ; three typical cycles of operation (approximately 20,000 effective l full-power hours). These are calculated using the l paratureswhich has been reviewed and approved Westinghouse fuel rod modely by the NRC.
D-3 _ _ _ _ _ . _ _ _ _ _ _ _ _ _ .
BVPS-1-UPDATED FSAR Rev. 8 (1/90) , References to Accendir 14D
~
- 1. Supplemental information on fuel design transmitted frem l Salvatori, Westinghouse, to D. Knuth, Atomic Energy l R.
Commission, as attachments to letters NS-SL-518 (12/22/72), I (1/4/73) NS-SL-521 (1/4/73), NS-SL-524and supplemental information and NS-SL-543 (1/12/73), l (Westinghouse Proprietary); on fuel i I design transmitted from R. Salvatori, to D. Knuth, as attach =ents to letters NS-SL-527 (1/2/73) and NS-SL-544 (1/12/73). j
- 2. K. Shure, " Fission Product Decay Energy" WAPD-BT-24, pp. 1-17, ,
Westinghouse Bettis Atomic Power Laboratory (December 1961) . ,
- 3. K. Shure and D. J. Dudziak, " Calculating Energy Released by Products," Trans. American Nuclear Society, 4 (1) p. 30 Fission (1961). ;
Teake, United Kingdom Atomic Energy Authority Decay Heat 4. Standard. (Private Communication)
- 5. J. R. Stahn and E. F. Clancy, " Fission-Product Radioactivity and !
Heat Generation" in " Proceedings of the Second United Nations the Peaceful Uses of Atomic Energy, International Conferance on
- Geneva, 1958," Volume 13, pp. 49-54, United Nations, Geneva, 1958. !
- 6. F. E. Obenshain and A. H. Foderaro, " Energy from Fission Product Decay," WAPD-P-652, Westinghouse Bettis Atomic Power Laboratory (1955).
7. MF - REPLACE WITH UJsskT D - " l 8. Deleted By Revision 8. M. Gutz, " MARVEL (A. Digital Computer Code for Transient -
- 9. J.
Analysis of a Multi-loop PWR System)," WCAP-7909, Westinghouse Electric Corporation (June 1972) . C. Baker, R. P. Rose,
~
- 10. T. W. T. Burnett, C. J. McIntyre, J.
Westinghouse Electric l "LOFTRAN Code Description," WCAP-7907, Corporation (June 1972) .
- 11. R. F. Barry, " LEOPARD, a Spectrum Dependent Non-Spatial Depletion Westinghouse Electric IBM-7094," WCAP-3269-26, ,
Code for the Corporation (September 1963).
- 12. R. F. Barry and S. Altomare, "The TURTLE 24.0 Diffusion Depletion (June 1968);
WCAP-7213, (Westinghouse Proprietary) Code": WCAP-7758 (September 1971). 14n-14
BVPS-1-UPDATED FSAR Rev. 10 (',92).
- t. mount of time available for the operator to determine caus f the dilutien, isolate tne reactor water makeup s. e, and 1.. te boration before the available shutdown r .n is lost.
With the rea :n manual control and no operator .on taken to terminate the _ sient, the power and temperatu- .se will cause the reactor to en the Overtemperature D -T trip setpoint resulting in a rea trip. The boren dilu* transient in this case is essentially equivalent to controlled RCCA bank withdrawal at power. . maximum reac* y insertion rate for a bcron dilution is conse. ively ated to be 2.5 pcm/sec, which is within the range inse- rates analyzed. Thus, the effects of dilution prior . ctor trip are bounded by the Uncontrolled ROCA Bank Withdr at Power analysis (FSAR Section 14.1.2). Following reacto .r. there are at least 16 minutes prior to criticality. ' is t.. mount of time available for tne operator to dete--
*he cause the dilution, isolate the reactor water =akeup arce, and in. te boration before the available shutdown ma is l'4t.
14.1.4.3 Conclusi , Because of tP procedure involved in the dil 'on process, an inadvertent .ution is considered to be h unlikely. Neverthele if an unintentional dilution of boron he reactor coolant occur, numerous alarms and indications a. vailable
, to ale the operator to the condition. Furthermore, t. maximum reac* .y addition due to the dilution is slow enough - allow tr . erator to determine the cause and take corrective ion re shutdown margin is lost.
14.1.5 Partial toss of Forced Reactor Coolant Flow 14.1.5.1 Identification of Causes and Accident Description A partial loss-of-coolant flow accident can result from a mechanical or electrical failure in a reactor coolant pump, or from a fault in the power supply to the pump. Each RCP is supplied by a separate bus. If the reactor is at power at the j time of the accident, the immediate effect of loss-of-coolant flow l is a rapid increase in the coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor is not tripped promptly. classified as an ANS Condition II incident. The ~ This event is necessary protection against a partial loss-of-coolant flow accident is provided by the low primary coolant flow reactor trip which is actuated by two out of three low flow signals in any , reactor coolant loop. Above approximately 30 percent power l (Permissiva 8), low flow in any loop will actuate a reactor trip. l Between approximately 10 percent power (Permissive P-7) and the ' power level corresponding to Per=issive P-8 low flow in any two loeps will actuate a reactor trip. Above P-7, two or more RCP 14.1-15
BVPS-1-UPDATED FSAR Rev. 10 (1/92) circuit breakers opening will actuate a reactor trip which serves as a backup to the low flow trip. Normal power for the pumps is supplied through buses connected to the main generatcr. Each pu=p is on separate bus. When a l
- generator trip occurs, the pu=ps are automatically transferred to and the pumps will bus supplied from external power lines, a
flow to the core. Following any continue to supply coolant are no electrical faults which require turbine trip wnere there network, the generator remains tripping the generator from the 30 seconds. The connected to the network for approximately connected generator thus reactor coolant pumps remain to the ensuring full flow for approximately 30 seconds after the reactor trip before any transfer is made. 14.1.5.2 Analysis of Effects and Consequences Method of Analysis The loss of one reactor coolant pump with three loops in operation has been analyzed. - Aco /mERT A - This transient is analyzed by three digital computer codes: 1) o the LOTTRAN Code (Reference 3) is used to calculate the loop and core flow during the transient, the time of reactor trip based on the calculated flows, the nuclear power transient, and the primary system pressure and temperature transients; 2) the FACTRAN Code (Reference 2) is then used to calculate the heat flux transient based on the nuclear power and flow from LOFTRAN; and 3) the THINC Code (Reference 9) is used to during calculate the departure from nucleate boiling ratio (DNBR) the transient based on the heat flux determined by FACTRAN and the flow determined by LOFTRAN. The DNBR transient presented represents the minimum of i the typical or thimble cell. Initial Conditions Initial operating conditions assumed are the most adverse with respect to the margin to DNB, i.e., maximum steady state power level, minimum steady state pressure, and maximum steady state for explanation coolant average temperature. of initial conditions.) 1 (See Section 14D.2AAFutTHER
- Apu MERT G -
peactivity Coefficients A conservatively large absolute value of the Doppler-only power The total integrated coefficient is used (See Table 14D-2). Doppler reactivity from 0 to 100 percent is assumed to be 0.016
. k.
The lowest absolute magnitude of the moderator temperature coefficient (0.0 ok/F) is assumed since this results in the maximum hot-spot heat flux during the initial part of the transient when the minimum DNBR is reached. ( 14.1-16 ,
SVPS-1-UPDATED TSAR Rev. 10 (1/92) Flow Coastdown o The flow coastdown analysis is based on a mementum balance around each reactor coolant loop and across the reactor core. Tnis sementum balance is combined with the continuity' equation, a pump
=c=entum balance and the pump characteristics and is based on high esti=ates of system pressure losses.
Results Figures 14.1-13 through 14.1-16 snow the transient response for the loss of a reactor coolant pump with three loops in operation. Figure 14.1-16 shows the DNBR to be always greater than the safety analysis limit. Since DNB does not occur, the ability of the primary coolant to remove heat from the fuel rod is not greatly reduced. Thus, the average fuel and clad temperatures do not increase significantly ; above their respective initial values. The calculated sequence of events for the case analyzed is shown in Table 14.1-2. The affected reactor coolant pump will continue to coast down and the core flow will reach a new equilibrium value associated with the two remaining operating pu=ps. With the ; reactor tripped, a stable plant condition will eventually be attained. Normal plant shutdown may then proceed. conclusions The analysis shows that the DNBR will not decrease below the limit value at any time during the transient. Thus, the DNB design basis as described in Section 3 is met and there will b'e no cladding damage and no release of fission products to the Reactor Coolant System.
. Startuo of an Inactive Peactor Coolant Loco 14.1.6. ntification of Causes and Accident De tion
-l 14.1.6.1.1 Wi - Stop Valves Open l If the plant is op-g with one out of service, there is reverse flow through . oop o the pressure difference !
across the reactor vesse . cold leg temperature in an ' inactive loop is' identical e cold leg temperature of the active loops (the reactor in mperature). If the' reactor is operated at power,
- is a tem re drop across the steam generator in the ina loop and, with reverse flow, the hot leg temperature 4 e inactive loop is r than the reactor core inlet tem re.
Administr procedures require that the unit -ought to a load Aess than 25 percent of full power prior arting a pump an inactive loop in order to bring inactive lo t leg 14.1-17
7 B'/PS-1-UPCATED PSAR Rev. 10 (1,92) Feferences to Section 14,1
- 1. D. B. Fairbrother, H. G. Hargrove, " WIT-6 Reactor Transient Analysis Computer Program Description", WCAP-7980,
, Westingheese Electric Corporation (November 1972).
n
- 3. Burnett, T. W. T., et al., "LOFTRAN Code Description,"
WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Preprietary), April 1984. -
- 4. S. Altomare, R. F. Barry, "The TURTLE 24.0 Diffusion Cepletion Code", WCAP-7758, Westinghouse Electric Corporation (September 1971).
S. F. M. Bordelon, " Calculation of Flow Coastdown After Loss of ' Reactor Coolant Pu=p (PHOENIX Code) ", WCAP-7969, Westinghouse Electric Corporation (September 1972).
- 6. M. A. Mangan, " Overpressure Protection for Westinghouse Pressurized Water Reactors", WCAP-7769, Westinghouse Electric Corporation (October 1971).
- 7. J. M. Geets, R. Salvatori, "Long Tern Transient Analysis Program for PWR's (BLKOUT Code), WCAp-7898, Westinghouse Electric Corporation (June 1972). *
- 8. J. M. Geets, " MARVEL - A Digital Computer Code for Transient Analysis of a Multiloop PWR System", WCAP-7909, Westinghouse Electric Corporation (June 1972).
- 9. J. Shefcheck, " Application of the TMINC Program to PWR !
4 Design", WCAP-7359-L, (August, 1969), Westinghouse Electric ; Corporation (Proprietary), and WCAP-7838, Westinghouse l Electric Corporation (January, 1972). l
- 10. D. H. Risher, Jr., R. F. Barry, " TWINKLE - A Multi-Dimensional l Neutron Kinetics Computer Code", WCAP-7979, Westinghouse l Electric Corporation (November, 1972).
9 14.1-46
133LE 14.1-2 fCONT'DL TTvr SECUrNCT OF ENVENTS FOR CONDITTON IT EVENTS l Accident EYRDE Time fsee) : Partial Loss of.1 Forced Coastdown begins o
- Coolant Flow (3 Loops operating, 1 RCP Coasting Low Flow Reactor Trip ' ' }, q Down)
Rods Begin to Drop 2.L Q,9 Minist.m DNBR occurs : . '_ i,1 l [ 5 i i
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7 , , ,, 71 W ' (3880003 FIGURE 14.H3 FLOW TRANSIE FOR PARTIAL LOSS O FLOW-THREE LOOPS IN OPE TION, i ONE PUMP COASTING DOWN i BEAVER VALLEY POWER STATION UNIT. I ! UPDATED FINAL SAFETY ANALYSIS REPORT : i _ __. __ -___ - _ ___- - __A
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d 06 - i b sW 04 - h 1 02 0 2 4 6 8 10 TME (SEC) FIGURE 14.1-13 . REACTOR VESSEL AND FAULTED LOOP FLOW TRANSIENTS FOR PARTIAL LOSS OF FLOW - ONE PUMP COASTING DOWN BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT
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- NUCLEAR POWER AND R(S PRESSURE FOR-PARTIAL LOSS OF FLOW !
THREE LOOPS OF OPERATION, i DNE PUMP COASTING ~ 00WN BEAVER VALLEY POWER ST4 TION UNIT 1 UPDATED FINAL SAFETY ANALYSIS . REPORT
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TIME (SEC) i FIGURE 14.1-14 NUCLEAR POWER AND PRESSURIZER PRESSURE FOR PARTIAL LOSS OF FLOW - ONE PUMP COASTING DOWN BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT
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FIGURE 14.1-15 AVERAGE AND HOT C NEL HEAT FLUX TRANSIENT FOR PARTI LOSS OF FLOW
~
THREE LOOPS IN OPERA N, ONE PUMP COASTING 00WN BEAVER VALLEY POWER STA DN UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT
g 12 b 3 _ O h oB - A 3
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^0 4 6 8 10 0 2 TIME (SEC)
FIGURE 14.1-15
*' AVERAGE CHANNEL AND HOT CHANNEL HEAT FLUX TRANSIENTS FOR PARTIAL LOSS OF FLOW - ONE PUMP COASTING DOWN BEAVE9 VALLEY POWER STATION UNIT 1 UPDAT5D FINAL SAFET( ANALYSIS REPORT
E ' ; REPLks worn PST gg ^' EXT /*% GE ou f 2.-
~
2.20 1.75 . 1.50 - i a: a 1.25 - 1.00
. .750-
.500 . . /
.250 0.000 -
0 1 2 3 4 5 I TIME ( ON05) 1
)
FIGURE 1-16
~
DER VE TIME FOR ; l PARTIAL L OF FLOW
- . TlftEE LOOP OPERATION, i OE PUW C ING 00WN l BEAVER YALLEY WER STATION UNIT 1C UPDATED FINAL ETY ANALYSIS REPDFj
3 l m 2.5 - W E
$ 2 -
O 1.5 -
=
0 2 4 6 8 10 TIME (SEC) FIGURE 14.1-16
. DNBR VERSUS TIME FOR PARTIAL LOSS OF FLOW - ONE PUMP COASTING DOWN BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT
non-LOCA UFSAR Page Changes
~
(Locked Rotor) i Paces Ficures i 14.2-40 14.2-15 through -18 14.2-41 14.2-42 14.2-43 14.2-44 14.2-51 14.2-52 Table 14.2-4A l i f t l
.. e .
BVPS-1-UPDATED FSAR Rev. 8 (1/90) 3 Conclusions 4
. Referenc ^ws limiting hot channel factors for a given - .ty insertion vn. et the limiting criteria of Se *' . 6.1.2.
The peak hot c. factors obtained for
- incidents in Section 14.2.6.2.3 a below the limiti- ..annel factors and thus meet all of the limi. iteria.
Even on a pessimistic ba ' analyses indicated that the described fuel and c1 .s t exceeded. It is concluded that there is no . of sudden fu ersal into the coolant. does not exce at which would cause Since the pea sure limits, t ed the faulted condition - it is stresses t there is no danger of further con tial damage to conclud the ry circuit. The analyses have demonstrated upper limit o sion product release as a result of a numbe uel rods ering DNB amounts to 10 percent. 14.2.7 Sincie Reactor Coolant Puro Locked Rotor 14.2.7.1 Identification of Causes and Accident Description The accident postulated is an instantaneous seizure of a reactor
. coolant pump rotor. Flow through the affected reactor coolant loop is rapidly reduced, leading to an initiation of a reactor trip on a low flow signal.
. Following initiation of the reactor trip heat stored in the fuel rods l continues to be transferred to the coolant causing the coolant to expand. At the same time, heat transfer to the shell side of the staam generators is reduced, first because the reduced flow results in a decreased tube side film coefficient and then because the side reactor coolant in the tubes cools down while the shell temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the coolant in the reactor core, combined with reduced heat transfer in the steam generators causes an insurge into the pressurizer and a pressure increase throughout the Reactor Coolant System. The insurge into the pressurizer compresses the steam volume, actuates the automatic spray system, opens the l safety power operated relief valves, and opens the pressurizer valves, in that sequence. The three power operated relief valves are !
designed for reliable operation and would be expected to function , However, for conservatism, their . properly during the accident. pressure reducing effect as well as the pressure This reducing effect of event is classified l the spray is not included in the analysis. as an ANS Condition IV incident. , 14.2.7.2 Analysis of Effects and consequences 14.2.7.2.1 Method of Analysis The Two digital computer codes are used to analyze this transient. (Reference 11) is used to calculate the resulting loop I LOFTRAN Code ' and core flow transients following the pump seizure, the time of reactor trip based on the loop flow transients, the nuclear power ; 14.2-40 l
3V?S-i-CPDATED FSAR Rev. 8 i, ? C i following reactor trip, and to determine the peak pressure. .The thermal behavior of the fuel located at the core hot spot is investigated using the FACTRAN Code (Reference 8), which uses the core flow and nuclear power calculated by LOITRAN. The FACTRAN Code includes a film boling heat transfer coefficient.
- ADD sustKr A - ^i At the beginning of the postulated locked rotor accident, i.e., at . the time the shaft in one of the reactor coolant pumps is assumed to seize, the plant is assumed to bs in operation under the most adverse steady state operating conditions, i.e., maximum steady state power level, =aximum steady state pressure and maximum steady state coolant average temperature. - ADD InszKT c - - ADD susGRT E -
When the peak pressure is evaluated, the initial pressure is conservatively esti=ated as 30 psi above nominal pressure (2,250 psia) to allow for errors in the pressurizar pressure measurement and 4 control channels. This is done to obtain the highest possible rise in the coolant pressure during the transient. To obtain the maximum pressure in the primary side, conservatively high loop pressure drops are added to the calculated pressurizar pressure. The pressure response shown in Figures 14.2-16 through 14.2-17 are the responses at the point in the Reactor Coolant System having the maximum pressure. I The maximum pressure in the reactor coolant systan following a locked rotor accident occurs at the pump discharge of pumps which have not experienced the locked rotor. Evaluation of the Pressure Transient After pump seizure, the neutron flux is rapidly reduced by control rod insertion effect. Rod motion is assumed to begin one second after the flow in the affected loop reaches 87 percent of nominal flow. The time delay of 1.0 second used in connection with the low i flow reactor trip is a vary conservativa allowance for the total time delay between the time the flow raschas 87 percent of nominal and the time the rods begin moving into the core. This total includes j delays associated with the follcwing: flow ' individual sensors / transmitters, solid state protection system input relays, solid state protection system, voltage drop on reactor trip breaker undervoltage and control rod gripper ralsasa. No credit is taken for l valves, the pressure reducing effect of the prassurizar relief , pressurizar spray, steam dump or controlled feedwatar flow after ' plant trip. Although these operations are axpected to occur and would result in a lower peak pressure, an additional degree of conservatism is provided by ignoring their effect. The pressurizar safety valvas are actuated at 2,575 psia and their capacity for staan relief is as stated. 14.2-41
I l BVPS-1-UPDATED FSAR Rev. B (1/90) l Evaluation of the Effects of DNB in the Core Durina the Accident For this accident, DNB is assumed to occur in the core and, therefore, an evaluation of the consequences with respect to fuel red thermal transients is performed. Results obtained from analysis of this " hot spot" condition represent the upper limit with respect to clad te=perature and zirconium water reaction. In the the rod power at the hot spot is conservatively assumed evaluation,/timestheaveragerodpower(i.e., F(Q) = 2.54/) to be 2.54 at the initial core power level. The number of rods in DNB was conservatively calculated to be 18% of the total rods in the core. Film Boilina Coefficient The film boiling coefficient is calculated in the FACTRAN code using the Bishop-Sandberg-Tong film boiling correlation. The fluid properties are evaluated at film temperature (average between wall and bulk temperatures). The program calculates the film coefficient at every time step based upon the actual heat transfer conditions at the time. The neutron flux, system pressure, bulk density and mass flow rate as a function of time are used as program input. For this analysis, the initial values of the pressure and the bulk density are used throughout the transient since they are the most conservative with respect to clad temperature response. For
- conservatism, DNB was assumed to start at the beginning of the accident.
Puel Clad Gao Coefficient The magnitude and time dependence of the heat transfer coefficient between fuel and cladding (gap coefficient) has a pronounced influence on the thermal results. The larger the value of the gap I coefficient, the more heat is transferred between pellet and clad. ! Based on investigations on the effect of the gap coefficient upon the maximum clad tamperature during the transient, the gap coefficient l was assumed to increase from steady state valug*consistant with the initial fuel temperature to 10,000 Btu /hr-ft F at the initiation of the transient. This assumption causes energy stored in the fuel to be released to the clad of the initiation of the transient and i maximizes the clad temperature during the transient.
- zirconium-steam Reaction The zirconium-steam reaction can become significant above 1,800*F (clad temperature). The Baker-Just parabolic rate equation shown below is used to define the rate of the zirconium steam reaction.
6 dfw 2) = 33.3 x 10 exp 45,500 dt 1.986T (14.2-3) 14.2-42 ,
BVPS-1-UPDATED FSAR Rev. 8 (1/90) where:
. v = amount reacted (mg/cm 2) t = time (sec)
! T = temperature (K) The reaction heat is 1,510 cal /gm The effect of zirconium-steam reaction is included in the calculation of the " hot spot" cladding temperature transient. 14.2.7.2.2 Locked Rotor Results The transient results with and without offsite power available are shown in Figures 14.2-15 through 14.2-18. The results of these-calculations are also summarized in Table 14.2-4a. .The peak Reactor Coolant System pressure reached during the transient is less than that which would .cause stresses to exceed - the faulted condition stress limits. Also, the peak clad surface temperature - is ' l considerably .less than 2700 degrees F. It should also be noted that- l the clad temperature was conservatively calculated assuming that DNB 1 occurs at the initiation of the transient. The calculated sequence of events is shown on Table 14.2-4a.. Figure- A
. 14.2-15 shows that the core flow rapidly reaches a new equilibrium- _
l value for the case with offsite power avsilable. With the reactor y tripped, a stable plant will eventually be attained. Normal' plant- q ' shutdown may then proceed. i Conclusions
- 1. Since the peak Reactor Coolant System pressure,-reached-during any of the transients is less than that which would cause stresses to exceed -the' faulted condition stress-limits, the integrity of the primary coolant system is'not endangered.
- 2. Since the . peak- clad surface _ temperature calculated forLthe
- j hot spot during the worst transient remains considerably ths.n 2,700* F_ and the amount of Zirconium-water'.
less reaction is small and the core willLremain'in place and-intact. with no consequential loss of core cooling capability. The radiological consequences of a single reactor. coolant pump locked:. 4 rotor were . analysed to:show compliance with the siting guidelines of-10 CFR 100 using methodch = consistent _-with Standard Review Plan Sections 15.3.3 - 15.3.4.- LThe percent of fuel rods projected'to experience Dets was determined"to be.EL "O 12:22 2. ;i;'. 5 02:1 ..d C
--/-%xThe projected :, r-' let based: on conservative fuel rod power census.
184 value' is expected to bound: future fuel cycles.-~ Although the peak clad temperature of the fuel will not exceed 2700 Degrees F, all of the fuel rods experiencing DItB are~ conservatively, assumed - to-fail. The radiological analysis assumes A the. instantaneous coincident loss of AC. release _of 184_.of gap activity to the RCS. 14.2-43
---- _ _ _ _ _ _ _ _ _ _ . . _ . {
l l i l BVPS-1-UPDATED FSAR Rev. 8 (1/90) power to station auxiliaries is assumed, resulting in an 8 hourtoplant
. cooldown via steam release from the secondary system the at=osphere. Table 14.2-4b tabulates significant analysis parameters.
The analysis projected 0-2 hour doses at the exclusion area boundary of 2.3 rem whole body, 1.4 rem beta skin, and 21.6 rem thyroid. The isotopic releases for the accident are tabulated on Table 14.2-4c. The dose results of the analyses are a small fraction of the 10 CFR 100 exposure guidelines and are, therefore, acceptable. 2.8 Inadvertent Leadina of a Puel Assembiv into an Incrocer Positi2D dentification of causes and Accident Deze on 14.2. and loading errors such as can ar rom the inadvertent Puel e loading of a or more fuel assemblies o improper positions, loading a fue od during manufacture vi e or more Fellets of the loading of full fuel assembly during wrong enrichme or the manufacture with llets of the b enrichment will lead to if the arr esults in placing fuel in core l increased heat f1 I positions calling to ual of les enrichment. Also included among trors the inadvertent loading of one or possible core loadin more fuel assemblies ir urnable poison rods into a new core without burnable poison r error in enrichment the normal manufacturing tolerances, Any can cause power shape whic e more peaked than those calculated with the correct en ents. The incore system of moveable flux detectors which is ed to vari over shapes at the start of life is capable of rev ing any ass y enrichment error or loading power shapes to b aked in excess of the design error which caus value. To reduce th robability of core loadin ors, each fuel assembly is marked vi an identification number and dad in accordance with During core loa , the identification a core loa g diagram. number vil be checked before each assembly i ved into the core. ently recorded on Serial n rs read during fuel movement are sub acing after the the load diagram as a further check on prope loading completed. of a laced fuel The p r distortion due to any combination would be e asse as would significantly raise peaking factors to the j read observable with in-core flux monitors. In addi about l onitors, thermocouples are located at the outle flu core. There i high ' ird of the fuel assemblies in the one thermocouples would also indica any ; pr ility that these re rmally high coolant enthalpy rise. In-core flux measurement a en during the startup subsequent to every refueling operation. 14.2-44
BVPS-1-UPDATED FSAR Rev. 8 (1/90) References to Section 14.2
- 1. J. M. Geets, " MARVEL - A Digital Computer Code for Transient Analysis of a Multiloop PWR Systam," WCAP-7909, Westinghouse Electric Corporation (June 1972).
- 2. F. S.' Moody, Transactions of the American Society of Mechanical Engineers, " Journal of Heat Transfer",
(February 1965), Figure 3, page 134.
- 3. T. W. T. Burnett, " Reactor Protection System Diversity in Westinghouse Pressurized Water Reactor," WCAP-7306, Westinghouse Electric Corporation (April 1969).
- 4. T. G. Taxelius, ed. " Annual Report-Spert Project, October .
1968 September 1969", IN-1370, Idaho Nuclear Corporation (June 1970).
- 5. R. C. Liimatainen and F. J. Testa, " Studies in TREAT of Zircaloy-2-Clad UO-Core Simulated Fuel Elements", January -
June 1966, ANL-7225, 177, Argonne National Laboratory (November 1966).
- 6. D. H. Risher, Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," WCAP-7588, Revision 1, Westinghouse Electric Corporation (December 1971) .
- 7. D. H. Risher, Jr., R. F. Barry, " TWINKLE -
A Multi-Dimensional Neutron Kinetics Computer Code", WCAP-7979, Westinghouse Electric Corporation (November 1972). 8.
- 9. A. A. Bishop, R. O. Sandberg and L. S. Tong, " Forced convection Heat Transfer at High Pressure After the critical Heat Flux," 65-HT-31, American Society of Mechanical Engineers (August 1965) .
- 10. F. M. Bordelon, " Calculation of Flow Coastdown After Loss of Coolant Pump (PHOENIX Code)," WCAP-7969, l Reactor
, Westinghouse Electric Corporation (September 1972) . l l
- 11. T. W. T. Burnett, C. J. McIntyre, J. C. Buker, R. P. Rose, "LOFTRAN Code Description," WCAP-7907, Westinghouse Electric
. Corporation (June 1972) .
14.2-51
BVPS-1-UPDATED FSAR Rov. 8 (1/90) References to Section 14.2 (cont'D) ,. 12. S. Altomare and R. F. Barry, "The TURTLE 24.0 Diffusion Depletion Code," WCAP-7758, Westinghouse Electric Corporation (September 1971).
- 13. R. F. Barry, LEOPARD -
A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094." WCAP-3269-26, Westinghouse Electric Corporation (September 1963).
- 14. " RADIOISOTOPE, A Computer Program for Calculating Residual '
Activities in a Closed System After One or More Decay Periods," RP-1, Stone & Webster Engineering Corporation (November 1972).
- 15. " ACTIVITY A Computer Program for Calculating Fission Product Activity in Fuel, Coolant, and Selected Tanks for a Nuclear 1 Power Plant," RP-3, Stone & Webster Engineering Corporation (January 1973).
- 16. "IONEXCHANGER, A Computer Program for Determining Gamma Activities in Ion Exchangers or Tanks as a Function of Time for Constant Feed Activity," RP-2, Stone & Webster
. Engineering Corporation (December 1972).
- 17. Baldwin, M.S., Merrian, M.M., Schankle, H.S., and van De Walle, D.J., "An Evaluation of Loss of Flow Accidents Caused
- by Power Systam Frequency Transiants in Westinghouse PWRs,"
WCAP-8424, Revision 1, June 1975.
- 18. DLC Calculation ERS-SFL--89-021, Safety Analysis of the Dose '
Consequences of a Locked Rotor Accident at BVPS-1 with 18% Fuel Failure -- EAB, LPZ, Control Room. l 1
)
14.2-52
1 i l BVPS-1-UPDATID FSAR Rev. 8(1/90) TABLE 14.2-4a
SUMMARY
OF RESULTS FOR TACKID ROTOR TRANSIENTS
^
-3 Loops operating 3 Loops-operating (with offsite never) (without offsite rever) ,
Maximum Reactor Coolant -2'*7 RG70 :::: - Q 73g ! Pressure (psia) Maximum Cladding Temp-erature at Core Hot Spot, ( r) un l8fl me/g93 - Zirc-Water Reaction at Core Hot Spot (weight %) +.-NH> 0.9 86 m o,450 e a D .o ; 1 of 2
- - ~ -
i BVPS-1' UPDATED FSAR Rov.'B (1/3 ! TABLE 14.2-da (CONT'D)
SUMMARY
OF RESULTS FOR LOCKED ROTOR TRANSIENTS TIME SEOUENCE OF EVENTS i
~
Accident Ev_ Rat Time (see) l Locked. Rotor ! 3 Loops operating i (with offsite power) I 1 RCP Rotor Locks 0.O Low Flow Trip Point Reached 4 J36- 0,04 Rods Begin to Drop L %l.Q4 , Maximum RCS Pressure Occurs e 3.+ ! Peak Clad Temperature occurs M 3.g Locked Rotor 3 Loops operating (without offsite power) ; 1 RCP Rotor Incks 0.0 l Iow Flow Trip Point Reached- e.;' O.Og # 1 Rods Begin to Drop J. 0 0 - l. 04 ~ : s RCPs Iose Power,
- u. ;,; /,of. l Coassdown Begins Maximum RCS Pressure occurs
- 3. 6 i Peak Clad Temperature occurs M37 o i
m ; i 2 of'2
[REv.B n/:E; 1.4 _ 12 '
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- a. z
$ E5 .50 o !
a5 9ca f g 0.00" b . .50- . 0 1 2 3 4 5 6 7 8 9 10 3' TIME (SECONDS) u FIGURE 14.2-15 ,
' FLOW TRANSIENTS FOR THREE LOOPS IN OPERATION, ONE LOCKED ROTOR BEAVER VALLEY POWER STATION UNIT 1
- UPDATED FINAL SAFETY ANALYSIS REPORT -
\
_ 12 5
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(0.4) 0 2 4 6 8 10 TIME (SEC) FIGURE 14.2-15
~
REACTOR VESSEL AND FAULTED LOOP FLOW TRANSIENTS FOR THREE LOOP OPERATION - ONE LOCKED ROTOR BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT
LRE/.3<i.;dl nue f : f i r/o arrstT - -" ases <- ,,,, t 0FFslTE ....
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Y PLc6 ON \, Q tasyT PMG twe < - o ' m, useo , i 5 i a s . -s ; , ; , ,, , ilWE (SECC*cS) j FIGURE 14.2-16 l 1 REACTOR COOLANT SYSTEM PRESSURE TRANSIENT! FOR THREE LOOPS IN OPERATION ONE LOCKED ROTOR BEAVER VALLEY. POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT
12
= -
$ 3 3 '
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$ 06 - i b
2 04 - 5 h 02
=
i 0 O 2 4 6 8 10 , TIME (SEC)
, 2.900 2.800 -
utnour arrstre powen _ 2.700 k - g 2.600 g 2.500 a - *TworrsntPowem g 2.400 S 2.300 - 2.200 - 2,100 O 2 4 6 8 10 i I TlW (SEC) FIGURE 14.2-16 NUCLEAR POWER AND REACTOR COOLANT SYSTEM PRESSURE FOR THREE LOOP OPERATION - ONE LOCKED ROTOR BEAVER VALLEY POWER STATION UNIT 1 ! UPDATED FINAL SAFETY ANALYSIS REPORT
l REY.3il/:* 1.4 t- <- REPLACE WITi+ m 51 O' 0
/ 4075 OA) 0 .
x
- y .< 0 .
.20 <
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g g 3 4 5 6 f 8 5 50 TIME (SECONDS) 3 FIGURE 14.2-17 NUCLEAR POWER TRANSIENTS, AVERAGE AND ~ HOT CHANNEL HEAT FLUX TRANSIENTS FOR THREE LOOPS IN OPERATION, ONE LOCKED ROTOR BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT
12 s 3 L-s . d d f_ 0 0 5 d 06 - E 04 - 02 - i a g 0 0 2 4 6 8 M TIME (SEC) s _ 12 Y g, <_ b d 08 - 1 e. 06 - b 6 I O4 - d Ig 02 - O 2 4 s a 10 TIME (SEC) s FIGURE 14.2-17 o AVERAGE CHANNEL AND HOT CHANNEL HEAT FLUX TRANSIENTS FOR THREE LOOP OPERATION - ONE LOCKED ROTOR BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT l
i i 3500 j REY. B n/si": 3000 ,, i e 0- gEPLAB WcfR E ' 2500< . ~'
, @g <
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=..,-
h1250, , 5S
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\
-0 1 2 3 4
5 6 7 8 9 10 TIME (SECONOS) FIGURE 14.2-18 MAXIMUM CLAD AND FUEL . CENTERLINE TEMPERATURES AT HOT SPOT FOR THREE LOOPS IN OPERATION , ONE LOCKED ROTOR BEAVER VALLEY POWER STAT [ON UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT-
t C 3 500 - h . wm.out errsat powa n
- L ;
5 3 000 -
/
/
w S { 2.500 g 3 / s . ..r...... E - y 2.000 d 2
'5 0 2 4 e e to TINE (SEC) $ 2.000 a
y [8 g - g 1.600 j .m.ouw,.m ,o . / g 1.co g m or,.n.,o... [ g 1.200 - x w g 1,000 o 800 - 0 2 4 6 8 10 TIME (SEC) i FIGURE 14.2-18'
# MAXIMUM CLAD AND FUEL CENTERLINE TEMPERATURES AT HOT SPOT FOR THREE LOOP OPERATION - ONE LOCKED ROTOR BEAVER VALLEY POWER STATION UNIT 1 UPDATED FINAL SAFETY ANALYSIS REPORT k
INSERT A The analysis is performed to bound operation with steam generator tube plugging levels up to y 30rt with a maximum loop flow asymmetn of Scr i INSERT B The initial conditions assumed in the analysis employ an RCS average temperature uncenainty of +4.5 "F and an RCS pressure uncenainty of -12 psia. INSERT C (Plant characteristics and initial conditions are further discussed in Section 14.D.2) INSERT D Hargrove, H. G., "FACTRAN - A Fortran IV Code for Thermal Transients in a CO2 Fuel Rod," WCAP-7908-A, December 1989. INSERT E The initial conditions assumed in the Locked Rotor DNB analysis employ an RCS average j temperature uncertainty of +4.5 "F and an RCS pressure uncertainty of -42 psia. 4 INSERT F unless otherwise stated in the sections describing specific accidents
?
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