ML20077C507

Increased Core Flow & Extended Load Line Limit Analysis for Hope Creek Generating Station Unit 1,Cycle 2

ML20077C507
Person / Time
Site:	Hope Creek
Issue date:	11/30/1987
From:	Chao G, Kim H GENERAL ELECTRIC CO.
To:
Shared Package
ML20077C502	List: ML20077C498 ML20077C503 ML20077C507 ML20077C513
References
NEDC-31487, NEDC-31487-01, NEDC-31487-1, NUDOCS 9412050154
Download: ML20077C507 (29)

Text

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NFU-7TD-GE-97-016-00 SELC-31487 CONTROLLED COPY a ERF L12-00765 Approved: 1 Class II November 1987 EAS 108-0987 FINAL REPORT INCREASED CORE FLOW AND EXTENDED LOAD LINE LIMIT ANALYSIS FOR HOPE CREEK GENERATING STATION UNIT 1 CYCLE 2 G. H. Chao H. T. Kim j l

Technical Project Engineers  !

l Approved: m Approved:

A. E. Rogers, Manager D. J. Robare, Manager l Plant Performance Engineering Plant Licensing Services l

D K _, o PD p l P j NUCLEAR ENEAGY BUSINESS OPERATIONS

• GENERAL ELECTRIC COMPANY

! SAN JOSE, CALIFOANIA 95125

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_- . -.- . . . . . - . _-. . -- ~ _ -. . - - - . . - - . .- -- - . . . . -

NEDC-31487-IMPORTANT. NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of General ~Klectric Company respecting information in this document are contained in the contract between Public Service Electrie an.1 Cas Company (PSE&G) and General Electric Company, 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 PSE6G or for any '

purpose other than that for which it is intended, is not authorized; and-with respect to any unauthorized use, General Electric Company makes no'

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representation or warranty, and assumes no liability as to the l

co npleteness , accuracy, or usefulness of the information contained in 'j this document.

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CONTENTS a

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1. INTRODUCTION AND

SUMMARY

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2. ANALYSIS AND RESULTS 2-l'  ;

i 2.1 MCPR Operating Limits 2-1 -

l 2.1.1 Abnormal Operating Transients 2-1 2.1.2 Rod Withdrawal Error 2-2 l 2.1.3- Slow Flow Runo'ut and KgBases 2-3 ,

2.1.4 MCPR Values 2-3 2.2 Overpressure Protection Analysis 2-4 l

2.3 Stability 2-4 l

! 2.4 Loss-of-Coolant Accident Analysis 2-5  !

2.5 Containment Analysis 2-7 2.6 Anticipated Transient Without Scram 7 2.7 Flow-Induced Vibration 2-7 ,

2.8 Load Impact on Reactor Internals 2-8 ]

2.8.1 Acoustic and Flow-Induced Loads Impact 2-8 l 2.8.2 Reactor Internal Pressure Difference Loads Impact 2-9 2.9 Control Rod Dro,p Accident 2-9

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( l j 3. REFERENCES 3-1 j i

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NEDC-31487 TABLES Table Title Eggg l

2-1 Transient Analysis Input and Initial Conditions 2-10 2-2 Transient Peak Value Results - 100% Power 2-11 2-3 MCPR Value Results 2-12 l

2-4 Overpressure Protection Analysis Results - MSIV 2-13 t

l Closure (Flux Scrara) 2-5 ATWS Analysis Summary - MSIV Closure 2-14 l

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NEDC-31487 ILLUSTRATIONS Firure Title Eagg 1-1 Hope Creek Power / Flow Map 1-3 2-1 Generator Load Rejection with Bypass Failure 2-15 100% Power /87% Core Flow 2-2 Feedwater Controller Failure, Maximum Demand 2-16 100% Power /87% Core Flow 2-3 Generator Load Rejection with Bypass Failure 2-17 100% Power /105% Core Flow 2-4 Feedwater Controller Failure, Maximum Demand 2-18 100% Power /105% Core Flow l

2-5 MSIV Closure, Flux Scram 2-19 102% power /105% flow t

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1. INTRODUCTION AND

SUMMARY

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1.1 INTRODUCTION

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Two factors which restrict the flexibility of a boiling water l j reactor (BWR) during power ascension in proceeding from the low-power / low-core-flow condition to the high-power /high-core-flow condition are: (1) the normal operation power / flow map (Reference 1) and (2) Preconditioning Interim Operating Management Recommendations (PCIOMRs).  ;

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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 l

l pellet-cladding interaction considerations inhibit withdrawal of control rods at high power levels. The combination of these two factors can l

result in the inability to attain rated core power directly.

I An expanded operating envelope above rated load line permits improved power ascension capability to full power. The increased core flow region provides an additional extended operating envelope to permit operation to compensate for reactivity reduction due to exposure during an operating cycle.

1.2 SUKHARY l l

This report presents the results of a safety and impact evaluation

for operation of the Hope Creek Cenerating Station (HCGS) Cycle 2 in a l

modified operating envelope called the Extended Load Line Limit Analysis (ELLLA) region. The purpose of this evaluation is to achieve improved power ascension capability to full power as well as to provide addi-tional core flow range at rated power, including an Increased Core Flow (ICF) region to compe'nsate for reactivity reduction due to exposure during an operating cycle. j i

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The extended operating domain is defined and illustrated in Fig- l ure 1-1. The ELLLA region is the operating region bounded by the 108%

APRM rod block line, (0.58 WD + 50%)*, rated power line, and the rated  ;

load line, and the ICF region is bounded by 100% and 105%. core flow lines and 100% rated power lina.

i The limiting operational transients reported in the HCGS Cycle 2 l l Licensing Submittal for rated core flow were reevaluated for end of ,

Cycle 2 at 871 and 105% core flow at 1001 power to evaluate the effects of the ELLLA and ICF operation on the Cycle 2 Operating Limit MCPR l l (OLMCPR) licensing basis. The transient analysis was also performed for [

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the ELLLA and ICF operation for the recirculation pump trip (R?T). system-inoperable option as contained in the current HCGS Technical Specifica- '

tions. The results are given in Tables 2-2 and 2-3. The Cycle 2 Option A and Option B OLMCPR licensing values (Reference 1) should be increased i

to the appropriate values as shown in Table 2-3 fer plant operation in 1

' the extended operating domain, j 9

i The analysis results also show that performance in the extended operating domain is within allowable design limit for overpressure protection, stability, loss-of-coolant accident (LOCA), containment, reactor internals, anticipated transient without scram (ATWS) events, l

flow-induced vibration, and reactor internal pressure difference (RIPD).

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• APRM Rod Block 5 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 Commission (USNRC) in Reference 2.

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. . . - _ .~. _ _ _ _ _ , _ . _ . _ _ _ , _ _ . , _ _ , _ _ . _ , _ _ __ _ . l

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, 140 -

4 I i I j W RECIACULATION LOOP j D = drive FLOW

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1 l (100/06) (10811001 i

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1100/87) (}oofgoo)

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(100/105) s APRM AOD BLOCK LINE

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• AN ALYSIS to.58WD + 50%)

= TECH SPEC 10 86WD+42%) ,w REGION O ~

100% LOAD LINE

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REGION

/ x

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PUwe SPEED

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\ l NATURAL \ j CIRCULATION 1 20 -

CAVITATION PROTECTION o . 20 40 so 80 too 120 CORE FLOW 1%)

Figure 1-1. Hope Creek Power / Flow Map i 1-3 ,

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2. ANALYSIS AND RESULTS 2.1 MCPR OPERATING LIMITS 2.1.1 Abnormal Ooeratine Transients Three limiting abnormal operating transients were reevaluated in the extended operating domain for the ELLLA and the ICF:

(1) Generator Load Rejection with Bypass Failure (LRNBP) )

(2) Feedwater Flow Controller Failure (FWCF) l (3) 100 F Loss of Feedwater Heating (LFWH) l l

l The reevaluation was performed at the power / flow conditions shown in Figure 1-1: (100P, 87F)* for the ELLLA and (100P, 105F) for the ICF.

The evaluation was performed at the end of Cycle 2. Plant heat balance, core coolant hydraulics and nuclear transient parameter data were developed and used in the above transient analysis. The initial condi-tions for the lowest and highest flow point at rated power are presented in Table 2-1. The computer model described in Reference 3 was used to simulate both the Load Rejection With Bypass Failure and Feedwater Controller Failure events. The transient peak values and critical power ratio results for the two cases analyzed are summarized in Tables 2-2 and 2-3, respectively, with the licensing values for comparison. The transient responses are presented in Figures 2-1 through 2-4. The results show that the CPR values for LRNBP and FWCF for the extended operating domain increase slightly compared to the licensing values.

The HCGS plant-specific analysis for the 100 F Loss of Feedvster Heating (LFVH) transient was performed for the ELLLA and ICF operation using the GE Three-Dimensional BWR Core Simulator (Reference 4). The

• This denotes 100% rated power and 87% core flow condtr;on.

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l CPR results are given in Table 2-3. The results show that the LFWH CPR

values for the extended operating domain are bounded by the CPR values i e . t of the LRNBP event.

l The LR'BP J transient was also analyzed for the extended operating l I l

domain for the equipment out-of-service option of the recirculation pump j trip (RPT) system inoperable. The results are included in Table 2-3 for l

i use in modifying the MCPR operating limit for the RPT system inoperable as contained in the current HPGS Technical Specifications. f 1

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l 2.1.2 Rod Withdrawal Error I

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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 l l

rod withdrawal. Thus, the licensing basis evaluation at-the rated  :

1 condition is conservative for operation in'the ELLLA region above the rated load line. )

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When the ICF is employed, the rod block monitor (RBM) setpoint  !

(which is flow biased) increases, giving higher MCPR limit. Thus, the RBM should be clipped at flows greater than 100% of rated so that the CPR values (Reference 1) determined without the ICF apply. The clipping j procedure includes an adjustment to the RBM circuit so that the high RBM tr.ip serpoint at flows greater than 100% rated is equal to the value at 100% rated flow.

The current licensing basis Rod Withdrawal Error (RWE) transient analysis is based on an RBM setpoint of 106% at rated core flow.*

Raising the RBM setpoint allows additional operating flexibility within the extended region of the power / flow map. This improvement can

• The analysis point bounds the HCGS Technical Specifications (Refer-  !

ence 5) value of RBM Upscale Setpoint 0.66W_n + 40%, where VD recirculation loop drive flow in percent or rated.

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be made, but tradeoffs occur. The RWE transient MCPR values are shown in Table 2-3.

With an RBM setpoint of 107% and 108%, the ACPR associated with the RWE increases by 0.02 and 0.07 respectively.

2.1.3 Slow Flow Runout Event and Kg Bases l The purpose of K g is to define MCPR operating limits at off-rated flow conditions. In particular, K is g designed to maintain core thermal margins in the event of a slow flow runout event. That is, K defines g a l

j set of off-rated MCPR limits such that a slow flow runout event ini-tiated from any given off-rated power / flow point will result in a minimum CPR no less than the safety limit MCPR. An evaluation of the slow flow runout event initiated from the limiting point in the ELLLA region up to 107% maximum core flow verified that the existing Kg curves I f are applicable for the expanded operating region. Although the flow will runout along a steeper rod line than would occur in the normal operating domain, the change in core power and MCPR from a given initial core flow will be limited by the recirculation system characteristics such that the K gcurves based on the normal power / flow maps bound the ELLIA and ICF results.

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2.1.4 MCPR Values The results of the safety evaluation show that the MCPR values for l LRNBP and FWCF transients in the ELLLA and ICF regions have increased i

slightly (see Table 2-3). For the ELLLA and ICF operation:

(1) Option A OLMCPR - 1.32 exceeds the value determined for the Cycle 2 Option A OLMCPR licensing value of 1.31.

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(2) Option B OLMCPR - 1.28 exceeds the value determined for the j Cycle 2 Option B OLMCPR licensing value of 1.27.

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For the RPT system inoperable, both Option A and Option B OLMCPR values are determined by LRNBP transient; the LRNBP event at the limit-ing ICF condition (100P,105F) bounds both the Licensing Basis and the ELLLA conditions, and the Option A and Option B OLMCPR values are 1.35 and 1.31, respectively, as shown in Table 2-3.

2.2 OVERPRESSURE PROTECTION ANALYSIS The main steam isolation valve (MSIV) closure with an indirect (flux) scram event is used to determine compliance to the American Society of Mechanical Engineers (ASME) pressure vessel code. This event was analyzed at the bounding condition of (102P, 105F). The results are compared to those '.or the licensing basis point in Table 2-4. As shown, the peak vessel pressure is well below the 1375 psig limit. The l

l transient responses are presented in Figure 2-5.

2.3 STABILITY GE has established stability criteria to demonstrate compliance to requirements set forth in 10CFR50 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 BWR fuel designs, this operating

! mode does not result in specified acceptable fuel design limits being

! exceeded. Furthe rmore , the onset of power oscillations for which corrective actions are necessary is reliably and readily detected and l

suppressed by operator actions and/or automatic system functions. The stability compliance of all licensed GE BWR fuel designs, including those fuels contained in the General Electric Standard Application for Reactor Fuel (CESTAR II, Reference 6), is demonstrated on a generic basis in Reference 7 (for operation in the normal as well as the ELLLA and ICF region). The NRC has reviewed and approved this in Reference 6;

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therefore, a specific analysis is not required. The HCGS Cycle 2 core contains licensed CE BWR core fuel and, hence, the generic evaluation in l

Reference 7 is applicable to HCGS.

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NEDC-31487 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 of operating recommendations has been developed as set forth in Refer-ence 8 and communicated to all GE BWRs. These recommendations instruct the operator on how to reliably detect and suppress limit cycle neutron flux oscillations should they occur, and the HCGS Technical Specifica-tions have implemented these recommendations. The recommendations were developed to conservatively bound the expected performance of all current product lines and are applicable to plant operation in the extended operating domain with ELLLA and ICF.

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! 2.4 LOSS-OF-COOIANT ACCIDENT The effect of increased core flow on LOCA analyses is not signifi-cant because the parameters which most strongly affect the calculated peak cladding temperature (PCT) (i.e., high power node boiling

! transition time and core reflooding time) have been shown to be relatively insensitive to increased core flow.

Results of the LOCA analysis performed for HCGS show that the PCT for ICF increases by less than 10 F for the limiting break compared to the rated core flow condition. PCT changes throughout the remainder of the break spectrum will be of a similar magnitude and thus will not alter the limiting break size or failure. Therefore, it is concluded that the LOCA PCT is acceptable and that the current maximum average planar linear heat generation rates (MAPLHGRs) for HCGS are applicable j for ICF.

i Also, as for ELLLA, the current MAPuiGRs for HCGS are applicable for plant operation in the ELLLA region.

A discussion of' low-flow effects on loss-of-coolant accident (IDCA) analyses for all operating plants (Reference 9) has been presented to I and was approved by the NRC (Reference 10).

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For 251-inch diameter BWR/4 plants like HCGS, the effect of low initial core flow was found to be small. These plants have the smallest-

" effective break ares." (ratio of largest break area to vessel water volume) or any BWR/4 plant. As a result, these plants reflood rapidly following a postulated break and have relatively large peak cladding-tempera ture (PCT)-margins to the 2200 F limit.

The LOCA responses from 102% power /100% flow and from 102% l power /87% flow are very similar'for a BWR/4-251 plant. A LOCA starting from 87% flow produces a slightly earlier-(>0.1 see) loss of nucleate  !

boiling in the top part of 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 followin5 a LOCA. This period is I l

dominated by the effect of counter-current flow limiting (CCFL) at the. j top of the fuel bundles. Since bott. cases being considered have the .

same core power, the steam generation rate in the core and the liquid l 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. l 1

In summary, the standard LOCA analysis for HCGS is applicable for plant operation in the ELLLA and ICF region.

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l NEDC-31487 2.5 CONTAINMENT ANALYSIS l

i The range of power / flow conditions in the ELLLA and ICF operating domain for HCGS was investigated to determine if there would be any impact on containment LOCA response, including the containment dynamic

loads. )

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The results show that all key containment parameters, including drywell pressure and temperature, wetwell pressure and temperature, and l

l suppression pool temperature, plus the LOCA-related containment dynamic loads, including vept thrust, pool swell, condensation oscillation and chugging loads, will remain below their respective design limits during ELLIA and ICF operations.

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l 2.6 ANTICIPATED TRANSIENT WITHOUT SCRAM (ATUS) l l l i

l The impact of the ELLLA and ICF operation on the ATWS response was  ;

i i evaluated.

Table 2-5 summarizes ATWS analysis results for rated conditions and

! operation in the ELLLA region. As indicated, the peak bulk suppression pool temperature limit of 190 F is satisfied at rated conditions with a l 1

4-minute SLCS time delay. However, a time delay of 2 minutes is required l 1

to satisfy the 190 limit for operation in the ELLLA region.

l The results show that ATWS performance in the ELLLA and ICF region is within allowable design limit for overpressure protection, core and fuel performance, stability, and containment structual integrity.

2.7 FLOW-INDUCED VIBRATION l l

To ensure that the flow-induced vibration response of the reactor internals is acceptable, a single reactor of each product line and size undergoes an extensive vibration test during initial plant startup.

After analyzing the results of such tests and assuring that all

responses fall within acceptable limits of the established criteria, the 2-7

NEDC-31487 reactor is classified as a valid prototype in acccrdance with Regulatory Guide 1.20. All other reactors of the same product line and size undergo a less rigorous confirmatory test to assure similarity to the base test. The acceptance criteria used for vibration assessment are based on a maximum allowable alternating stress intensity of 10,000 psi.

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The increased core flow vibration analysis was performed by analyz-ing the startup test vibration data for the valid prototype plant (Browns Ferry 1). Based on the results of the analysis and a review of the test data, the reactor internals response to flow-induced vibration is expected to be within acceptable limits for plant operation in the ICF region.

2.8 LOAD IMPACT ON REACTOR INTERNALS 2.8.1 Acoustic and Flow-Induced Loads Imoact The acoustic loads are lateral loads on the vessel internals that result from propagation of the decompression wave created by a sudden recirculation suction line break. The acoustic loading on the vessel internals is proportional to the total pressure wave amplitude in the vessel recirculation outlet nozzle. The total pressure amplitude is the sum of the initial pressure subcooling plus the experimentally deter-mined pressure undershoot below saturation pressure. An increased downcomer subcooling, as is seen in the ELLLA region, results in a lower saturation pressure, thereby having a larger total pressure amplitude and resulting in larger acoustic loads.

The high-velocity flow patterns in the downcomer resulting from a recirculation suction line break create lateral loads on the shroud and the j e t pump . These loads are proportional to the square of the criti-cal mass flux out of the break. The additional subcooling in the downcomer resulting f' rom operating in the ELLLA region leads to an increase in critical flow and, therefore, in flow-induced loads.

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l The reacter internals most impacted by acoustic and flow-induced loads l are the shroud, shroud support, and j e t pump . The_ impact on these components was evaluated throughout the ELLLA region, The analyses concluded that these components have enough design margin to handle the larger loading during operation in the ELLLA region.

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l 2.8.2 Reactor Internal Pressure Difference Loads Imoacts A reactor internal pressure difference (RIPD) analysis was per-formed for the ELLLA and ICF regions. The reactor internals most l

l affected are the core plate, shroud support, shroud, top guide, shroud head, steam dryer, control rod guide tube, control rod drive housing and ,

i jet pump.  ;

l l l The results show that the RIPDs for the ELLLA operation are bounded by the results for rated power / flow conditions. The increased RIPDs l 4

across the reactor internals and the fuel channels for the ICF operation l were generated for the maximum core flow (105% rated) at normal, upset and faulted conditions. The reactor internals and fuel channels were 1

evaluated using the bounding loads under these conditions. It is I l

concluded that the stresses produced in these components are within the l allowable design limits given in the Final Safety Analysis Report. I I

2.9 CONTROL ROD DROP ACCIDENT HCGS uses banked position withrawal sequences (BPWS) for control rod movement. Control Rod Drop Accident (CRDA) results from BPWS plants have been analyzed statistically, ihe results show that, in all cases, the peak fuel enthalpy in a CRDA would be much less than the correspond-  ;

ing design limit even with a maximum incremental rod worth corresponding l t

to 95% probability at the 95% confidence level. Based on these results, it was proposed to the USNRC, and subsequently found acceptable, to delete the CRDA from the standard GE-BWR reload package for the BPWS plants (Reference 6, Section 5.2.5.1.3, Item 1). Hence, the CRDA is not i analyzed specifically for HCCS.

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NEDC-31487 Table 2-1 TRANSIENT ANALYSIS INPUT AND INITIAL CONDITIONS Power Flow Condition (100P.105F) il22P.87F)

Thermal Power (MWt) 3293 3293 Steam Flow (M1b/hr) 14.17 14.14 .

Core Flow (M1b/hr) 105.0 87.0 Feedwater Temperature ( F) 420 420 Vessel Dome Pressure (psia) 1020 1020 Core Midplane Pressure (psia) 1036 1033 Core Coolant Inlet Enthalpy 527.2 522.8 (Btu /lb)

Turbine Inlet Pressure (psia) 965 965 l

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l Table 2-2 TRANSIENT PEAK VALUE RESULTS  ;

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Initial Peak Peak Peak Peak power / Neutron Heat Steamline Vessel flow Flux -Flux Pressure Pressure Transient (% NBR) (% NBR) (% NBR) (esia) (esir) De1{a)

CPR ,

LRNBP 100/100I ") 404.7 120.3 1164 1199 0.18 100/ 87 369.8 118.5 1164 1196 0.14

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100/105 424.7 121.3 1164 1201 0.20 i

I 233.3 114.0' 1134 1169 0.11 FWCF 100/100 ")

100/ 87 227.1' 112.1 1134 1167 0.09 100/105 252.4 115.7 1136 1172 0.14 LRNBP 100/100 454.3 123.4 1164 1203 .0.21 I) 401.2 120.9 1165 1199 0.16 w/o RPT 100/ 87 l 100/105 454.3 124.7 1165 1205 0.23 l

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(a) Licensing basis point (Reference 1).

(b) Provided to evaluate Tech Spec RPT out-of-service option.

(c) Delta CPR based on an initial CPR which yields a safety

• 1 limit MCPR of 1.07; uncorrected for ODYN Options A and B.

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NEDC-31487 Table 2-3 MCPR VALUE RESULTS i

l Cycle 2 Licensing Basis ICF ELLLA (100P.100F) ,,(100P,105F) (100P,87F)

Ootion A Ootion B Ontion A Ootion B Ootion A Ootion B l

I LRNBP 1.31 1.27 1.32 1.28 1.26 1.22 l

FWCF 1.22 1.20 1.25 1.23 1.20 1.20 l

i LFWH 1.17 1.17 1.17 RWE RBM 1.22 1.22 1.22

-106%

RWE RBM 1.24 1.24 1.24

-107%

RWE RBM 1.29 1,29 1.29

-108%

1 LRNBP 1.34 1.30 1.35 1.31 1.28 1.24 w/o RPT(*}

l (a) Provided to evalu* ate Tech Spec RPT out-of-service option.

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Table 2-4 OVERPRESSURE PROTECTION ANALYSIS

SUMMARY

MSIV CLOSURE (FLUX SCRAM) 102% Power I l

l Peak Peak Steam Line Vessel

~

Initial Core Flow Pressure Pressure

(% NBR) (esir) (esic) 1 1

100 1199 1232 l

I 105 1199 1233 )

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• Licensing basis point (Reference 1). l l

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1 NEDC-31487 Table 2-5 ATUS ANALYSIS

SUMMARY

- MSIV CLOSURE 2-Pump SLCS 2-Pu=p SLCS 2-min Timer 4-min Timer Rated Rated Condition ELLLA Condition ELLLA Initial Power / Flow (% NBR) 100/100 100/87 100/100 100/87 Peak Neutron Flux (% NBR) 710 716 710 710 Peak Average Heat 144 144 144 144 Flux (% NBR)

Peak Vessel Pressure (psig) 1316 1341 1316 1341 Peak Suppression Pool 178 187 189 >190*

Temperature ( F)

Peak Containment 9.0 10.9 11.0 >11.0*

Pressure (psig)

• Estimated O

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NDC - 314S7 1

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l Figure 2-1. Generator Load Rejection With Bypass Failure j

I (100% Power, 87% Core Flow) 2-15 i

NEDC - 31487 i

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! LEdit!N d REF-SEP-SKRT) 1V0:pREt*fiv1TV l 2 DU8rLErgag4:71v!TY 2 WESSEL STE AMFLov 3e --- s u m R!<fiv!TV-3 e--.

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f 150.0 m - _

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19.0 20.0 0.0 19.9

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Figure 2-2. Feedwater Controller Failure, Maximum Demand (100% Power, 87% Core Flow) i i

2-16 I

- - - . . _ ~ . .-

NEDC - 31467 r-I NEUTRON rLun 1 VE5SEL PnE95 #15E(PSI) 2 AVE Sver ACE west rLux 25d[TY VAL 4 rLos 3 CORE INLET rLOW 3'a=>'

RELIEr vALg rLov 15 0. 8 3 00.0 ^

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100.2 ~ l \ 2 00.0 5 ~?  %,

b -

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50.0 100.O 1  :  :  : N

0. 8 . 0.0 _ _ , _ , . , ,

0. 0 2. 0 4.0 6. 0 0.0 2.0 4. 0 6.0 flRE (SE;3NC$1 TIME ISECOCS) i 1 LEVELI!NCh-REr.stp.snmT) 1 v01D REACTIk!TY 2 VESSEL STE A rLOW 2 DOPPLER REACT 1v!TY )

mr;.CW 5Cs i 3.---TJf!NE---STE -, ar _EEey-ACr;Iv1.T.Y.

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8. 0 2.0 4.0 $. 5 0,8 2.8 4. 0 E.8 TIPE (SECCCS) 11ME ISECDeCS)

Figure 2-3. Generator Load Rejection with Bypass Failure (100% Power, 105% Core Flow) i 2-17 1

"DC - 31437 I

150.e

.-ets rtur 1 vEs LEt ratss *!$Etestl f

, ave SM ACE *E AT rLux 2 SAF :TY v 'SE FLOW i 3 :-r; COE[*Lt ittET 3 REL IEr v E rtos 152. 8 ,> r* Flow

• 4 BYP (SS V L E FLOW 1 to. 8 l

5 . . - l 1 C 0. 3 l e

15 54.0 a f.

n E ,. \, .

$ 3. 8 h

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0. 9

g. 0 18.0 20.0 0. 0 18.3 20.0

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2 VESEEL STEAmrLOV 2 DOSFLE RDCT!v!TY 3 $0F6M 3, ere TUSB _i INE

- r STE *MTL0w 1.8 .- m er 4:e I V.cI. T+. Y 1

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• • • . . . . 2.8 t C.0 10.0 2 0. C C. C 18.8 20.0 l

l 71*[ 15t:0C51 f!nt (SECDC $1 I

I e Figure 2-4. Feedwater Controller Failure, Maximum Demand (100% Power, 10$% Core Flow) 2-18 I

NynC - 31/*87 l l

l 1 NEUTRON FLUX 1 VE5SEL P8tESS RISE CP51) 2 ave SURrACE M AT FLUX 2 SAFETY W AlvE FLCv '

3 CCRE INLEf Flow 3 RELIEr m-= ce valve

=r FLCs 3 g. i a 300.0 a &-

er-E I c:. 3

- -) 1 200.0 ,

1 \

l j b z

l l 53.0 ,9 300.0 y

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l 1 1

1 LEVELCIN i nEr-SEP.SMRT)  ! VOID REACfftv!Tv i l 2 WESSEL STEAP5 LOW 2 DDPPLER 3CACT!v1TT 3, SCR,* AM RE ACT I v ! TY, i 200.0 3'cer*-

TUA914..e_CE n er %

- - - AMFLOW 3.0 r_ _ ne e ,wt.

=.

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5 88 fI 100.0 '

g-  ; _ g -

-=

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I 1

f N. . E.

3 t

i E* 8 .f -. . - M .l.0 j

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5. 0 5.0 0. 0 5.0 TIME (SECDW51 TIME ISEC045) 1 I

Figure 2-5. MSIV Closure, Flux Scram (102% Power, 105% Flow) 2-19 i

l. .

NEDC-31487 i

3. REFERENCES l

l

1. HCCS Cycle 2 Supplemental Reload Licensing Submittal, 23A5854, Rev. O, October, 1987.

2. Safety evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 59 to Provisional Operating License No.

DPR-19, Arendment No. 52 to Facility Operating License No. DPR-25, Amendment No. 70 to Facility Operating License No. DPR-20, and j Amendment No. 64 to Facility Operating License No. DPR-30, Common-wealth Edison Company and Iowa-Illinois Gas and Electr.ic Company, Dresden Station Unit Nos. 2 and 3, Quad Cities Statiot Unie Nos. 1 and 2, Docket Nos. 50-237, 50-249, 50-254, and 50-265,

3. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors", Oct. 1978 (NEDO-24154).

l 4. "Three-Dimensional BWR Core Simulaltor," January 1977 (NEDO-20953-A).

l

5. "HCGS Technical Specifications" as amended through April, 1986 (Docket No. 50-354).

6. " General Electric Standard Application for Reactor Fuel (Supplement J for United States)," May 1986 (NEDE-24011-P-A-8-US, as amended). l

7. " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria,' October 1984 (NEDE-22277-P-1). l

8. "BWR Core Thermal Hydraulic Stability," SIL No. 380 Revision 1, February 10, 1984

9. Letter, R.L. Gridley (GE) to D.G. Eisenhut (NRC), " Review of Low Core Flow Effects on LOCA Analysis for Operating BWRs," May 8,  !

1978.

l

10. Letter, D.G. Eisenhut (NRC) to R.L. Gridley (GE), enclosing " Safety Evaluation Report Revision of Previously Imposed MAPLHCR l (ECCS-LOCA) Restrictions for BWRs at Less Than Rated Flow," May 19 )

1978.  ;

i O

1 3-1

l 1

i. .

! LCR 94-02 l

l NLR-N94029 l

l I

l ATTACHMENT 4 l

HOPE CREEK CYCLE 6/REIDAD 5 ANALYSES '

t i

FACILITY OPERATING LICENSE NPF-57 HOPE CREEK GENERATING STATION 7 DOCKET NO. 50-354 i

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