ML20206A569
| ML20206A569 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 04/30/1986 |
| From: | Robare D, Rogers A, Wolf S GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19292H074 | List: |
| References | |
| DRF-L12-00748, DRF-L12-748, NEDC-31139, NUDOCS 8704080076 | |
| Download: ML20206A569 (26) | |
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. . - _ __ _ .__ . _
l DRF L1 0 8 APR L 98 DAC 314 GENERAL ELECTRIC BOILING WATER REACTOR EXTENDED LOAD LINE LIMIT ANALYSIS FOR LIMERICK GENERATING STATION UNIT 1, CYCLE 1 1
p?R4888?R8?8886 GEN ER AL $ ELECTRIC p PDH
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NEDC-31139 DRF L12-00746 Class II April 1986 DAC 314 l
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i GENERAL ELECTRIC BOILING WATER REACTOR EXTENDED LOAD LINE LIMIT ANALYSIS FOR LIMERICK GENERATING STATION ,
UNIT 1, CYCLE 1 S. Wolf Technical Project Engineer Approved 0"F Approved: 41.V -
A E. Rogers, Manager D.J. Robare, Manager 4
Plant Performance Licensing Services Engineering I
NEDC-31139 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Philadelphia Electric Company (PECO) for PECO's use in supporting the operation of Limerick Generating Station (LGS). The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting information in this document are contained in the General Electric Company Extended Load Line Limit Analysis Proposal No. 414-1820-HH1, January 17, 1986. The use of this information except as defined by said proposal, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither the General Electric Company ner any of the contributors to this document maku any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document, or that such use of ruch information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
ii L
NEDC-31139 CONTENTS 8.ag
- 1.
SUMMARY
1-1
- 2. INTRODUCTION 2-1
- 3. DISCUSSION 3-1 3.1 Background 3-1 3.2 Analytical Basis 3-1 3.3 Analysis and Results 3-2 3.3.1 MCPR Operating Limits 3-2 3.3.1.1 Abnormal Operating Transients 3-2 3.3.1.2 Rod Withdrawal Error 3-3 3.3.1.3 Slow Flow Runout and Kf Bases 3-4 3.3.1.4 MCPR. Values 3-4 3.3.2 Overpressurization Analysis 3-5 3.3.3 Stability 3-5 3.3.4 Loss-of-Coolant Accident 3-6 3.3.5 Containment Analysis 3-7 3.3.6 Reactor Internals 3-8 3.3.7 Impact on Anticipated Transients Without 3-8 Scram (ATWS) 3.4 Conclusion 3-8
- 4. REFERENCES 4-1 iii
p NEDC-31139 TABLES Table Title Page, 3-1 Transient Analysis Input And Initial Conditions 3-10
- ELLLA Unique And FSAR 3-2; Core-Wide Transient Analysis Results 3-11 3-3 Required Operating Limit MCPR Values 3-12 3-4 overpressurization Analysis Summary 3-13 MS1V Closure (Flux Scram) 3-5 ATWS Analysis Summary - ATWS MSIV Closure 3-14 r
ILLUSTRATIONS Figure Titic Page 1-1 Limerick Unit 1 Power / Flow Map 1-2 iv
- 1.
SUMMARY
This report justifies the expansion of the operating region of the power / flow map for the Limerick Generating Station (LGS) Unit 1, Cycle 1.
The operating envelope is modified to include the extended operating region bounded by the 108* average power range monitor (APRM) rod block line (0.58 WD + 50%), the rated power line, and the rated load line, as shown in Figure 1-1. In this report, rated power is defined as 3293 MWt.
The technical analysis contained in this report is referred to as the extended load line limit analysis (ELLLA) and the shaded area in Figure 1-1 is referred to as the ELLLA region.
The results of the safety evaluation show that the Load Rejection with Bypass Failure and Feedwater Controller Failure transients at the ELLLA 100% power intercept point (100P, 87F)* are slightly more limiting for critical power ratio than at the licensing basis point (104.3P, 100F). However, for no equipment-out-of-service (E005) and the Rod Block Monitor (RBM) less than or equal to 107% at rated flow, the minieur critical power ratio (MCPR) operating limit is not affected by ELLLt.
operation. For turbine bypass and recirculation pump trip ECOS and/or the RBM at 108% A . rated flow, the LGS Technical Specifications with the incorporation of the MCPR limits of Table 3-3 are adequate for operation in the ELLLA region for Cycle 1 (Figure 1-1).
The analyses results also show that performance in the ELLLA regior.
is within allowable design limits for overpressure protection, stability, loss-of-coolant accident (LOCA), containment, reactor internals and anticipated transient without scram (ATWS) events.
- P = % power, F = % core flow.
1-1
NEDC-31139 160 140 -
WD = RECIRCULATION LOOP DRIVE FLOW 120 -
1100/89) (108 1001 1100/87) 100 -
1100 10o3 I
~
m APRY ROD BLOCK LINE
- A 80 -
SIS ( . BWD + 50%)
- TECH SPEC to.66WD + 42 %)
ELLLA REGION 60 -
100% LOAD UNE
/
f MINIMUM PUMP SPEE s NATURAL CIRCULATION
~
20 -
CAVITATION PROTECTION O I i O 20 40 EO 80 100 120 CORE FLOW (%)
Figure 1-1. Limerick Unit 1 Power / Flow Map 1-2
- 2. INTRODUCTION Two factors which restrict the flexibility of a boiling water reactor (BkTs) during power ascension in proceeding from the low power /
low-core-flow condition to the high power /high-core-flow condition are:
(1) the updated Final Safety Analysis Report (FSAR) power / flow curve, and (2) Preconditioning Interim Operating Management Recommendations (PCIOMRs).
If the rated load line control rod pattern is maintained as core flow is increased, changing equilibrium xenon concentrations will result in less than rated power at rated core flow. In addition, fuel pellet-cladding interaction considerations inhibit withdrawal of control rods at high power levels. The combination of these two factors can result in the inability to attain rated core power directly.
This report provides the analytical basis for operation of the LCS Cycle 1, under a modified operating envelope to permit improved power ascension capability to full power within the design bases previously applied.
The operating envelope is modified to include the ELLLA regien bounded by the 108% APRM rod block line (0.58 WD + 50%), the rated pover line, and the rated load line as shown in Figure 1-1.
2-1
- 3. DISCUSSION
3.1 BACKGROUND
Operation of the LGS utilizing the standard power / flow cap is described in Chapter 4 of the LGS FSAR (Reference 1). This section of the FSAR describes the basic operating envelope (Figure 4.4-2) within which normal reactor operations are conducted and provides the basic philosophy behind the power / flow curve.
This analysis expands the operating domain along the 108% AFRM rod block
- line to 1001 power at 87% flow. Rated power operction at any flow between 87% and 1001 is acceptable. The expanded operating map is shown in Figure 1-1.
3.2 ANALYIICAL BASIS A codified power / flow curve has been derived to provide relief fro:
the operating restrictions inherently imposed during ascension to power by the existing power / flow curve and PCIOMRs. Five design basis cbjectives were specified in deriving this operating curve:
- a. For those transients and accidents that are sensitive te variations in power and flow, the 104.3% power /100% flow (licensing basis) point must be shown to be a more limiting condition than any condition within the FT.T.TA region (i.e. , the shaded region of Figure 1-1). Otherwise, revised operating limits for the ELLLA region must be defined.
- b. In no instance shall the ratio of power to flow intentionally exceed the ratio defined by the APRM rod block line.
AFRM Rod Block < 0.58D W +50%, where n W is recirculation drive flow in percent of rated. This less restrictive equation was approved by the United States Nuclear Regulatory Coc:=ission (USNRC) in Reference 2.
3-1 l
- c. The slope of the APRM rod block line must be such that flow k increases are capable of comp'ensating for xenon buildup while increasing reactor power to rated power at rated core flow.
- d. The consequences of all accidents and transients analyzed in the FSAR must remain within the limits normally specified for such events.
- e. Reactor power ascension from minimum recirculation pump speed to full power shall be directly attainable through combined control rod movement and recirculation flow increase without violation of either the APRM rod block power / flow line or PCIOMRs.
3.3 ANALYSIS AND RESULTS 3.3.1 MCPR Operating Limit 3.3.1.1 Abnormal Operating Transients All abnormal operating transients described in Limerick FSAR Chapter 15 were examined for ELLLA operation. The most limiting transients are the Load Rejection With Bypass Failure (LRNBP), the Feedwater Controller Failure (FWCF), and 100*F Loss of Feedwater Haating (LFWH) events.
For the ELLLA, these events were analyzed at the 1001' power intercept point (100P, 87F). These analyses were performed using the nuclear parameters resulting from end-of-cycle 1 (EOC1) target exposure shape consistent with rated operation.
The 100*F LFWH results in Limerick FSAR Chapter 15 are applicable to the ELLLA region. A generic statistical LFWH analysis using the computer model described in Reference 3 and methodologies described in Reference 4 3-2 e
NEDC-31139 utilizes a large data base to establish bounding results for this event.
It is found that the LFWH response initiated froc all ELLLA power / flow conditions is bounded by the rated condition 55 probability 95'.
confidence values. This generic bounding ACPR value is bounded by the value documented in Chapter 15 for this event.
The co=puter model described in Reference 5 was used to simulate both the LRNBP and FWCF events. The analysis inputs and initial conditions unique for ELLIA operation are presented in Table 3-1, with the licensing basis case for comparison. The transient peak value results and critical power ratio ( ACPR) results are sum =arized in Table 3-2. The results show that the CPR values for LRNBP and FWCF in the ELLIA region have increased slightly. However, the LCPR values are still below the bounding FSAR ACPR value of 0.16 based on the LFWH event.
The transients were also examined for equipment-cut-of-service options, including turbine bypass (BP) inoperable and both BF and recirculation pu=p trip (RPT) inoperable. Results for the liciting transient with these options out-of-service (FWCF w/o BP and FWCF w/o BF w/o RPT) are included in Tables 3-2 and 3-3.
3.3.1.2 Rod Withdrawal Error The current rod block monitor (RBM) setpoint is a function of power and flow. Above the rated rod line, the rod block will occur with less rod withdrawal. Thus, the licensing basis evaluation at the rated condition is conservative for operation above the rated load line.
The current licensing basis Rod Withdrawal Error (RWE) transient analysis is based on an RBM setpoint of 107% at rated core flow.*
The analysis point buunds the Limerick Technical Specifications (Reference 6) value of RBM Upscale Setpoint t 0.66 WD + 401, where WD is recir:ulation loop drive flow in percent of rated 3-3 l
l
NEDC-31139 Raising the RBM setpoint allows additional operating flexibility within the expanded region of the power flow map. This improvement can be made, but tradeoffs occur.
With an RBM setpoint of 108% st rated flow, the ACPR associated with the RWE increases by 0.06, from 0.13 to 0.19. With this RBM setpoint, the MCPR operating limit for the Technical Specifications condition of BP inoperabic and RPT operable is determined by the RWE event and increases to 1.25 for both Option A and Option B.
3.3.1.3 Slow Flow Runout Event and KfBases The purpose of K is to define MCPR operating limite at off-rated f
flow conditions. In particular, K is designed to maintain core thermal f
cargins in the event of a slow flow runaut event. That is, K defines a f
set of off-rated MCPR limits such that a slow flow runout event initiated from any given off-rated power / flow point will result in a minimum CPR no less than the safety limit MCPR. Application of the existing K curves f
to the ELLLA region, then, must show that all points in the extended operating region are bounded with respect to the slow flow runout by previous analyses of the normal power / flow cap. Detailed evaluations of the slow flow runout event initiated from the limiting point in the ELLLA region verified that the existing K curves are applicable for ELLLA.
f Although the flow will runout along a steeper rod line than would occur in the ncrmal operating domain, the change in core pcwer and MCPR from a given initial core flow will be limited by the recirculation system characteristics such that the K curves g based on the normal power / flow maps bound the ELLLA results.
3.3.1.4 MCPR Values Based on the fuel cladding integrity safety limit and the results of the transient analyses, the operating limit MCPR values for ELLLA are listed in Table 3-3, with the licensing basis values for co=parison.
3-4
NEDC-31139 3.3.2 Overpressurization Analysis The main steam isolation valve (MSlV) closure with an indirect (flux) scrac event is used to determine compliance to the American Society of Mechanical Engineers (ASFE) pressure vessel code. This event was analyzed at the 100% power intercept point (100P, 87F) for the LGE using the nuclear parameters resulting from the EOC target exposure shapc consistent with rated operation. The results are compared to those for the licensing basis point in Table 3-4. As shown, the peak vessel ,
pressure is vall below the 1375 psig limit and is bounded by the licensing basis point for the LGS.
3.3.3 Stability The General Electric Company has established stability criteria to de=onstrate co=pliance to requirements set forth in 10CFE50 Appendix A, General Design Criteria (GDC) 10 and 12. These stability compliance criteria consider potential limit cycle response within the limits of safety system or operator intervention and assure that for GE Bk1 fuel designs this operating mode does not result in specified acceptable fuel design limits being exceeded. Furthermore, the onset of power esci11ations for which corrective actions are necessary is reliably and readily detected and suppressed by operator actions and/or automatic system functions. The stability compliance of all licensed GE Bk1 fuel designs, including those fuels contained in the General Electric Standard Application for Reactor Fuel (GESTAR, Reference 7), is demonstrated on a generic basis in Reference 8 (for operation in the normal as well as the ELLLA region). The NRC has reviewed and approved this in Reference 9; therefore, a specific analysis is not required. The LGS Unit 1 Cycle I core contains licensed GE Bk'R initial core fuel and, hence, the generic evaluation in Reference 8 is applicable to LGS Unit 1.
To provide assurance that acceptable plant performance is achieved during operation in the least stable region of the power / flow map, as well as during all plant maneuvering and operating states, a generic set 3-5
NEDC-31139 of operating recommendations has been developed as set forth in Reference 10 and communicated to all GE Bk'Rs. These recommendations instruct-the operator on how to reliably detect and suppress limit cycle neutron flux oscillations should '. hey occur, and the LGS Technical Specifications have implemented these recommendations. The recommendations were developed to conservatively bound the expected performance of all current product lines and are applicable to operation with FTT.TA.
3.3.4 Loss-of-Coolant Accident A discussion of low-flow effects on loss-of-coolant accident (LOCA) analyses for all operating plants (Reference 11) has been presented to and was approved by the NRC (Reference 12).
The standard loss-of-coolant accident analysis for LGS is applicable for plant operation in the ELLLA region.
For 251-inch diaceter Bk'R/4 plants like LGS, the effect of low initial core flow was found to be small. These plants have the smallest
" effective break area" (ratio of largest break area to vessel water volume) of any Bk'R/4 plant. As a result, these plants reflood rapidly following a postulated break and have relatively large peak ~ cladding temperature (PCT) margins to the 2200'F limit.
The LOCA responses from 102% core power /100% core flow and from 102*
core power /87% core flow are very similar for a BWR/4-251 plant. LOCA starting from 87% core flow produces a slightly earlier (<0.1 see) loss of nucleate boiling in the top part cf the limiting fuel bundle.
However, it does not affect the dryout time of the high power node where the maximum peak cladding temperature occurs. The reduced initial core flow also has little impact on the reflood phase following a LOCA. This period is dominated by the effect of counter-current flow limiting (CCFL) at the top of the fuel bundles. Since both cases being considered have the. same core power, the steam generation rate in the core and the liquid 3-6
NEDC-31139 downflow rate through the fuel bundles will be similar for both cases.
Thus, the effect of reduced initial core flow on the reflooding time is negligible.
In sue:ary, the standard LOCA analysis for LGS is applicable for plant operation in the ELLLA region. No additional operating restrictions are required to satisfy the requirements of 10CFR50.46 for ISS in the ELLLA region.
3.3.5 Containr.ent Analysis The impact of plant operation in the ELLLA region on the contain=ent LOCA response was evaluated.
The calculated peak values for dryvell pressure and wetwell pressure under ELLLA operation are bounded by the corresponding values for the FSAR (Chapter 6) conditions. The peak value for drywell floor differential pressure (download) for ELLLA operation is virtually the same as for the licensing basis condition and is bounded by the corresponding design limit. All other containment para =eters (e.g.,
drywell and suppression chamber temperatures, drywell and suppression cha=ber external pressures, dryvell and suppression chamber maxicut allowable leak rates) are bounded by the results reported in the FSAR.
The LOCA-related pool swell, condensation oscillation and chugging loads were evaluated at the worst power / flow conditions under ELLLA operation. The conditions which establish pool swell loads under ELLLA operations were compared with the corresponding FSAR (Chapter 6) conditions. Pool swell, condensation oscillation and chugging loads with ELLLA conditions are bounded by the corresponding design loads.
3-7
NEDC-31139 3.3.6 Reactor Internals The reactor internal pressure differences (RIPD) for ELLLA operation are bounded b." the results for rated power / flow conditions. The peak acoustic and flow-induced loads have been evaluated for ELLLA operation.
The reactor internals have been evaluated taking into account the increased peak acoustic and flow-induced loads for ELLLA operation. The results show that the reactor internals have adequate design margin for operation in the ELLLA region.
3.3.7 Impact on Anticipated Transient Without Scram (ATWS)
The ATWS design analysis for the Limerick plant at rated conditions is presented in Reference 13. The limiting Limerick ATWS MSIV closure events were reevaluated for ELLLA operation for the limiting condition of 100P/67F, including both the two pump and three pump Standby Liquid Control System (SLCS) cases. The results show that ATWS performance from the ELLLA region is within allowable design limits for overpressure protection, core and fuel performance, and stability. ATWS analysis results are summarized in Table 3-5. Containment performance results show that the peak pressure suppression pool temperature is below a bulk suppression pool temperature limit of 190'F with a 2-minute timer delay for the two pump SLCS and a 5-minute timer delay for the three pump SLCS.
Automatic Depressurization System (ADS) trip avoidance criteria are also met.
3.4 CONCLUSION
The results of the safety evaluation show that the Load Rejection with Bypass Failure and Feedwater Controller Failure transients at the 3-8
NEDC-31139 ELLLA 100% power intercept peint (100P, 87F) are slightly more limiting than at the licensing basis point (104.3P, 100F). However, for no equip-ment-out-of-service (E005) and the Rod Block Monitor (RBM) less than or equal to 107% at rated flow. the MCPR operating limit is not affected by ELLLA operation. For turbine bypass and recirculation pump trip E005 and/or the RBM at 108% at rated flow, the LGS Technical Specifications with the incorporation of the MCPR limits of Table 3-3 are adequate for operation in the ELLLA region for Cycle 1 (Figure 1-1).
The analyses results also show that performance in the ELLLA region is within allowable design limits for overpressure protection, stability, LOCA, containment, reactor internals and anticipated transient without scram (AIVS) events.
3-9 i
NEDC-31139 Table 3-1 TRANSIENT ANALYSIS INPUT AND INITIAL CONDITIONS -
ELLLA UNIQUE AND FSAR FSAR ELLLA Licensing Basis Point 100% Intercept Poir.t (104.3P/100F) (100P/87F)
Thermal Power (MWt/%) 3435/104.3 3293/100 Steam Flow (M1b/hr/%) 14.86/105 14.14/100 Core Flow (M1b/hr/%) 100.0/100 87.0/87 f Feedwater Te=perature (*F) 425 420 Do=e Pressure (psig) 1020 1005 Core Exit Pressure (psig) 1031 1015 Core Coolant Inlet Enthalpy 526.6 522.8 (Btu /lb)
Turbine Pressure (psig) 960 950 Core Average Rated Void 43.45 45.48 Fraction (1) 3-10 l
NEDC-31139 Table 3-2 CORE-WIDE TRANSIENT ANALYSIS RESULTS Peak Peak P k Initial Peak Heat team ine essel ACPR Power / #*88" * #888"#*
Neutron Flux P8x8R Flow Flux Q/A (% s1 y (unad-Transient" (% NBR (% NBR) Initial) (psig) (psig) justed)
C LRNBP 104.3/100 222.5 106.2 1196 1225 0.08 d
LPSBP 100/87 226.7 107.7 1193 1222 0.09 .
FWCF 104.3/100 C 156.3 105.0 1165 1194 0.06 d
FWCF 100/87 159.3 105.7 1165 1186 0.07 d
FWCF 100/87 188.6 110.9 1191 1221 0.12 w/o BP' d
FWCF 100/87 303.0 119.9 1199 1235 0.19 w/o BP, w/o RPI'
- a. LRNBP = Load rejection with bypass failure, FWCT = feedwater controller failure to maximum demagd, v/o BP = turbine bypass in-operable, w/o RPT = RPT inoperable,
- c. FSAR licensing basis point.
- d. ELLLA 100% power intercept point.
- e. Provided to evaluate Technical Specifications equipment-out-of-service options.
3-11
NEDC-31139 Table 3-3 REQUIRED OPERATING LIMIT MCPR VALUES Licensing Basis Point 100% Intercept Point 104.3P/100F 100P/87F CPR CPR Transient" Option A Option B Option A Option B LRNBP 1.19 1.11 1.20 1.12 FWCF 1.17 1.14 1.18 1.15 c
FWCT w/o BP . 1.22 1.19 1.23 1.20 FWCF w/o BP, w/o RPT c 1.30 1.23 1.30 1.23 CPR CPR LFWh d d 1.22 1.22 RWE RBM = 107' 1.21 1.21 RWE RBM = 108% --
1.25'
- a. LRNBP = Load rejection with bypass failure, FWCF = feedwater controller failure to maximum demand, w/o RPI = RPT inoperable, w/o BP = turbine bypass inoperable, RWE = rod withdrawal error, RBM = rod b.ock monitor.
- b. Includes adjustment factors as specified in the NRC safety evaluation report on ODYN (NEDO-24154 and NEDE-24154P).
- c. Provided to evaluate Technical Specifications equipment-out-of-service options,
- d. Required OLCPR using Option A or Option B with RBM i 107% at rated flow.
- e. Required OLCPR using Option A or Option B with RBM $ 108% at rated flow.
3-12
NEDC-31139 Table 3-4 OVERPRESSURIZATION ANALYSIS SLWARY MSIV CLOSURE (FLUX SCRAM)'
Peak Peak Steam Line Vessel Pressure Pressure Initial P P
- l Power / flow
(% NBR) (psig) (psig) 104.3/100* 1227 1260 100/87** 1215 1250 I
- FSAR licensing basis point.
j ** ELLLA 100% power intercept point.
3-13
NEDC-31139 Table 3-5 ATVS ANALYSIS SLT'.ARY - ATWS MSIV CLOSURE 2-Pump SLCS 3-Pump SLCS 2-min Timer 5-min Timer Rated Condition ELLIA ELLLA Initial Power / Flow (i NBR) 100/100 100/87 100/87 Peak Neutron Flux (% NBR) 698 688 688 Peak Average Heat 148 146 246 Flux (% NBR)
Peak Vessel Pressure (psig) 1337 1349 1349 Peak Suppressicn Pool 182.7 189.8 186.9 Temperature ('F)
Peak Containment 9.6 11.2 11.0 Pressure (psig) 3-14
, 4. REFERENCES
- 1. " Final Safety Analysis Report - Limerick Generation Station," as amended through Revision 45, December 1985.
- 2. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 59 to Provisional Operating License No.
DPR-19, Amendment No. 52 to Facility Operacing License No. DPR-25, Amendment No. 70 to Facility Operating License No. DPR-20, and Amendment No. 64 to Facility Operating License No. DPR-30, Commonwealth Edison Company and Iowa-Illinois Gas and Electric Company, Dresden Station Unit Nos. 2 and 3, Quad Cities Station Unit Nos. I and 2, Docket Nos. 50-237, 50-249, 50-254, and 50-265.
- 3. "Three-Dimensional BWR Core Simulator," January 1977 (NEDO-20953-A).
- 4. Letter, J.S. Charnley (GE) to F.J. Miraglia (h1C), " Loss of Feedwater Heating Analysis," July 5, 1983 (MFN-125-83).
- 5. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," October 1978 (NEDO-24154).
- 6. " Technical Specifications Limerick Generating Station, Unit 1," as anended through February 6, 1986 (Docket No. 50-352).
- 7. " General Electric Standard Application for Reactor Fuel (Supplement for United States)," August 1985 (NEDE-24011-P-A-7-US, as amended).
- 8. " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," October 1984 (NEDE-22277-P-1).
- 3. Letter, C.O. Thomas (NRC) to H.C. Pfefferien (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011, Revision 6, 2
Amendment 8, Thermal Hydraulic Stability Amendment to GESTAR II,"
April 24, 1985.
- 10. "Bk1 Core Thermal Hydraulic Stability," SIL No. 380 Revision 1, February 10, 1984.
- 11. Letter, R.L. Gridley (GE) to D.C. Eisenhut (NRC), " Review of Low-Core Flow Effects on LOCA Analysis for Operating Bhis," May 8, 1978.
- 12. Letter, D.G. Eisenhut (NRC) to R.L. Gridley (GE), enclosing " Safety Evaluation Report Revision of Previously Imposed MAPLHGR (ECCS-LOCA)
Restrictions 1or Bh%s at Less Than Rated Flow," May 19, 1978.
4-1
- 4. REFERENCES (Continued)
- 13. J.M. Grau, " Design Analysis and SAR Inputs f or ATWS Perforc:ance and Standby Liquid Control System (Limerick Units 1 and 2)," General Electric Company, September 1985 (NEDE-22042, Rev. 2).
4-2
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