ML20135E688
| ML20135E688 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 02/29/1996 |
| From: | Hoang H GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19310D739 | List: |
| References | |
| GE-NE-B13-01760, GE-NE-B13-01760-R02, GE-NE-B13-1760, GE-NE-B13-1760-R2, NUDOCS 9612110428 | |
| Download: ML20135E688 (55) | |
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4 GE-NE-B13-01760
' Revision 2 ClassI February 1996 SAFETY REVIEW FOR LASALLE COUNTY STATION UNITS 1 AND 2 SAFETY / RELIEF VALVES REDUCTION AND SETPOINT TOLERANCE RELAXATION ANALYSES I
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H. X. Hoang V Project Manager l
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GE-NE-B13-01760, Rev. 2 ;
l IMPORTANT NOTICE REGARDING l CONTENTS OF THIS REPORT PLEASE READ CAREFULLY The only undertakings of the General Electric Company (GE) respecting information in this !'
document are contained in the contract between Commonwealth Edison Company (Comed) and
>l GE, as identified in the purchase order for this report and nothing contained in this document !
shall be construed as changing the contract. The use of this information by anyone other than i Comed or for any purpose other than that for which it is intended, is not authorized; and with i 1
respect to any unauthorized use, GE makes no representation or warranty, and assumes no !
liability as to the completeness, accuracy, or usefulness of the information contained in this '
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GE-NE-B13-01760, Rev. 2 t
TABLE OF CONTENTS U
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SUMMARY
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1.0 INTRODUCTION
1-1 l 1.1 Purpose 1-1
!. 1.2 Background 1-2 1.3 Present Performance Requirements 1-2 1.4 Proposed Performance Requirements Change 1-4 2.0 ANALYSIS APPROACH 2-1 3.0 VESSEL OVERPRESSURE ANALYSIS 3 3.1 Overpressure Analysis Assumptions 3-1 3.2 Overpressure Analysis Results 3-1 4.0 HIGH PRESSURE SYSTEM PERFORMANCE 4-1 4.1 High Pressure Core Spray System Evaluation 4-1 j 4.2 Reactor Core Isolation Cooling System Evaluation 4-4 4.3 , Standby Liquid Control System Evaluation 4-10 a 5.0 CONTAINMENT DYNAMIC LOADS 5-1 ,
, 5.1 LOCA Containment Response 5-1 l 5.2 Safety / Relief Valve Dynamic Loads 5-2 1 5.3 Conclusion 5-6 I
. 6.0 ATWS MITIGATION CAPABILITY 6-1 7.0 SRV AVAILABILITY 7-1 i
8.0 CONCLUSION
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9.0 REFERENCES
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GE-NE-B13-01760, Rev. 2 c
TABLES .
Table . Title Enss 1-1 Comparison of Present to Proposed Performance 1-6 Requirements '
i 3-1 SRV Safety Mode Configuration for MSIV Flux 3-3 Scram Analysis
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3-2 MSIV Closure Event Analysis Results 3-4 l
.4-1 HPCS System Performance Comparison 4-14 4-2 HPCS Pump Head Design Requirements 4-15 4-3 RCIC System Performance Comparison :
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6-1 SRV Relief Mode Configuration for MSIV 6-3 No Scram Analysis 6-2 Initial Operating Condition for ATWS Analysis- 6-5 6-3 Equipment Perfoitnance Characteristics for ATWS Analysis 6-6
. 6-4 MSIV Closure (No Scram) Transient Responses 6-7 6-5 Sequence ofEvents for MSIV Closure No Scram 6-8 l
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l ILLUSTRATIONS Figure litic P_agt 1
l' 3-1 MSIV Closure Flux Scram,102P/105F Nominal +3%,8 SRVs OOS 3-5 6-1 MSIV Closure No Scram Event,102P/87F, 5 SRVs OOS 6-9 1h l
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GE-NE-B13-01760, Rev. 2
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SUMMARY
This report documents the analyses performed for LaSalle County Station (LSCS) Unit 1 !
and 2 in suppon of Commonwealth Edison's (Comed) effort to reduce the number of ,
s Safety / Relief Valves (SRVs) currently installed at the LSCS units. In addition, the analyses also l provide the technicaljustifications to support the relaxation of the SRV safety mode setpoint tolerance from the current + 1% value to + 3%. The setpoint tolerance relaxation ponion of this report was previously submitted for NRC review by Comed (and subsequently approyed) for !
l implementation during L2R06 (Unit 2) and LIR07 (Unit 1). These portions of the report are '
retained in this revision for historical purposes, while other portions denhng with SRV removal have been enhaneM to address some of the safety concerns associated with this proposed change.
This evaluation will be used as part of or as reference by Comed in its licensing change submittal. j 1
The analyses results show that along with the setpoint tolerance relaxation to +3%, up to ,
4 SRVs can be ehminated from the current SRV configuration at the LSCS units without ;!l adversely impacting the safety ofplant operation. In addition, the design basis assumption for one !)
SRV out-of-service may be retained. '
I The limiting transient event for vessel overpressure protection was re-analyzed for LSCS !
Unit 2 Cycle 7 at the +3% safety mode valves openmg setpoints in conjunction with a reduction in :
number of SRVs. The results show that with the tolerance setpoint relaxation and reduction in the number of operable SRVs, the maxunum vessel pressure still remains within the ASME Upset Code limit of1375 psig.
The containment LOCA and the suopression pool boundary loads response, the Anticipated Transient Without Scram (ATWS) and the high pressure make-up system performance were also evaluated to justify operation with the increase in valve setpoint opening and reduced number of valves Results of the evaluation reported herein show that there is no impact on those areas.
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2 GE-NE-B13-01760, Rev. 2
1.0 INTRODUCTION
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1,1 PURPOSE The purpose of this report is to present the results of an evaluation of the LaSalle County Station (LSCS) Unit 1 and 2 Safety / Relief Valves (SRVs) performance requirements. With the current excess steam relief capacity at the LSCS units, the total number of'SRVs can be reduced and yet the remaining configuration would still achieve the design basis requirement to support
, safe plant operation. In addition, the SRVs safety mode opemng setpoint tolerance are relaxed from +1%/-3% to 3% to nuninuze the impact on plant resources due to SRV opening setpoint l drift, and yet maintaming high SRV reliability. Commonwealth Edison (Comed) has requested .
that these SRV performance changes be evaluated to suppon the following pressure relief system '
performance requirements:
1 (1) Relaxation of the LSCS surveillance requirement tolerance from current +1% to +3%
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for the SRVs opemng setpoint in the safety mode. There is no change to the current l performance requirements for the SRVs openmg setpoint in the relief mode.
l (2) Justification for continuous plant operation with a reduction in the current number of SRVs. ~ ,
The current performance requirements for the LSCS SRVs are discussed in Section 1.3.
Each of the present performance requirements peninent to this analysis, as well as the associated limitation and the remedial actions for exceeding the limit, are identified. Section 1.4 discusses l .
the proposed performance requirement changes, the associated limits and the analyses required to suppon each proposed change. A comparison of the present and proposed performance requirements is shown in Table 1-1.
The analysis approach and the listing of the type of analyses performed to suppon the l proposed changes are described in Section 2.0.
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GE-NE-B13-01760, Rev. 2
1.2 BACKGROUND
i The nuclear pressure relief system at the LSCS units consists cf Crosby dual mede SRVs located on the main steam lines between the reactor vessel and the first isolation valve within the
. drywell. The SRVs provide three main protection functions:
l (1) Overoressure relief ooeration. The SRVs open automatically to limit the vessel pressure excursion during a postulated pressurization transient event.
l (2) Overoressure safety function (sorinn safety model. The SRVs, functioning in the self-actuated safety mode, open to prevent the reactor vessel overpressunzation.
i (3) Deoressurization ooeration. The Automatic Depressunzation System (ADS) function is performed by selected SRVs and these valves open automatically as part of the . Emergency i Core Cooling System (ECCS) for events involving small breaks in the reactor vessel l
l process barrier.
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1.3 PRESENT PERFORMANCE REQUIREMENTS ,
1.3.1 SRV_Setooint Tolerance From Reference 1, the current SRV configuration and nominal opening setpoint for LSCS is as follows:
ReliefMode, psig Safety Mode, psig Number of SRVs Nominal As Analyzed Nominal As Annivzed 2 1076. 1091. 1150. 1162.
I 4 1086. 1101. 1175. 1187.
4 1096 1111 115 1197.
4 1106. l'121. 4l'5. 1207. !
4 1116. 1131. .405. 1217.
, The margin for the relief mode opening setpoint between the nominal trip and the
) analytical limit for LSCS is based on a pressure switch setpoint error of +/-15 psi.
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1 GE-NE-B13-01760, Rev. 2 A narrow +1% tolerance band on the safety mode opening setpoint of the SRVs stems from an acceptance criterion defmed by the American Society of Mechanical Engineers (ASME) for vessel overpressure protection- Section 3/4 4.2 of the current Technical Spe:ificctions for LSCS states that the allowable opening setpoint errors for each SRV in the safety mode shall be
+1%.
l The ASME has since revised the criterion for demonstrating valve operational readmess from 1% to 3% (Reference 2) within the plant design basis. The 1% tolerance applies to several limitations which have to be addressed if this tolerance is exceeded. These limitations are as follows:
l (1) The LSCS Technical Specification 3/4.4.2 delineates that the SRVs in safety mode are operable within +1% of the nominal setpoint.
(2) Licensing basis analyses for vessel overpressunzation have been performed assummg the l
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valves opening pressures are +1% above the nominal setpo'mts. If the SRV safety mode opening pressures are greater than 1% above the nominal setpoint, then the plant could potentially operate in an unanalyzed condition. Such a condition warrants a review for a l
Licensee Event Report (LER) and a safety evaluation.
! (3) Valve refurbishment and removal of additional valves from the plant for testmg are j necessary if valve opemng pressures are demonstrated to be beyond the limiting condition for operation 3/4.4.2 (+ 1%/-3% of the nominal SRV safety mode settmgs).
(4) If surveillance testing demonstrates that the safety mode openmg pressures are beyond l +1% of the nommal setpoint, setpoint adjustment to the +1% tolerance is required prior to returning the valves to service.
Consequently, valve opening setpoint drift to > +1% above the nominal setpoint causes each of the above remedial actions to be taken, thereby increasing valve surveillance testing costs, and the number of reportable events, thus increasing activities that use additiorial utility resources.
Although the +1% tolerance is specified in the LSCS Technical Specifications and has been used 1
in plant safety evaluations, it does not represent the limiting setpoint required to ensure plant
, safety. Several BWRs have experienced 'SRV setpoint drift in excess of the Technical Specification limitations. Each time, safety evaluations were performed on a cycle specific basis 1-3
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GE-NE-B13-01760, Rev. 2 demonstrating that setpoint drift did not compromise plant safety. The consequences ofvalve opening setpoint drift can be minimiraA by increasing the setpoint tolerance assumed in licensing analyses and resultant plant operating limits.
1.3.2 SRV Reduction l The current reload licensmg basis for LSCS assumes one SRV declared OOS for mmimum critical power ratio (MCPR) and vessel overpressure protection calculations.
1.4 PROPOSED PERFORMANCE REQUIREMENT CHANGE l '
This section discusses the effect of each set of the proposed performance requirement changes on LSCS and the analyses necessary to support the changes. The present and proposed L SRV performance requirements are shown in Table 1-1.
l 1.4.1 SRV Safety Mode Tolerance Setooint Palavation The ASME has expanded the acceptance criterion for SRV performance testing from +1%
to +3% per Reference 2. Consequently, as long as the maximum valve opening pressure remains below the nommal +3'/ range, the plant is still wnhin analyzed conditions and the valves are considered capable ofperforming their relieffunction.
The acceptance criterion defines the range of expected in-service performance of a valve.
Beyond this criterion,. valve refurbishment is required and additional valves must be removed from the plant for testing The increased tolerance on the acceptance criterion potentially reduces the number of valves that will exceed the in-service performance testing requirements, thus reducing the cost of valve surveillance testing. '
Prior to placing new or refurbished valves in service, the valves setpoints are adjusted to j within +1% of the nommal nettinge Inet=11=+ ion of the valves within a +1% tolerance ensures that there is margin to the +3% in-service testing criterion for openmg pressure. Thus', valve integrity and the benefits of the increased surveillance requirement tolerance are maintained from cycle to cycle.
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They include vessel overpressure protection, ECCS/LOCA performance, fuel thermal limits, containment loads and high pressure system performance (High Pressure Coolant System, Reactor Core Isolation !
Cooling System, and Standby Liquid Control System). l l
l GE work scope consists of the tasks identified above, with the exception of the ECCS/LOCA performance, fuel thermal limits impact and mam steam pipmg loads which are
., Comed's responsibility and thus are not part of this report.
1.4.2 SRV(s) Reduction 4 To take advantage of the current over-designed steam relief capacity at LSCS, it is >
l proposed to reduce the number of SRVs from the current eighteen-valve configuration to a ,
smaller number based on the safety analyses results. For LSCS, the proposed changes include justifying continued plant operation with less than eighteen SRVs available. However, the SRVs available for potential elimmation cannot be part of those required to perform the Automatic Depressurization System (ADS) and the Low-Low-Set (LLS) Logic function. !
l The potential consequences from the SRV reduction will be evaluated. The final number l
of SRVs available for permanent removal will be based on the maximum number of SRVs required to comply to the reactor vessel overpressure protection (during normal transients as well
! as ATWS events), ECCS/LOCA performance, fuel thermal limits, containment and main steam piping loads, high pressure system performance and Emergency Procedures Guidelines (EPGs).
GE work scope consists of the tasks identified above, with the exception of the ECCS/LOCA performance, fuel thermal limits impact, main steam piping loads and EPGs which are Comed's responsibility and thus are not part of this report.
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Table 1-1 i COMPARISON OF PRESENT TO PROPOSED !
PERFORMANCE REQUIREMENTS !,
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,, Performance Reauirement i Prasant I imil New T imit i i
! 1. Opening pressure (relief mode) up to which the 15 psi 15 psi SRVs are capable of performing their intended
- i p function (operable).
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- 2. Opening pressure (safety mode)up to which the + 1%/-3% 3%
the SRVs are capable ofperforming their intended l
i function (operable), Technical Specification 3/4.4.2
- 3. Opening pressure up to which licensing basis + 1.% +3%.
analyses have been perfonned. !
- 4. Tolerance beyond which valve refurbishment + 1%/-3% i' 13% ;;
l and additional valve testing is required as demonstrated by surveillance testing,
- 5. Tolerance on the as-left SRV settmg prior to the 1% i1% ;
valve being returned to service. l
. Note for SRV Radar + ion:
l Although the analyses included in this document support up to six SRVs as available for reduction, only five SRVs can be removed (with one valve assumed out-of-service) and the two SRVs with the lowest opening setpoint must not be part of this group of six to comply with HPCS performance criteria (section 4.1.3).
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GE-NE-B13-01760, Rev. 2 i i' 2.0 ANALYSIS APPROACH This section identifies the areas which nuy be affected by the proposed SRV performance !
l requirement changes shown in Table 1-1. The following safety and regulatory concerns are l
identified as bemg potentially affected as a result of the SRV safety mode opening setpoint tclerance increase to +3% and operation with a reduction in the number of SRVs:
- . 1. Vessel overpressurization.
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Thermal limits during anticipated operational occurrences (AOOs).
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- 3. Emergency Core Cooling System (ECCS) performance durmg postulated LOCA. '
- 4. Anticipated Transients Without Scram (ATWS). I I
- 5. High pressure system performance '
l 6. Containment LOCA responses and suppression pool boundary dynamic ~ loads. i l 7. Main steam piping loads, includmg loads on attached SRV discharge lines.
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- 8. Emergency Procedure Guidelines.
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- 9. SRV availability.
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! GE's work scope includes item 1, 4, 5, 6 and 9 and Comed or its Architect-Engmeer (Sargent and Lund ) is. responsible for the remaining tasks Although GE does not have the updated main steam piping analyses of the LSCS units as performed by Sargent and Lundy, the l GE work scope does not require the availability of this information.
For the scope of work performed by GE and documented in this report, the SRV safety mode tolerance setpoint is increased from +1%/-3% to 3% in conjunction with a reduction in the number of SRVs. Due to the different applicable criteria applicable, the number of SRVs
, available for ehmination will be specific to each tasks and the smallest value will be recommended for subsequent implementation.
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) ,. 3.0 VESSEL OVERPRESSURE ANALYSIS The AShE Code requires peal: vessel pressures to be less than the upset transient limb of 1375 psig dudng transient events. The limiting ovegressure event for the LSCS units is the Main !
Steam Isolation Valve (MSIV) closure with flux scram event (Reference 3). The reactor is shutdown by the backup indirect high neutron flux scram due to vessel pressurization and subsequent collapse of voids.
- The greatest challenge to the ASME Upset code limit is provided by assummg that all the l SRV safety mode setpoint have drifted upward to +3% above the nommal trip setpoint, l coincident with a reduction in the number of SRVs.
l 3.1 OVERPRESSURE ANALYSIS ASSUMPTIONS The following assumptions and initial conditions were used in analyzmg the MSIV closure with flux scram for LSCS Unit 2:
1 (1) Initial core thermal power at 102% ofrated.
(2) Initial core flow at 105% ofrafed (3) .
(4) Reduction in the number of SRVs such that the ASME overpressunzation criteria (peak i vessel pressure less than 1375 psig) is maintained.
(5) Credit taken for the available SRVs in the safety mode. .
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The SRV safety mode opening pressures analyzed were +3% above the nommal trip setpoint (see Table 3-1).
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3.2 OVERPRESSURE ANALYSIS RESULTS The overpressure analysis results are provided to serve as an indicator for feasibility of application to Units I and 2. However, this analysis must be reveri6ed each cycle through the s
normal reload licensing process. The reactor response with the SRV safety mode opening setpoint at +3% above the nominal is shown in Figure 3-1. The event is initiated as the MSIVs begin to
! close.
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With only 10 SRVs available out of a total of.18, the calculated pc4 vessel pressure at the bottom of the reacte msselis 1341 psig, thus providing significant margin to the ASME Upset code limit of 1375 psig. Since the current reload licensing basis for the LSCS units is to assume one SRV OOS, the net number of SRVs available for elimmation, based on the ASME overpressure upset criteria, would be seven valves. The SRVs OOS calculations assumed that the first 7 valves, starting with the lowest opening setpoint and going upward, were removed, and the next lowest valve was also not credited, but left as a potential SRV OOS. Section 6.0 "ATWS Mitigation Capability" provides further discussion on calculations relating to the maximum number of SRVs available for ehmination.
Table 3-2 shows the resultant peak vessel pressures for the MSIV closure flux scram event analyzed and Figure 3-1 shows the time histories of key parameters during this transient event.
The Unit 2 Cycle 7 reload licensing analyses results (Reference 3), with 17 out of 18 SRVs available and with a setpoint tolerance of-3%/+1%, are also included in Table 3-2 for comparison l Purposes.
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! i Table 3-1 SRV SAFETY MODE CONFIGURATION FOR MSIV FLUX SCRAM ANALYSIS ;
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!7 SRV Safety Mode Number of SRV* Nommal Setnoint. nsig As Analyzed. nsig l
l- 2 1150 1185 4 1175 1210 4 1185 1221 I i
4 1195 1231 ',
4 1205 1241 +
i Lowest 8 valves not credited.
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l Table 3-2 MSIV CLOSURE FLUX SCRAM EVENT ANALYSIS RESULTS l
l Peak Peak
. Peak Peak Steamime Ves.sel l SRV Neutron Flux Heat Flux Pressure Bottom Pressure Power / Flow Conhration (% NBR) (% NBR) pig Esig 102/105m Nom. + 1%, 486 132 1240 1275 17 SRVs l*
102/105 Nom + 3% 486 132 1316 1341 10 SRVs i
Note: (1) LSCS Unit 2 Cycle 7 reload analysis (Reference 3), with -3%/+1% setpoint tolerance .
range.
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I Figu p 3-1 MSIV Closure Hun Scrans Event,102P/195F, Noniinal + 3%,
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l GE-NE-B13-01760, Rev. 2 1
, 4.0 HIGH PRESSURE SYSTEM PERFORMANCE 1
The purpose of this section is to evaluate the impact of the SRV safety mode opening
! setpoint tolerance change and the SRV reduction on the high pressure make-up system ;
performance at LSCS. The followmg system are meluded in the evaluation: ;
l l - High Pressure Core Spray (HPCS) 1
- Reactor Core Isolation Cooling (RCIC)
, - Standby Liquid Control System (SLCS) l 4.1 - HIGH PRESSURE CORE SPRAY SYSTEM EVALUATION ll' The most significant impact of the SRV setpoint tolerance relaxation and SRV reduction l program on the HPCS system is the resulting higher reactor pressure due to the increase in the SRV upper analytical opening setpoint. For LSCS, the HPCS system was originally designed to provide injection into the reactor pressure vessel up to at least 1% above the lowest safety ,
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- setpoint of the SRVs, which corresponds to a reactor pressure of I162 psig. With the setpoint
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tolerance relaxation program, the SRV safety setpoint tolerance is being increased from 1% to 3% This change increases the maximum reactor pressure for HPCS system injection by 23 psi, to 11,85 psig. :
4.1.1 System Function and Reauiramants The HPCS system, an ECCS component, is designed to provide sufficient core cooling and prevent excessive fuel cladding temperuure in the event of a LOCA. The HPCS system accomplishes this function by injecting coolant makeup water into the pressure vessel to cool the reactor core when coolant is lost through any design basis break of the nuclear system process barrier. The HPCS also supplies makeup water to the reactor vesselin the event of a transient l
which results in the loss of all feedwater flow or reactor isolation and a failure of the RCIC system. The HPCS system is designed to deliver water to the reactor vessel at a rate equal or greater than 516 gpm, with the reactor vessel pressure 1160 psi above the pressure at the source of suction (suppression pool).
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GE-NE-B13-01760, Rev. 2 e
4.1.2 Inouts and Assumotions
- The following values constitute the present high pressu.e design point for the HPCS i system
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I System Flow Rate = 516 gpm l
Pump Flow Rate = 1156 gpm l Reactor Operating Pressure = 1160 psig 1
The HPCS system changes required by the SRV setpoint tolerance relaxation and SRV ,
l reduction program will be based upon maintaining the same system flow rate and injection time at ,,
l the new maximum system operating pressure. The HPCS system requires that the current setpoint for the lowest group of SRVs must be maintain *A in order for the system to meet its design basis requirements. Table 4-1 lists the parameters used to evaluate the effect ~of the SRV setpoint tolerance relaxation and SRV reduction upon the HPCS system perfonnance. l l
l l 4.1.3 System Evaluation i
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l 4.1.4 Component Evain= tion !
System components were evaluated by comparing the system's current operating and design temperatures and pressures with the expected system operating temperatures and pressures associated with the increased SRV setpoint tolerance. This examination demonstrated that the '
l current operating values as well as the projected operating values are bounded by the cunent (
design. Therefore, the individual system components will be subjected to temperatures and pressures that are within the current design.
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l 4.1.5 Interfacing Systems Evahintion l
Systems interfacing with the HPCS with potential interface changes are identified in this I section. The Primary Containment, Condensate Storage System, Reactor Vessel System, Service I
, Air System, Residual Heat Removal System, Radwaste System and Leak Detection System interface with the HPCS System, but do not have signi6 cant changes to the system interfaces.
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l 4.1.6 Conclusion The HPCS system was found to have the capability to deliver the required flow of 516 gpm at the increased reactor pressure resulting from relaxation of the SRV setpoint tolerances. l l The higher reactor pressure with the SRV tolerance relaxation program does not impact the design of those system. components directly impacted by the increased reactor pressure, including the valves, because the system was designed to operate at the higher pressures expected dunng l system operation at shutoff head conditions (no flow to the reactor vessel).
However, for the SRV reduction program, the HPCS design performance imposes a restriction on the SRVs selected for potential removal, such that the lowest opening setpoint SRV ;
! group must be maintained. '
l 4.2 REACTOR CORE ISOLATION COOLING SYSTEM EVALUATION The most significant impact of the SRV setpoint tolerance relaxation'and SRV reduction program on the RCIC system is the r=1 ting higher reactor operating pressure due to the increase' in the SRV upper analytical opemng setpoint. For LSCS, the RCIC system is origmally designed ;
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to provide injection into the reactor pressure vessel up to at least 1% above the lowest safety l setpoint (analytical limit) of the SRVs, which corresponds to a reactor pressure of 1162 psig.
l With the SRV setpoint tolerance relaxation, the SRV safety setpoint tolerance is beic; increased from 1% to 3%. This increases the maximum reactor pressure for RCIC system injection by 23 pski to 1185 psig.
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l The RCIC System, classified as a Power Generation System, is designed to maintain the I !
reactor vessel water level above Level 1 in the event of a transient occurrence which r tsults in the loss of all feedwater flow or reactor isolation. The system is also designed to allow fx complete shutdown by maintaining sufficient water inventory until the reactor is depressurized to a level where the shutdown cooling mode of the Residual Heat Removal (RHR) system can be plaud into operation. The RCIC system accomplishes this function by injecting coolant makeup water l
into the reactor pressure vessel with a turbine driven pump.
l.
The system design basis requirement for the RCIC is a developed head of 2890 ft at a j
reactor pressure of 1158 psig (high reactor pressure operating mode). i 4.2.2 Inputs and Assumptions The following values constitute the present high pressure design point for the RCIC system:
]
- System Developed Head = 2890 ft '
Reactor Operatmg Pressure = 1158 psig i Pump Speed = 4530 rpm Pump Shut-OffHead = 1476 psig The RCIC system changes required by the SRV setpo' m t tolerance relaxation and SRV reduction program will be based upon maintaming the same system design requirement @Ly at the new reactor operating pressure. The RCIC system changes will also take into consideration ;
any hmitations on the program imposed by.other systems. The HPCS system requires that the 4-5
GE-NE-B13-01760, Rev. 2 ,l t
'l i
current setpoint for the lowest group of SRVs must be maintained in order for the system to meet its design basis requirements. Consequently the RCIC system changes will be based on changing the SRV setpoint tolerance for the lowest group of SRVs. Table 4-3 lists the parameters u.-ed to l evaluate the effect of the SRV setpoint tolerance and SRV reduction upon RCIC system i performance.
4.2.3 System Evaluation '
l t
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GE-NE-B13-01760, Rev. 2 1
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4.2.4 Comoonent Evm1 nation h
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i-t GE-NE-B13-01760, Rev. 2 l!
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4.2.'5 Interfacine Systems Evaluation Systems interfacing with the RCIC with potential interface changes are identified in this ;
, section. The Primary Containment, Condensate and Condenser, Reactor Water Cleanup, and !
! Radwaste systems interface with the RCIC system, but do not have significant changes to the
[ )
system interfaces.
4-8
GE-NE-B13-01760, Rev. 2 i
t
- 4.2.6 Conclusion The RCIC system was found to have the capability to deliver its design rated flow of 600 gpm at the increased reactor pressure resulting from relaxation of the SRV setpoint tolerances.
This capability was achieved by increasing the turbine / pump maxunum rated operating speed to obtain an increase in the pump developed head while maintaining the original system design margins.
l The RCIC turbine has the capacity to develop the horsepower and speed required by the pump to meet its new discharge pressure requirements while continuing to use the original 4-9
GE-NE-B13-01760, Rev. 2 system design margins. The change in the system design point requires a new pump and turbine rated speed of 4580 rpm. This speed is below the maxunum continuous operating speed speci6ed by the pump and turbine manufacturers. The increased turbine rated speed requires the acceptance of a reduced overspeed trip margin since the maximum trip speed cannot be raised
. above the speci6ed manufacturers limit.
t The steam supply isolation setpoint of 300% of steady state Dow for steam line leak !'
detection will need to be re-evaluated as de6ned in GE SIL 475 (RCIC and HPCI High Steam Flow Analytical Limit) for the 3.8% higher steam flow rates. '
The RCIC System valves that are impacted by the increase in reactor pressure will require ;
re-evaluation for operability at the increased operating pressures. The specified full differential pressure values for the RCIC steam supply and pump discharge valves should be adjusted >
accordingly to reflect the effect of the new SRV setpoint tolerances.
~
The impact of the SRV setpoint relaxation program on the remainder of the system components was determined to be negligible because of the very small increase in operatmg pressure and/or teirweure.
The following modi 6 cations /setpoint changes are required for the RCIC System to perform at the new design point:
- 1. Turbine control system adjusted for a rated speed of 4580 rpm
- 2. Steam supply line isolation differential pressure setpoint re-evaluated j .
- 3. Valve operability confirmed for higher differential pressures 4.3
!I STANDBY LIQUID CONTROL SYSTEM EVALUATION l
The SRV setpoint tolerance relaxation and SRV reduction program does not impact the ;
l performance of the SLCS. The SLCS was originally designed to provide injection into the reactor l pressure vessel from zero pressure up to a maxunum reactor pressure of 1150 psig at the point of-l injection. The performance of the SLCS was conservatively based on the SRV relief setpoint pressure (with 1% setpoint tolerance) for the highest valve group. Since the SRV setpoint i
1 1
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' 4-10 j
_. - - _ ,~ .
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GE-NE-B13-01760, Rev. 2 tolerance relaxation program increases the SRV spring safety setpoint tolerance from 1 to 3%
without impacting the SRV relief function setpoint tolerance, the operation of the SLCS !
will not be impacted. The removal of SRVs under the SRV reduction program will not impact the l performance of the SLCS since the maximum system injection pressure is based on the upper f, analytical pressure for highest valve group. l Since the calculations for maximum pressure at the discharge of the S,LCS pumps were completed by the utility for implementation of ATWS, this report will not include an asseissment
., of SLCS operation.
The ability of the SLCS pump to inject its design flow rate into the reactor vessel is not directly affected by this analysis since there was no change in the reactor pressure for system operation.
i l
l 4.3.1 System Functions and Reauiremente ,
i The Standby Liquid Control System (SLCS) is a reen* reactivity control system capable of shutting down the reactor from rated power condition.to cold shutdown in the postulated condition that all or some of the control rods cannot be inserted. It is a manually operated system that will pump a sodium pentaborate solution into the vessel in order to provide neutron absorption and achieve a subcritical reactor condition.
Since this analysis does not change the reactor power level or shutdown margin
. requirements, it has no impact on the SLCS shutdown capability. The proposed change in SRV setpoint tolerances increases the maximum reactor pressure during injection, thus increasing the pump discharge pressure for injection. I The design criterion for this system is to provide a prescribed boron concentration in solution into the reactor (660 ppm). Technical Speci6 cation hauts are placed on this system to assure adequate reactor shutdown margin. These limits are expressed in terms of acceptable solution volume and concentration operating regions. The operation of a single SLCS pump at a nominal flow rate of 41.2 gpm, meets the boron injection rate requirements for continued decreasing reactivity as the core cools down.
(
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I; GE-NE-B13-01760, Rev. 2 l
1 The maximum reactor pressure at which the SLCS pumps could be called upon to mject sodium pentaborate into the reactor is determmed by the upper analytical pressure for the highest group of SRVs operating in the relief mode. The maximum pressure at the discharge of the SLCS pumps is therefore the SRV setpoint pressure plus the head of water in the reactor and the pump discharge system flow and head losses with the operation of either one or both pumps in operation.
4.3.2 Inouts and Assumotions j The following values constitute the present design of the SLCS :
t i
Pump Nommal Flow Rate = 41.2 gpm (each) i
, Reactor Operating Pressure Range =
0 to 1150 psig '
f Injection Rate (Boron) =
6 to 25 ppm / min Reactor Boron Concentration =
660 ppm Pump reliefvalve nommal setpoint = 1400 psig
! 4.3.3 System Evaluation I y e l .
t
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4-12
GE-NE-B13-01760, Rev. 2 4.3.4 Interfacinn Systems Evaluation Systems interfacing with the SLCS with potential interface changes are identified in this section. The Pnmary Containment, Plant Air system, Deminerahzed Water, and Radwaste systems interface with the SLCS, but do not have significant changes to the system interfaces.
l
[
4.3.5 Conclusions ~ ~,
The SLCS for LSCS was designed to inject the neutron absorber solution at a maximum reactor pressure of 1150 psig measured at the outlet of the control sparger. The results of the evaluation found that SRV setpoint tolerance relaxation and SRV reduction program ~does not impact the system capability to deliver the required flowrate of neutron absorber solution to the reactor pressure vessel at the higher reactor pressures.
The impact of this program on the remainder of the system components was determined to be negligible because the system operating pressures do not change.
No modifications or setpoint changes are required for the SLCS as a result of this program l l
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, 4-13
e
- 4 GE-NE-B13-01760, Rev. 2 I
Table 4-1 HPCS SYSTEM PERFORMANCE COMPARISON s SRV Safety Mode Setpoint Tolerance Nom. 1% Nom. 3%
As -Analyzed SRV Safety Mode Openmg Sctpoint, psig 1162 1185 1187 1210
'd 1197 1221 1207 1231 1217 1241 Reactor Pressure, psig (above suction source) 1160 1185 Required System Injection Rate, gpm 516 516 Minimum Flow Line Rate, gpm 640 640 Total Required Pump Flow Rate, gpm 1156 1156 Required TDH, feet 2908.3 2967.0 I
Pumo Characteristics !
Pump Total Dynamic Head Required, ft 3000 3000 Pump Flow Rate, gpm ^ l156 1156 j
~ Margin ft 91.7 33.0 :
.. ...m,..
Note: For SRVs reduction purposes, the lowest openmg setpoint SRV group (two SRVs at 1076 psig nominal setpoint) must be retained to maintain the required HPCS
! performance.
l i 3
4-14
l GE-NE-B13-01760, Rev. 2 !
, Table 4-2 -
HPCS PUMP HEAD DESIGN REQUIREMENTS l Design:
i l SYSTEM AVAILABLE DESIGN DESIGN ,
SRV AL(+1%) REQD. TDH PUMP TDH MARGIN MARGIN ,
j' Groun (osin) (feet) (feet) (feet) (osie) >i i
i l 1 1160 2908.3 3000 91.7 39.1 Proposal:
SYSTEM AVAILABLE DESIGN DESIGN SRV AL(+3%) REQD. TDH PUMP TDH MARGIN MARGIN i l Group (osie) (feet) (feet) (feet) (osie)
. 1 -1184.5 2965.8 3000 34.2 14.6 2 1201.2 3026.1 3000 -26.1 n/a i 3 1220.6 3050.5 3000 -50.5 . n/a i!
t i
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R' GE-NE-B13-01760, Rev. 2 '
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Table 4-3
)
RCIC SYSTEM PERFORMANCE COMPARISON i
SRV Safety Mode Setpoint Tolerance Nom.i1% Nom. 3%
As -Analyzed SRV Safety Mode Opening Setpoint, psig 1162 1185 1187 1210 1197 1221 !
3 1207 1231 i 1217 1241 System Flow Rate, gpm 600 600 Pumo Characteristics Total Dynamic Head, ft 2890 2960 Pump Flow Rate, gpm 625 625 Shaft Speed, RPM 4530 4580 Brake Horsepower, HP 702 730 !
, Turbine Characteristics Turbine Steam Supply Press., psig 1158 1185 Inlet Pressure (mimmum required), psig 410 420 ,
Steam Flow Rate, Ibm /hr 28,250 29,330 Design Rated Speed, RPM 4530 4580 Nominal Overspeed Trip Speed, RPM 5625 5625 Maximum Overspeed Trip Speed, RPM ,,5740 5740 Overspeed Trip Setpoint Margin, % speed 124.2 122.8
- Suggested speed values l
" Based on rated and nominal trip speeds j Note: For SRVs reduction purposes, the RCIC perfonnance evaluation assumes that the lowest opening setpoint SRV group (two SRVs at 1076 psig nommal setpoint) are retained, based on the HPCS performance requirement.
4-16
.- - .- = - . .-. -- - - - - - . . - . - - . . . .
GE-NE-B13-01760, Rev. 2 5.0 CONTAINMENT DYNAMIC LOADS The Safety Relief Valvc (SRV) safety mode setpaint tolerance relaxation to +3% was assessed for potential impr.ct on the containment hydrodynamic loads. The results of this
- , assessment also considers plant operation with a reduction of up to 5 SRVs (4 SRVs available for removal plus one design basis SRV OOS) out of a total of I8 SRVs currently avadable.
5.1 LOCA CONTAINMENT RESPONSE '
5.1.1 Containment Pressure and Temnerature i
The effect on the peak containment pressure and temperature response and on the peak suppression pool temperature for the respective limiting events was considered. The most limiting event in terms of peak containment temperature response is the design basis accident (DBA) LOCA, a double-ended guillotine break of the steam line. For peak containment pressure and peak suppression pool temperature responses, the limiting DBA LOCA is the recirculation line break. Relaxation of the SRV setpoint tolerance and reduction of 5 SRVs (4 SRVs removed and one SRV assumed OOS) has no effect on both of these events because the vessel depressurizes without any SRV .actuations. Therefore, there is no impact .on the DBA-LOCA peak containment pressure and temperature and on the peak DBA-LOCA suppression pool temperature.
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GE-NE-B13-01760, Rev. 2 i
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Therefore, relaxation of the SRV safety open setpoint tolerance to +3% and elimmation of 5 SRVs (4 SRVs removed and one SRV assumed OOS), will not impact the containment pressure l and temperature response for smaller breaks.
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5.1.2 LOCA Hydrodynamic Load >
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- 5.2 SAFETY / RELIEF VALVE DYNAMIC LOADS The SRV dynamic loads defined for LSCS Unit I and 2 were reviewed to detemune the '
effect of a relaxation of the SRV safety open setpoint tolerance to +3%. The purpose of the 1
, review was to determine if sufEcient conservatism and margins in the LSCS defined SRV loads are available to offset the effects of an increase in the SRV opening pressure of +3% and reduction in the number of SRVs available.
SRVs provide pressure relief during reactor transients. Steam discharged from the SRVs 5-2
. i-GE-NE-B13-01760, Rev. 2 is routed through the SRV discharge lines (SRVDLs) and through the SRVDL quencher into the suppression pool. Actuation of SRVs introduces high pressure steam in the SRVDL which
! quickly pressunzes the SRVDL resulting in the forced expulsion of the waterleg initially in the 1 SRVDL and subsequently the air in the SRVDL. The SRV loads r=*ing from SRV operation I include the reaction and thrust loads acting on the SRVDL and quencher and the air-bubble loads
.which are transmitted to the submerged boundaries and structures. These loads and the basis for these loads as applied to LSCS are summarized in Reference 7. I An increase in the SRV safety open setpoint tolerance to +3% from the current value of 1% will result in an increase in the SRV opening discharge flow rate into the SRV discharge line. !
This in turn results in an increase in the loads associated with SRV openings. Therefore, to support operation with the SRV safety open setpoint tolerance relaxed to +3%, an evaluation of I:
the impact on the SRV loads was performed. i l For the condition where 5 adjacent SRVs are elimina% (4 SRVs removed and one SRV assumed OOS), the potential impact of asymmetric hydrodynamic load distribution in the '
suppression pool during the initial discharge of air in the pool event is addressed as follows. The l LSCS Design Assessment Report (DAR), [ Reference 7], considers several load cases including an l asymmetric load case due to the discharge of 3 adjacent SRVs in the suppression pool. The DAR describes this case as non-realistic due to its low probability but is used as a conservative l .
asymmetric load. This load contains two conservatisms. As discussed in Reference 8, one conservatism is tied to the conservatism in the load t agnitude which includes a 50% multiplier.
l A second conservatism is due to the fact that this load case, as the other load cases, assumes that the bubbles created in the suppression pool are formed simultaneously and in phase. This latter conservatism is considered unlikely.
l
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GE-NE-B13-01760, Rev. 2 t r l I
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l Therefore, the current design load for LaSalle, as desenhd in Reference 7, remains boundmg with reduction of 5 SRVs (4 SRVs removed and one SRV amanmad OOS).
~
.l The SRV loadi evaluation was divided into two parts: 1) the loads on the SRVDL and quencher and, 2) the loads on the submerged suppression pool boundary and on the submerge structures in the suppression pool. .
5.2.1 SRVDL and Onanchar Imde This task is not part of the GE scope of work and the results will be provided by Comed '
or Sargent and Lundy.
5.2.2. Submerned Pool Bound =rv and Suhmarned Struchure Imde l ,
The loads on the submerged boundary and on submerged structures are based on thej pe i
bubble pressures detwa44 with the generic ==*M described in RJewss 7 and 8. The conservatism in the generic methods were reviewed to address load increases due to the setp i tolerance relaxation. l i
i 5-4 .
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! l I' GE-NE-B13-01760, Rev. 2
- l, L
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l
- l. SubmereedPoolBoundarv Load LSCS uses the KWU T-Quencher at the end of the SRV discharge line, therefore the
{
l design pool boundary loads for the LSCS units are based on the KWU T-Quencher methodology.
l According to Reference 7, the LSCS T-Quencher load uses the KWU T-Quencher methodology l which is also described in Reference 8 and is identified as the " Alternative Methodology" for
! i defining the T-Quencher design load. According to Reference 8, the basis for the " Alternative l
Methooology" are the results ofX-Quencher tests conducted by KWU I
1
)
l
\
~
l Therefore, the LSCS SRV load definition for the pressure
'on the pool boundary is not impacted by the increase in the SRV open setpoint tolerance to +3%.
i SubmergedStructure Lavh. \
s 5-5 i
l GE-NE-B13-01760, Rev. 2 Therefore, the submerged structure SRV load definition of Reference 7, is not impacted by an increase in the SRV setpoint open pressure tolerance to +3%.
l l
5.3 CONCLUSION
Due to the significant conservatism and margins avadable in the SRV loads, an increase in the LSCS SRV safety opening setpoint tolerance to +3% and the reduction of up to 6 SRVs (5 SRVs removed and one SRV assumed OOS) will not adversely impact the current design basis SRV hydrodynamic loads analyses results.
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GE-NE-B13-01760, Rev. 2 c
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! ,l l- 6.0 ATWS MITIGATION CAPABILITY l
The potential impact of the SRV reduction program on the LSCS Anticipated Transient !
Without Scram (ATWS) performance is in the compliance to vessel overpressure criteria of 1500 i
psig (Emergency Condition) due to reduced relief capacity The SRV setpoint tolerance I h'
i relaxation of +3%, applicable to the safety mode setpoint, does not affect the ATWS analysis !
results because the SRVs are credited here only for their function in the relief mode.
t l5 The limiting event for this ATWS condition is the main steam isolation valve closure (MSIVC) transient. For such an event, it is conservatively assumed that the reactor scram does not take place on any reactor protection system signals. Thus, the eventual shutdown of the plant 3 for this postulated event is by the use of the Standby Liquid Control System (SLCS). The initial l
reduction in power occurs by the use of the ATWS high dome pressure recirculation' pump trip (RPT) signal. For this analysis, the event was termin=*ad after the ATWS RPT function is actuated by its upper analytical limit and following the actuation of the SRVs causing the dome i pressure to start decreasing l The SRV setpoint tolerance relaxation does not apply to the ATWS scenario because this event takes credit only for the SRV relief mode actuation. Therefore, the following r2sumptions
. were used to study the effect of SRV reduction on this limiting ATWS event:
i
- 1. The reactor is operating at 100% power /87% flow. This is the limiting state point I
for ATWS analysis consideration because at this maximum power / minimum core flow condition, the effectiveness of the recirculation pump trip is less pronounced than at higher core flows (rated or increased core flow conditions). ;
- 2. The MSIVs are assumed to close within 4 seconds.
- 3. The SRV relief mode statistical upper analytical limits are calculated based on the i nominal trip opening setpo'mt conservatively 'mcreased by 30 psi over the current
, nominalvalues (see Table 6-1). l
) 4. The number of SRVs is reduced such that the 1500 psig criteria is still met.
b 1
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GE-NE-B13-01760, Rev. 2 ii I
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l The initial plant op=ing conditions and the equipment performance characteristics assumed in the analysis are shown in Table 6-2 and 6-3, respectively. For this MSIV Closure with No Scram event analyzed with 12 SRVs available (out of the total number of 18 SRVs), the peak reactor vessel bottom pressure was calculated to be 1378 psig, which is less than the ASME l service level C (Esrpocy) value of1500 psig. The available margin to the limit is reserved for l : potential future power uprates. In addition to one design basis SRV OO!!, the mimber of SRVs
- - available for reduction, asjusti6ed by the ATWS analyses results, is 5 vahes. The SRVs OOS are assumed to be from the lowest opening setpoint valve and going upward. The transient peak values are 9 =marized in Table 6-4 and key parameters time histories are presented in Figure 6-1.
The sequence of events for this transient is stumterM in Table 6-5.
The peak pressure resuhs of the ATWS event are generally cycle independent. The I
ASME ovepressunzation event is performed each cycle to ensure vessel overpressunzation does
- not exceed 1375 psig (see section 3.0 of this report). This requirement also ensures that ATWS i
6-2
_ . ~ , _ _ _ . -
. . . . . . . 1 i
GE-NE-B13-01760, Rev. 2 criteria continue to be met each cycle, as plant specific results show the ASME overpressure peak pressure value to contain less margin to the required limit than the ATWS results. The peak responses of fuel clad /:ng temperature and suppression pool temperature are not significantly affected by the reducw in the number of SRVs. These parameters are significantly impacted only if changes such as fuel type, power uprate or changes in the maximum licensed rod line assumed for the ATWS scenario are made in the initial core thermal power conditions. Therefore, ATWS need only be reevaluated if any such changes are considered.
i
[TEs analysis was perfortned for LSCS Unit 1, Cycle 8 core and plant configuration and the results are applicable to Unit 2 also].
l 1
Therefore, it is concluded that the ahminatian of 6 SRVs (5 valves removed in addition to !
one SRV assumed OOS) from the current 18-valve enaA= Mon does not adversely impact the i vessel overpressurization criteria for the hmiting ATWS event.
l 4
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6-3 i
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GE-NE-B13-01760, Rev. 2 l
Table 6-1 SRV RFLTEF MODE CONFIGURATION FOR MSIV CLOSURE NO SCRAM ANALYSIS l s Nominal Trip Nonunal + 30 psi As used in analysis
- Setooint. psig psig p.Eia l 1076 1106 1113 "
1076 1106 1130**
1086 1116 1116**
i 1086 1116 1146**
1086 1116 1127**
1086 1116 1136 "
1096 1126 1126 1096 1126 1157 1096 1126 1137 10 % 1126 1146 1106 1136 1137 1106 1136 1166 1106 1136 1147
, 1106 1136 1156 1116 1146 1146 1116 1146 1176 1116 1146 1157 1116 1146- 1166 Note: HPCS performance criteria requires that the two lowest setpoint valves not be eliminated.
The analysis was performed with the six lowest setpoint valves not credited 'and is conservative.
- With Statistical spreading "' '
l
" Valves ehminated in the analysis NTS = Nominal Trip Setpoint l
6-4
ii ji GE-NE-B13-01760, Rev' 2 i
l Table 6-2 INITIAL OPERATING CONDITIONS FOR ATWS ANALYSIS l Dome Pressure (psia) 1020 Core Flow (Mlbm/hr / % Rated) 94.4/g7 Power (MWt / %) ~3323/100 i i
i Steam /Feedwater Flow (Mlbm/hr / % Rated) 14.3/100 l Feedwater Tsrweiare (*F) 419.0 '.
Initial Void Reactivity Coef5cients (centsrA) 11.0 Core Average Void Fraction 41.3
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6-5
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GE-NE-B13-01760, Rev. 2 3
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Table 6-3 EQUIPMENT PERFORMANCE CHARACTERISTICS FOR ATWS ANALYSIS l
l Nominal Closure Time ofMSIV (sec) 4 l
ReliefValve System Capacity, % Current NBR Steam Flow at 1111 psig/ # ofValve 68.2 %/12 Relief Valve Setpoint Opemng Range (psia) 1126/1176 l
ReliefValve Opening Delay (sec)* -- 0.4 -
ReliefValve Opening Time (sec) 0.15 Relief Valve Closure Delay Time (sec) 0.3 Relief Valve Closure Setpoint Range (psia) 1002/1047-ATWS High Pressure Upper Analytical Limit Setpoint (psia) 1165 ATWS Dome Pressure Sensor Time Constant (sec) 0,1 ATWS Dome Pressure RPT Delay (sec) 0'.374 Relief Valve Out-of-Servicem 2@l076 psig 4@l086 psig
- Note:
(1) Based on the six lowest statistically calculated setpoints. See table 6-1.
- Includes Relief Valve Sensor Tune Constant (sec) l l l l
6-6
GE-NE-B1341760, Rev. 2'
. Table 6-4 MSIV CLOSURE (NO SCRAM)
TRANSIENT RESPONSES
, Peak Peak Peak Peak Heat Flux NeutronFlux Steamhne Vessel
(% NBR) . (%NBR) Press (osie) Press (osin) y MSIV Closure (No Scrae) 138 309 1358 1378 {
Event,100%P/87%F, 12 SRVs in-service
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i 6-7
GE-NE-B13-01760, Rev. 2 Table 6-5 SEQUENCE OF EVENTS FORMSIV CIASURE NO SCRAM Events Time
- 1. MSIV Closure (Nommal 4 sec) Osec !
All Normal Scrams Fail i
., 2. Madmum Poweris Reached 4.0 sec
- 3. Relief Valve Trip on Dome Pressure First Valve 4.13 see Last Valve 4.46 sec
- 5. Maximum VesselPressure Occurs 11.60 sec e 1 9
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GE.NE.B13 01760, Rev. 2 i
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Figure 61 MSIV Closure No Scram Event,100P/87F,6 SRVs OOS 69
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GE-NE-B13-01760, Rev. 2
,' 7.0 SRV AVAILABILITY i
?
The design of the safety relief system is required to meet certain reliability requirements to insure acceptable BWR performance. Availability is defmed as the probability that enough valves (out of all the valves installed and operable) open properly to prevent overpressunzation of the reactor vessel for the design basis vessel overpressure y
event, MSIV Closure with Flux Scram. The valve configuration provided in a design is required to provide a reliability of 5 nmes or greater. The acceptable number ofvalves is i
1 established from this criteria at the design (or in this case design moddication) stage. 'The purpose of the SRV availability part of the SRV Reduction analysis is to ensure that the minimum SRV reliability factor is maintained when a new valve configuration is chosen.
The SRV Reduction project is different from the reloads or special performance plant improvement analyses because a more permanent change to the SRV configuration will be made and this change is like re-establishing the plant design basis configuration.
, Therefore, the minimum number of SRVs required for safe plant operation should maintain compliance to the design basis SRV availability requirement.
The evaluation assumes the SRVs in the safety mode configuration because the
, design basis for vessel overpressure analysis credits only the SRV in the safety mode.
Inputs to the availability study includes:
- 1) The minimum number of valves required to prevent vessel overpressurization, assuming no valve failure-to-open. Based on the MSIV Closure with Flux Scram result (see Section 3), the mmimum number of SRVs is 10.
- 2) The valve-failure-to-open rate is estimated at 0.2E-6 failures / hour and this input is based on over 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />'of plant operating experience for a total of 86 7-1
- i l
GE-NE-B13-01760, Rev. 2 l
Crosby SRVs at LaSalle 1 and 2, Susquehanna 1 and 2, and WNP-2. Only 2 failures-to-open were noted at these units during the data collection period.
1:
- 3) The time interval between scheduled SRV testing and maintenance, in hours.
l The value for this input is 26,784 hours0.00907 days <br />0.218 hours <br />0.0013 weeks <br />2.98312e-4 months <br /> and is based on testing one-third of the installed SRVs during each outt.ge, with a cycle length of 18 months.
The minimum number of operable SRVs required to meet GE's design criteria is 13 valves. This is based on 14 valves installed of which one valve is assumed OOS, and the calculated availability is greater than 0.99999 that 10 of the f
i 13 opearable valves will function when required. As such, up to 4 SRVs can be j l removed in addition to allowing one licensmg basis SRV OOS, for the LSCS units, without affecting the availability requirement of the SRVs.
l However, with only 13 valves installed, of which one valve is as-M i I
i OOS, the calculated availability of 10 functional valves of a total of 12 operable valves is' -0.999986. As such, CECO can consider removing up to 5 valves in addition to allowing assumption of one of the installed valve to be Out Of Service.
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8.0 CONCLUSION
S Based on the GE valuations described herein, the proposed SRV performance requirement changes for LSCS Unit I and 2 as depicted in Table 1-1 have no significant safety impact on l ECCS/LOCA performance, high pressure system (HPCS, RCIC and SLCS) performance, -
containment structuralintegrity, and ATWS analysis results.
l
- e Additionally, this analysis examined cycle dependent safety concerns, such as vessel overpressure margin and thermal limits, demonstratmg that the SRV safety mode tolerance setpoint relaxation up to 3% above the nominal setpoint, combined with up to 6 SRVs OOS (5 SRVs available for removal in addition to one SRV assumed OOS), has no significant impact on plant safety. The SRVs OOS are analyzed based on starting with the lowest opening setpoint l valve and going upward. For future cycles, it is recommended that the LSCS reload licensmg evaluations verify the cycle specific applicability of the vessel overpressure analysis.
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9.0 REFERENCES
1.
I
- 2. ANSI /ASME OM-1-1981, as referenced in Subsection IWV-3500 of the ASME Code, f Section XI,1986 Edition. . 1 3.
l l i l 4.
5.
6.
I
- 7. Commonwealth Edison Company, "LaSalle County Station Mark II Design Assessment Repon", Revision 10, September 1982.
- 8. NUREG-0802, " Safety / Relief Valve Quencher Loads Evaluation for BWR Mark II and III Contamments".
9, -
10.
11.
12.
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14.
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