Text

Extended Operating Domain & Equipment Out-of-Svc for Quad Cities Nuclear Power Station Units 1 & 2. B Wolfe Affidavit Also Encl
	ML20235C487
Person / Time
Site:	Quad Cities
Issue date:	07/31/1987
From:	Charnley J, Hoang H, Sozzi G; GENERAL ELECTRIC CO.
To:	;
Shared Package
ML19304B493	List: ML20235C340; ML20235C382; ML20235C433; ML20235C487;
References
	NEDC-31449, NUDOCS 8709240468
	Download: ML20235C487 (73)
	v • d • e

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NEDC-31449 Class II July 1987 EXTENDED OPERATING DOMAIN AND EQUIPMENT OUT-OF SERVICE FOR QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 H.X. HOANG 1

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Approved: Approved:

l t W G.L. Sozz , Ma a 5 k $N

'.5 W harn1'ey, Application Engineer. Fuels Licensing Ma ger Services l

l GENER AL $ ELECTRIC i

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j IMPORTANT NOTICE REGARDING

( CONTENTS OF THIS REPORT i

Please Read Carefully The only undertakings of the General Electric Co:rpany respecting information in this document are contained in the contract between the customer and the 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 used of this information by anyone other than the customer or for any purpose other than that for which it was intended, is not authorired; and with respect to any unauthorized use, the General Electric makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.

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NEDC-31449 i

TABLE OF CONTENTS Pare l

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SUMMARY

PART A: EXTENDED OPERATING DOMAIN ANALYSIS FOR QUAD CITIES h'UCLEAR POWER STATION UNITS 1 AND 2 A.1 INTRODUCTION A.1-1 A.2 PLANT OPERATIONAL STATUS A.2-1 _. ___

A.2.1 Recirculation System Capability A.2-1 A.2.2 Minimum End-of-Cycle Coastdown Power Level A.2-1 A.3 SAFETY ANALYSIS A.3 1 A.3.1 Abnormal Operational Transients A.3 1 A.3.1.1 Limiting Transients A.3-1 A.3.1.2 Overpressurization Analysis A.3-1 A.3.1.3 Rod Withdrawal Error A.3-2 A.3.2 Fuel Loading Error A.3-2 A.3.3 Rod Drop Accident A.3-2 A.3.4 Loss-of-Coolant Accident A.3-3 A.3.5 Thermal-Hydraulic Stability A.3-3 A,4 MECHANICAL EVALUATION OF REACTOR INTERNALS A.4-1 AND FUEL ASSEMBLY A.4.1 Loads Evaluation A.4-1 A.4.2 Loads Impact A.4-2 A.4.2.1 Reactor Internals A.4-2 A.4.2.2 Fuel Assemblies A.4-2 A.5 FLOW-INDUCED VIBRATION A.5-1 .

A.6 FEEDWATER N0ZZLE FATIGUE ANALYSIS A.6-1 A.6.I Methods and Assumptions A.6-1 A.6.2 Results A.6-3 A.7 CONTAINMENT ANALYSIS A.7-1 iii

NEDC 31449 CONTENTS (Continued)

Pare PART B: PLANT EQUIPMENT OUT-OF-SERVICE FOR QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2 B.1 INTRODUCTION B 1-1 B.2 FEEDWATER HEATER OUT-OF-SERVICE B.2-1 B.2.1 Abnormal Transients Evaluation B.2-1 B.2.2 LOCA Analysis B.2-3 B.2.3 Feedvater Nozzle Fatigue Analysis B.2-4 B.2.4 Containment Loads Evaluation B.2-5 B.3 ONE SAFETY / RELIEF VALVE OUT OF-SERVICE B.3-1 B.3.1 Abnormal Transients Evaluation B.3-1 B.3.2 LOCA Analysis B.3-2 B.4 ONE RECIRCULATION PUMP OUT-OF SERVICE B43 B.4.1 Minimum CPR Fuel Cladding Integrity B.4-1 Safety Limit B.4.2 Minimum CPR Operating Limit B.4-2 B,4.3 Stability Analysis B.4-2 B.4.4 LOCA Analysis B.4-3 REFERENCES R-1 ACKNOWLEDGEMENTS R-2 iv

NEDC-31449 j,

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'o TABLES Table Title by,.c ,

A 3-1 Core-Wide Transient Analysis Results at ICF A.3-5 and ICF/FFWTR /

A 3-2 Core-Wide ACPR Results at ICF and ICF/FTWTR A.3-6 , , -

A.3-3 MCPR Operating Limits at ICF and ICF/FFWTR A.3-7 I '-

for QCNPS Unit 1, EOC10 A.3-4 Overpressurization Analysis Results A 3-7 A.6-1 Feedwater Duty Map Indices Added for FFWTR A.6-4 and Coastdown A.6-2 Nozzle Fatigue Usage for an 11-Year Seal A.6-5 Refurbishment Period for FFWTR and Coastdown (Location I)

B.2-1 Transient Analysis Results for QCNPS at B.2 6 100P/100F, Feedwater Heater Out-of-Service B.2-2 Operating MCPR Results for QCNPS at B.2-7 100P/100F, Feedwater Heater Out-Of-Service B.2-3 Feedwater Duty Map Indices Added for B.2-8 Feedwater Hester Out-of Service B.2-4 No le Fatigue Usage for an 11-Year Seal B.2-9 Refurbishment Period for Feedwater Heater Out-of-Service (Location I)

B.3-1 S/DV Setpoints Used for Transient Analysis B.3-4 v

NEDC-31449 ILLUSTRATIONS lignre Title Pare A.1-1 Extended Operating Domain Power / Flow A.1-2 Map for QCNPS A.31 Plant Response to Load Rejection without A.3-8 Bypass, 100% Power 108% Flow, Rated Feedwater Temperature, EOC+128 MWD /MT A.3-2 Plant Response to Load Rejection without A.3-9 Bypass, 100% Power 108% Flow, FFWTR, EOC+582 MWD /MT A.3-3 Plant Response to Turbine Trip without A.3 10 i

Bypass, 100t Power 100% Flow, Rated l

Feedwater Temperature, EOC A.3-4 Plant Response to Turbine Trip without A . 3 - 1:.

Bypass, 1004 Power 108% Flow, Rated Feedwater Temperature, EOC+128 MWD /MT A.3-5 Plant Response to Turbine Trip without A.3 12 Bypass, 1004 Power 108% Flow, FFWTR, EOC+582 MWD /MT A.3-6 Plent Response to Feedwater Controller A.3-13 Failure, 100% Power 108% Flow, Rated Feedwater Temperature, EOC+128 MWD /MT A.3-7 Plant Response to Feedwater Controller A.3-14 Failure, 100% Power 108% Flow, FFWTR EOC+ 582 MWD /MT A.3 8 Plant Response to MSIV Closure Flux Scran. A.3-15 100% Power 10S% Flow, Rated Feedwater Temperature, EOC+128 MWD /MT l

l A.6-1 Example of Linear Method of Determining A.6-6 Seal Refurbishment Intervals B.2-1 Plant Response to Load Rejection Without B.2-10 Bypass, Feedwater Heater In Service (Bounding Power Shape)

B.2-2 Plant Response to Lead Rejection Without B.2-11 Bypass, Feedwater Heater Out of Service (Bounding Power Shape) vi

NEDC-31449 B.2-3 Plant Response to Feedwater Controller B.2-12 Failure, Feedwater Heater In Service (Bounding Power Shape)

B.2-4 Plant Response to Feedwater Controller B.2 13 Failure, Feedwater Heater Out-of-Service (Bounding Power Shape)

B.3-1 Plant Response to Load Rejection Without B.3-5 Bypass, Full Relief Capacity B.3-2 Plant Response to Load Rejection Without B.3-6 Bypass, One S/RV Out-of Service l

i B.3-3 Plant Response to MSIV Flux Scram, B.3-7 One S/RV Out of-Service vil

NEDC-31449

SUMMARY

This two-part report documents a comprehensive set of analyses performed for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2 to:

(1) Support the expansion of the operating domain of the current power / flow map.

l (2) Justify continuous operation with one of the following i equipment out-of-service: one feedwater heater string, ene safety / relief valve, one recirculation loop.

The first part of the report addresses the technical bases to justify the proposed operating domain expansion, and the second part presents analyses results and associated technical specif f cation limits to support plant operation with certain equipment out-of-service.

The standard operating domain for QCNPS Units 1 and 2 was previously modified to include the operating region above the rated rod line bounded by the 108i average power range monitor (APRM) rod block line, the rated power line and tne rated core flow line (Reference 1). For this analysis, the standard operating envelope is modified to include the increased core flow (ICF) region (Figure A.1-1). This safety evaluation is also applicable to operation beyond nominal end-of-cycle with ICF and up to 100 degree F final feedwater temperature reduction ( FWTR) or beyond nominal end of-cycle at less than rated core flow with reduced feedwater temperature. The cycle extension is then followed by a coastdown to 20%

power.

As part of the expanded operating domain analysis, the limiting abnormal operational transients at rated core flow condition are reevaluated for 100% power and 108% core flow with and without FWTR to support the ICF operation mode throughout the cycle and ICF/FWTR operation ix

NEDC-31449 beyond nominal end-of-cycle. The operating limits obtained for ICF with FFWTR also bound operation at less than rated core flow with reduced feedwater temperature.

I The loss of-coolant accident (LOCA), the containment LOCA response, l and the reactor stability compliance criteria were also evaluated to justify operation at ICF/FFWTR conditions. In addition, the effect of the increased pressure differences due to ICF on the reactor internal components, fuel chant.els, and fuel bundles was also analyzed to show that their design limits will not be exceeded. The effect of ICF on the flow-induced vibration response of the reactor internals was also evaluated to ensure that the response is within acceptable limits. The increase in the feedwater nozzle usage factors due to reduced feedwater temperature was also addressed, _ __ _

Results of the analysis reported herein show that there is no impact on LOCA performance, containment design loads, reactor internal loadings capabilities, or feedwater nozzle fatigue resulting from operation in the expanded power / flow domain. The recirculation system performance data for QCNPS indicate that the currently achievable core flow rate may be less than 10 of rated. However, this is a system performance consideration rather X in a safety concern. Therefore, the analyses were performed at the bot. / ng condition of 108% rated core flow.

For the out-of service equipment mentioned in Item 2 above, the analyses per. med assumed a single failure only and established the licensing bas for continuous plant operation in the expanded power / flow map excluding the ICF region. The feedwater heater out-of service (corresponding to a 100 degree F reduction in feedwater temperature) is included as part of the transient analysis input assumptions. Specific cycle independent operating MCPR limits are established to allow continuous plant operation with this equipment failure. In the case of a safety / relief valve (S/RV) out-of-service, transient analysis results X

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NEDC-31449 showed that there is no impact on the calculated MCPR and that the i l

I overpressure margin to the ASME upset code limit is still satisfied. An analysis justifying recirculation system single loop-operation (SLO) was l previously performed for QCNPS (Reference 2), This analysis has been l

l reviewed and shown to be applicable with the new GE8x8EB fuel design. In i addition, the impact on plant operating limits were also evaluated for SLO l

b mode with or without one safety / relief valve out-of-service in the normal i operating domain as well as in the region above the rated rod line. The analyses of the above-mentioned equipment out-of-service also showed that '

l there is no impact on the LOCA containment response, reactor stability -

performance, or fuel peak clad temperature.

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NEDC-31449

PART A EXPANDED OPERATING DOMAIN ANALYSIS FOR QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2

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NEDC-31449 A.1 INTRODUCTION Most boiling water reactors (Bk'Rs ) have recirculation. system pumping capabilities in excess of that required to provide 100% rated core flow.

The use of increased core flow (ICF) above 100% rated core flew can provide greater operational flexibility in reaching and maintaining full power during tha cycle and can extend the operating cycle at rated power. The magnitude of this extension is dependent on the characteristics of the core and on the maximum allowable core flow. In general, operation with ICF can extend full power operation by approximately one week.

Final feedwater temperature reduction (FFWTR) at the end-of-cycle (EOC) can further extend the operating cycle. In general, a 100 degree F reduction in feedwater temperature provides approximately two weeks of additional full thermal power operation. The ICF region (Figure A.1-1) is referred to as the extended operating domain.

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NEDC 31449 A.2 PLANT OPERATIONAL STATUS A.2 1 RECIRCULATION SYSTEM CAPABILITY Based on recirculation system performance data for QCNPS, the maximum core flow is expected to be somewhat less than 108% of rated core flow.

Given that the ICF capability results from a system performance .

-l consideration rather than a safety concern, the analyses are performed at ,

the bounding condition of 108% of rated core flow.

I A.2.2 MINIMUM END OF-CYCLE COASTDOWN POWER LEVEL Standard licensing transient analyses are performed at the full power, i end-of-cycle (EOC), all-rods-out condition. Once an individual plant !

J reaches this condition, it may shut down for refueling or it may be placed l in a coastdown mode of operation. For QCNPS, this type of operation I involves coasting down to a lower percent of rated power while maintaining 1 constant core flow. For the purpose of this cafety evaluation, coastdown to 20% power is assumed.

The power profile during this period is assumed to be a linear f function with respect to exposure. It is expected that the actual profile will be a slow exponential curve. An analysis using the linear approximation, however, will be conservative, since it bounds the power level for any given exposure. In Reference 3, evaluations were made at I 90%, 80% and 70% power level points on the linear curve. The results showed that the pressure margins from the limiting pressurization transient and the MCPR operating limits exhibit a larger margin for each of these i points than the EOC full power, full flow case. The LOCA analysis results l

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NEDC-31449 for the rated power /108% of rated core flow case is conservative for the coastdown period since the power will be decreasing and ICF will be maintained. Reference 4 presents the results of analyses to justify coastdown power operation.

i A.2-2 A

NEDC-31449 i

A3 SAFETY ANALYSIS A.3.1 ABNOPJiAL OPERATIONAL TRANSIENTS A.3.1.1 L.lyitinn Transients The limiting operational transients analyzed in the QCNPS Unit 1 Reload 9, Cycle 10 supplemental reload licensing submittal (Reference 5) were re-evaluated for ICF followed by final feedwater temperature reduction (FFWTR) as follows:

Nuclear transient data for 100% power, 108% core flov (200P/108F) with and without FFWTR (corresponding to a 100 degree F reduction in feedwater temperature) were developed for rated poaer at exposures beyond end of-cycle 10 (EOC 10). These nuclear data were then used to analyze the load rejection without bypass (LRNBP), turbine trip without bypass (TTNBP) and the feedwater controller failure (FWCF) events at the 100P/108F condition. The transient events were analyzed based on core characteristics with both BP/P8x8R and CE8xBEB.

The results of the transient analyses are presented in Tables A. 3 1, A . 3 2 and A . 3 3. As shown in Tables A. 3-2 and A. 3 3, the limiting event (LRNBP) from Reference 5 bounds all of the taCPR results from the ICF and ICF/FFWTR analyses with one exception: the BP/P8x8R limits for LRNBP and TTNBP at ICF with rated feedwater temperature. Therefore. the current technical specification MCPR operating limits should be modified to incorporate the changes as shown in Table A.3-3 for operation at ICT conditions. The transient performance responses for the LRNBP, TTNBP and FWCF events are shown in Figures A.3-1 through A.3-7.

A.3.1.2 overpressurization Annivsis The limiting transient for the ASME code overpressurization analysis

[ main steam isolation valve (MSIV) closure with flux scram (direct scrax failure)], was evaluated for extended EOC10 conditions with ICF without A.31

NEDC-31449 FFWTR (Table A.3-4 and Figure A.3-8). For this evaluation, ICF without FWTR is more severe than with FFWTR, and the MSIV closure with flux scram event provides the most limiting overpressure transient response. The transient analysis (Table A.3 4) for the ICF condition produced a peak vessel pressure of 1327 psig, which is below the ASME Code upset limit of 1375 psig and, therefore, is acceptable.

A.3.1.3 Rod withdrawal Error The rod withdrawal error (RWE) transient was evaluated under ICF conditions. When ICF is employed, the rod block monitor (RBM) setpoint (which is flow biased) increases and gives a higher MCPR limit. Thus, the REM should be clipped at flows greater than 100% of rated so that the 6CPR values determined without ICF (Reference 5) apply. The clipping procedure includes an adj us tment to the REM Ifreuit so that the high RBM trip setpoint at flows greater than 100% of rated is equal to the value at 100%

rated flow. These results are independent of whether FWTR is implemented or not, and are therefore bounding.

A.3.2 FUEL LOADING ERROR Operation with ICF and/or FFWTR does not impact the analysis of the dotnted bundle fuel loading error event. Thus, the results reported in the Cycle 10 reload licensing submittal (Reference 5) are applicable for operation with ICF And/or TFWTR, A.3.3 ROD DROP ACCIDENT The rod drop accident (RDA) event is a startup accident evaluated at cold and hot standby core conditions which are unaf fected by ICF and FFWTR operation. Therefore, there is no change to the RDA analysis bases as presented in Reference 5, and the RDA requirements of Reference 5 are applicable for operation with ICF and/or FFWTR.

A.3-2

,s NEDC 31t:49 1

I A.3.4 LOSS-OF-COO 1 ANT ACCIDENT (LOCA) ANALYSIS - i s

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For core flows 1c,wer than a critical value, boiling transition at the' limiting fuel node (i.e., the high power node) can occur sooner than observed at rated, operating conditions. This phenomenon is referred.to as early boiling trarsition (EBT). If EBT occurs for the high power nude at reduced flow, tip resultant peak cladding temperature tPCT) can exceed the

, t rated condition results. If thqre is no PCT margin 't> regulatory limits, j i

it may be necessary to apply a maximum average plaur linear heat i generation rate (ME LilGR)__ multipl:xr for' operetion in cdctain flow ranges.

Low flow effects weh generically addrtsssed in ciklerence 6, whihh sas.

approved by the Reference 7 NRC Safety Evaluation, he' port. It show d that no MAPLHGR multiplier is required for the QCNPE,ciass of plant.

t The effects of .ICF anc'/or JFWTR on LOCA analyses are insignificant ,

because the parameters which most strongly affect tt.e ulculated WT (i.e. , .

high power node boiling transition time and core rc;1oWiing time) haw been shown to be relatively insensitive to core f1w and feedwater traperature changes of this magnitude. Both of these modes of operation tewi to slightly improve the results. With the lower initjal care void fraction, there is more !iquid trass to be lost out cf the break ta fore core uncovery results. The not effect of void fractio @ add other effects will result in a LOCA PCT change of less then U degrees V, which is i. significant in view o f tha- l arf;e PCT margins from the new SAFER /GESTR LOCA analysis (R(ference f 8). ,

Therefore, it is concluded that t%s ' LOCA analysis rest its fdr QCNPS are applicabli and ' insensitive to operation with ICF and/or FFs'TR. _

A.3.5 Thi.RMAL HYDRC'LIC : STABILIT'i The core .ir.d channel hydredyr.anic decay ratio were evaluated itr !CF snd/or FFWTR'o.petation. If tW reactor initially oper;tes at 1CF and at or A.3-3

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.e i f. only FFWTR is utilized, operation should be at or below the rated rod lie. However, the combined effect of operating the reantor with ICF first and -:. hen with FFWTR would result in a lower overall core and channel decay r tio than for normal operation.

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Therefore, it is concluded that the reactor core stability and the .

channel hydrodynamic stability performance with ICF and/or FFWTR are within j the established criteria.

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Table A.3-2 l

L CORE-WIDE ACPR RESULTS Uncorrected Option A Option'B Iransients Exposure EZE FFWTR BP/8X8R GE8X8FB BP/P8X8R GE8X8EE HZZf8X8R GE8XPEB' LRNBP EOC 100/100 No 0.19 0.19 0.26 0.26 0.21 0.21 EOC + 128 100/108 No 0.20 0.19 0.27 0.26 0.22 0.21 EOC + 582 100/108 Yes 0.18 0.18 0.25 0.25 0.19 0.20 TTNEP EOC 100/100 No 0.19 0.19 0.25 0.26 0.20 0.21 EOC + 128 100/108 No 0.20 0.19 0.26 0.26 0.21 0.21 EOC-+ 582 100/108 Yes. 0.17 0.17 0.23 0 23 0.18 0.18 F'iCF EOC 100/100 No 0.14 0.14 0.20 0.20 0.15 0.15 EOC + 128 100/108 No

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NEDC-31449 l

Table A.3-3 MCPR OPERATING LIMITS FOR QCNPS UNIT 1, EOC 10 Option A Option B Transient- EP/P8X8R GE8X8EB BP/P8X8R GE8X8EB

-LRNBP 1.33 1.33 1.28 1.28 (100P/100F)

LRNBP 1.34 1.33 1.29 1.28 (100P/108F, Rated I Feedwater Temperature)

LRNBP 1.32 1.32 1.26 1.27 (100P/108F, FF'n'IR)

Table A.3-4 OVERPRESSURIZATION ANALYSIS RESULTS Maximum Maximum Initial Iritial Steamline Vessel Power Flow Pressure Pressure IRANSIENT (4) (%) (PSIGT (Psir) Finure Nu-ber MSIV Closure - Flux Scram 100 100 3295 1319 Reference 5 (Reference 5, EOC)

MSIV Closure - Flux Scram 100 108 1290 1327 Figure A.3-8 (ICF w/o FF'a'TR, EOC+128 MWD /MT) i A 3.7 i

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A.3-9

NEDC-31449 fvti$rdfuraf Ftur ,

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l A.3-10

NEDC-31449 egurn> Flux tts 4L PetEss mistres!3 2 Avt sunrACE e(AT FLUX SAF :TT WALVE FLDs 8 coar '* n F'o'

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A 3-11

NEDC-31449 i

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A.3-12

NEDC-31449 t

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Figure A.3.6 - Flant Response to Feedwater Controller Failure, 200% Power 1081 Flow, Rated Feodwater Temperature 20C+128 WD/MT.

A.3-13

NEDC-31449 i 94.0

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1 A.3-14 1

NEDC-31449 tEuroch Flux

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A.3-15

WEDC-31449 A.4 MECHANICAL EVALUATION OF REACTOR INTERNALS AND FUEL ASSEMBLY

'A.4.1 LOADS EVALUATIONS Evaluations were performed to determine the bounding acoustic and flow-induced loads, reactor internal pressure difference loads, and fuel-support loads for ICF and/or FWTR operation.

A postulated sudden break in the recirculation line is accompanied by the propagation of a decompression wave which originates at the break and ,

propagates back toward the vessel. Once in the vessel, the wave would broaden and lose intensity. However, it can create lateral loads on the vessel internals located opposite the recirculation suction line connections to the vessel. The pressure wave amplitude will be larger if the subcooling to the downcomer is increased and, therefore, the lateral loads could increase. The high velocity flow patterns in the downcomer resulting from a recirculation suction line break also create lateral loads on the reactor vessel internals. These loads are proportional to the square of the critical mass flow rate out of the break. The additional subcooling in the downcomer resulting from FWTR operation can lead to an increase in the flow-induced loads. The reactor internals most impacted by acoustic and flow induced loads are the core shroud, shroud support and j e t pumps.

A reactor internals pressure difference (RIPD) analysis was performed to evaluate the effect of ICF operation on the reactor internal components loadings. The increased internal pressure differences across the reactor internals were computed for the 108% rated core flow at normal, upset and faulted conditions for the reactor internals impact evaluation.

Based on the results from plant-specific fuel lift analyses performed at 108% core flow, the resulting impact of ICF operation on the fuel-support loads and fuel bundle lift for QCNPS were evaluated. Fuel-support loads and fuel bundle lift were evaluated for upset, faulted and fatigue load combinations. It was shown that the fuel bundle lift is expected to be A.4-1

NEDC-31449 minimal, and the design basis vertical loads on the fuel assembly and its supports remain valid.

A.4.2 LOADS IMPACT A.4.2.1 Reactor Internals The reactor internals most affected by ICF and/or FFWTR operation are the core plate, shroud support, shroud, top guide, shroud head, steam dryer, control rod guide tube, control rod drive housing and jet pump. These and l

other components were evaluated using the bounding pressure differential loads, as calculated in Section A.4.1, under normal, upset and foulted conditions. It is concluded that the stresses produced in these and other f components are within the allowable design limits & iven in the Final Safety Analysis Report or the ASME Code,Section III.

A.4.2.2 Fuel Assemblies 1

i' The fuel assemblies, including fuel bundles and channels, were evaluated for ICF operation considering the effects of loads discussed in Section A.4.1 under normal, upset and faulted load combinations. Results of the evaluation demonstrate that the fuel assemblies are adequate to l withstand ICF effects up to 108% core flow.

The fuel channels were also evaluated under normal, upset and faulted conditions for ICF operation. The channel wall pressure gradients were found to be within the allowable design limits. {

k 1

l l

A.4-2 )

I

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ __. __ _ _J

r NEDC 31449 l

l A.5 FLOW-INDUCED VIBRATION 1

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 test and assuring that all responses fall !

within acceptable limits of the established criteria, the reactor is classified as a valid prototype in accordance 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.

Both QCNPS Units 1 and 2 are BWR/3 251 inch size plants. Reactor internal vibration measurements were conducted at QCNPS Unit 1. Since Unit 2 internal components are similar to Unit 1, the same approximate vibration levels are expected at Unit 2. Test data from Unit I were analyzed based on current GE standard design bases procedures and acceptance criteria based on a maximum allowable alternating stress intensity of 10,000 psi. The results showed that, at the rated core flow condition (98 M1b/hr), the maximum vibration amplitude observed was 40% of the acceptance criteria. In addition, the amplitude of vibration is assumed to be proportional to the square of the flow velocity. Therefore, ICF operation at 108% of rated core flow would increase the vibration level to approximately 65% of the acceptance criteria. Therefore, it is concluded that the vibration level remains acceptabic for operation at 108% core flow.

A.5-1 L-__-_-__.

r NEDC 31449 I

l A.6 FEEDWATER N0ZZLE FATIGUE ANALYSIS j i

i An evaluation of the effect of FFWTR and end-of-cycle coastdown on 1 feedwater nossle fatigue was performed for _ the QCNPS with the following assumptions:

(1) An 18-month fuel cycle.

i i

(2) FWTR to 230 degrees F (equivalent to 100 degree F reduction) 1 for 14 days was followeo by a coastdown to 70% power over a f

. i period of 12 weeks. The feedwater temperature _at the end of the i 1

coastdown was 210 degrees F. ;

l A 6.1 METHODS AND ASSUMPTIONS The fatigue experienced by the feedwater nozzle results from two phenomena: (1) system cycling and (2) rapid cycling. System cycling is I caused by maj or temperature changes associated with system transients.

Rapid cycling is caused by small, high frequency temperature fluctuations caused by mixing of relatively colder no :le annulus water with the hot reactor water. Yhe colder water impinging the nozzle originates from leakage past the thermal sleeve secondary seal and from the boundary layer of cold water formed by heat transfer through the thermal sleeve.

FWTR affects only the rapid cycling f atigue usage for two reasons: ,

(1) the transient temperature variation associated with these modes of operation is small and thus does not affect the system cycling usage l factor, and (2) the time spent at a reduced feedwater temperature is a significant contributor to rapid cycling fatigue usage. An updated rapid I

cycling analysis performed in Reference 10 was revised to include the J condition for FWTR and coastdown to 70% power. l I

I The feedwater duty map, (Table 3-2 in Reference 10), was modified to include the additional indices shown in Table A.6-1 for FWTR operation I

with coastdown. These additional indices model the coastdown as a 4

A.6-1 l i

1

NEDC-31449 four-step process. The temperatures and flow rates are set for each step to give conservative results. The percentage of time spent in the FFWTR and coastdown is subtracted from the percentage spent in normal operation.

Seal ring refurbishment time is determined so that by the end of the feedwater nozzle life, the sum of the system cycling and the rapid cycling usage factor for each the feedwater nozzle locations will not exceed the allowable value of 1.0. It is assumed that the system cycling usage factor l l

is linearly dependent on the number of years since the beginning of feedwater nozzle life. After each year, the total rapid cycling usage ,

factor from the beginning of life is compared to the maximum allowable rapid cycling usage factor for each year, which is determined as follows:

U - UF!MX - SCUF. yrs LIFE where:

U, - Maximum allowable rapid cycling usage f actor from beginning of life for specific year UFMAX - Maximum allowable rapid plus system cycling usage factor by the end of the feedwater nozzle life SCUF - Total system cycling usage factor for the feedwater nozzle life LIFE - Feedwater nozzle design life (years)

I yrs - Number of years since beginning of feedwater nozzle life ;

J l

If the total rapid cycling usage factor since beginning of life '

j exceeds U,,x for any feedwater nozzle locations, seal ring refurbishment is A.6-2 1

l NEDC-31449 l

l l

assured at the end of the previous year. This method is illustrated in Figure A.6-1.

l A.6.2 RESULTS The analysis documented in Reference 10 indicated that refurbishrtent of the thermal sleeve . seals after 11 years would be necessary to keep the 40-year total fatigue usage (system cycling plus rapid cycling) below a value of 1.0. Keeping the refurbishment schedule constant for the analysis, the 40-year total fatigue usage was calculated as shown in Table .' ]

A.6 2. The fatigue damage per cycle for FWTR operation is conservatively )

estimated by taking the dif ference between the FWTR fatigue and the normal I

operation fatigue and dividing that quantity by the number of cycles in 40 j years. j l

If FWTR and coastdown were used for every cycle during the plant's life, the 40 year total fatigue usage factor would be greater than 1.0, assuming that the seals were replaced after 11 years. Satisfactory fatigue )

t usage can be achieved by reducing the refurbishment interval to seven (7) {

years, as noted in Table A.6 2 assuming FWTR at the end of every cycle, the refurbishment interval is impacted by four (4) years.

The results of this analysis are based on the expected coastdown )

1 operational strategry (Section A.6) and on leakage correlations developed i during testing of the triple-sleeve design. A shorter end of-cycle coastdown period or a smaller temperature reduction would increase the j refurbishment interval. Also, the leakage is based on several geometric factors and assumed corrosion rates for the sleeve and safe-end materials.

The resulting fatigue results are conservative for the expected plant I operation mode. Since leakage is the primary contributor to rapid cycling l I

{ fatigue, a more accurate evaluation of rapid cycling could be made by l

i

! monitoring seal leakage and considering actual plant performance.

l, 1

A.6-3 l

t NEDC-31449 Table A.6-1 l l

l FEEDWATER DUTY MAP INDICES ADDED FOR FFWTR AND COASTDOWN ,i Cycle Feedwater Feedwater Region A Time Index Flow Temperature Temperature Year-(% rated) ( F) ( F) (%)

20 100 225 546 2.56 21 100 220 546 3.84 i

22 92.5 215 546 3.84 23 85 210 546 3.84 -

)'

24 77.5 205 546 3.84 Notes: (1) The feedwater temperature is based on a lower value of a +3% variation on the nominal temperature.

(2) The time spent at this mode of operation (2.56 + 3.84 + 3.84 + 3.84 + 3.84 - 17.92%)

was subtracted from normal operation (i.e., index 1 - 65.20 - 17.92 - 47.28%).

A.6-4

NEDC-31449 Table A.6-2 N0ZZLE FATIGUE USAGE FOR A 11-YEAR SEAL REFURBISHMENT PERIOD FOR FFWTR AND COASTDOWN (LOCATION I) 18-Month Cycle FFWTR/Coastdown Normal to 70% Power Operation Each Cvele 40 Year Total Fat 15ue Usage 0.5735 2.1843 Additional Usage Due To FFWTR and Coastdown -- 1.6108 .

l Additional Usage Fer Cycle -- 0.0600 Note: The total 40-year usage factor for FFWTR operation after every cycle can be kept to below 1.0 by refurbishing the seals af ter 7 years (at location D).

I I

i A.6-5

hTDC-3144 9 j i

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1 u -

GUFM AXI ALLOWABLE USAGE

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YEARS i Figure A.5.1 Example of Linaar Method of Determining Seal Refurbish:ent Intervals A.6-6

NEDC-31449 A.7 CONTAINMENT ANALYSIS t

.The impact on the containment LOCA response was evaluated for QCNPS with regard to operation in the expanded power / flow map (i.e., including operation above the rated _ rod line and in the ICF region with and/or without FFWTR).

I The important containment parameters considered in the analysis include:

(1) Drywell pressure and temperature (2) Suppression chamber airspace pressure and temperature (3) Drywell-suppression chamber differential pressure (4) Suppression pool temperature (5) Annulus pressurization loads (6) Hydrodynamic loads.

Results of the analysis showed that the peak values of drywell pressure and temperature, suppression chamber airspace pressure and temperature, suppression pool temperature, and annulus pressurization loads are bounded by the values reported in the Plant-Unique Load Definition (PULD) (Reference 11). Major containment hydrodynamic loads postulated to occur in a hypothetical LOCA were evaluated and included pool scell load, condensation oscillation (CO), and chugging loads. All these dynamic loads are bounded by their corresponding design values except for the vent line thrust load.

The peak calculated value for the vent line thrust load is 21% higher than the PULD reported value. This load represents only a part of the total maximum vent system discharge load (i.e., the vent thrust load is just one component of the maximum vent discharge load combination). From QCNPS Plant-Unique Analysis Report (Reference 12), the margin for the f

maximum vent system discharge load (allowable versus calculated) is A.7-1

NEDC-31449 l

estimated at 25%. The vent line thruct load contributes 19% to the total ,

load combination. Therefore, while this vent thrust load component exceeds the PULD-value, the total vent discharge load remains well within the existing design margin.

S/RV loads on the containment are not affected because there is no change in the S/RV setpoints or reactor operating pressure associated with operation in the extended operating domain.

l 9

A.7-2

)

, . d

NEDC-31449 PART B PLANT EQUIPMENT OUT-OF-SERVICE ANALYSIS FOR QUAD CITIES NUCLEAR POWER STATION UNITS 1 AND 2

NEDC-31449 B.1 INTRODUCTION

\.

The purpose of this section is to present the results of a study ,

prepared for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2 to establish the licensing bases for continued plant operation with a single

\.

failure of the following equipment:

(1) Last-stage feedwater heater string ,

(2) One safety / relief valve (S/RV) -

(3) One recirculation pump loop The ability to operate at full power or at a reduced power level throughout an entire or partial reactor fuel cycle with one of the abcee equipment out-of-service would be of significant economic value.

Operational flexibility and capacity factor are increased because the plant can continue to operate until the out-of-service equipment can be repaired or until the next convenient outage occurs. Assuming only a single equipment failure, the resultant operating MCPR limits are applicable to the expanded power / flow map (Figure A.1-1) with the exception of the increased core flow region.

To establish the technical specification operating limits for each of the equipment assumed out-of-service, one or more of the following concerns need to be addressed:

(1) Core-wide transient performance (2) Containment dynamic loads (3) Feedwater nozzle fatigue (4) Loss of-Coolant accident (LOCA)

B.1-1 s .

NEDC-31449 B.2 FEEDWATER HEATER OUT-OF-SERVICE This analysis justifies operation with 100 degree F reduction in feedwater temperature in the expanded operating docain with the exception of the ICF region (Figure A.1-1). The feedwater heater out-of-service =

analysis supports a contingency operating mode allowing continued operation with reduced feedwater temperature over a full fuel cycle.

l i l

Operation with feedwater heater out-of-service is similar to operation wi th FFWTR , except that (1) the duration of operation can be longer and, (2) operation can occur at any time during the cycle. Therefore, trar:1ent analyses are performed to develop a cycle independent operating MCPR limit applicable to plant operation at the reduced feedwater temperature. In ]

addition, the impact on other safety analyses and design bases such as containment, LOCA and feedwater nozzle fatigue is evaluated.

B.2.1 ABNORMAL TRANSIENTS EVALUATION Operating with a feedwater heater out-of service could potentially impact plant transient analysis as follows:

(1) The direct effect of reduced feedwater temperature is to increase the core inlet subcooling which in turn affects the core pressurization rate and reactivity during postulated J transients R

(2) The potential change in core inlet conditions can affect the 5-reactor nuclear parameters sucli as the power shape and core {E void fraction. Changes in these parameters can affect the plant responses for the transient events analyzed.

To esttblish cycle-independent operating limits for reactor operation with a feedwater heater out-of-service, a bounding - end-of cycle (EOC) exposure condition is used to develop nuclear input to the transient

=

B.2-1 4

"E

NEDC-31449 analysis model. The severity of the transient results is strongly

' dependent on the effectiveness of the contaol rod scram action. For this reason, the EOC bounding exposure condition assumes a rnore top-peaked axial power distribution than the nominal power shape, thus yielding a poorer

t. cram response. Analyses with this bounding power shape result in a ACPR 0.04 worse than similar analyses with the nominal power shape and, therefore, should provide reasonable conservatism for operating MCPR limits in future cycles.

' [ load rejection without The limiting transient event from Reference 5, bypass (LRNBP)) was analyzed with the feedwater heater out-of-service.

With reduced feedwater temperature, the LRNBP will be less severe because of thd reduced core steaming rate . and lower initial void fraction. The feedwater controller failure (FWCF) event, although not lirsiting in terms of oCTR, has the potential to become more severe with a feedwater heatet out-of-service and could become the lirnit ing transient. Therefore, both LRNBP and FWCF were reanalyzed with the bounding power shape at 1004 core power /100% core flow with a 100 degree F feedwater temperature reduction.

The results for the above transient analyses are presented in Table B.2-1 and time histories of the key parameters are shown in Figures B.2-1 through B.2-4. Table B.2-2 presents the 6CPR results for the events analyzed. To account for plant operation in the regien above the rated rod line, the above transients were also evaluated at 100% power /87% flow and found to be less limiting than the rated condition case. As expected, the LRNBP event with feedwater heater out-ofiservice is less limiting than the base case with feedwater heater operable. The opposite trend is observed for the FWCF event. However, with feedwate r heater out-of-service, both the LRNBP and FWCF event yield similar operating limits for this operating condition.

Therefore, the operating limits associated with feedwater heater out-of-service are 1.37 (Option A) and 1.32 (Option B)

B.2-2

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9 .. 'NEDC 31449 .1'

.t i +

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The<above lin'tts are valii[for ,all future cycin ak QCNPS UhEts, x1 a@ -

2 ' lauding' current GE foal designs pro vidid Thati 7.

t i i U (1) The standar 1 reloaA licehsing LRNBP and ITNBP events. r,edu? t h operating 1Loit MCPE' i,0L6PR) values less r.han or/ ecud . tu ! {37 +

and 1.32 for Optjo.T aand B, respectively.- ' y.

,f f .)

,g (2) The standard reload' licensing analysis NI event les f.tx ir -

1 ' l, . ..

OU1CPR less^then 1.34 and 1.29 for Option A aniUOption B, '

s y,o!,< ,

resMetively. 3his conditOb'it"iaposed to as.iure 'that the ACPR {

, pre a'd between the FWCF and the' ' LRNBP/TTNBP observed for the

^

bound!$g power shape is main;a1ned.

'i These two criteria are not expectedL Oe r ic t M.L. since tney' -

.l l represent conservative limit; 4t.A ned with the tquading, poqr :.dape . The i- current cycle 10- reload licensirig analysiu Jaef$nnee Si results are\.t1A

"' included in Table ' n.2.2 and confirm the 0.04 b CPR urb! ,n established it' fnn

, , , v ,

bounding basis. '

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. B.2.2 LOCA ANALYSIS .

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Operation"' with. a feeNate r tester out of-serv.iv; irnreases the .

, ,j s3.bcooling < i n the ,downcomer And at tyd core int e r. , Thie.%ould c@u' an ,

,j

.o ,

increare in Slowdawn flow out ' of a tdlated break ' in tb recirculi.tl.o.-

line during the early stages of a LOCA. [This increase in subco Jjnt pd in

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3 blowdown f1pw out of the break pai causo seve ral small ef fects on tM ECCS ,

< ?nrmal hydraulic ena!nis:

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,(1) The decay in core fSlet flow could occur more rapid 1v :i*>jsb of -

t .

-;- > rhe higher inventory lors and cadse ;

a slightly earlier fuel '

.cledding dryout (boiling transitfon). , f 4-(2) The core uncovery time:could change slightly becaurr ol'two

. competing effects: wp1 / mass loss out of fthe bruak, Mt More mass 1w -

in the core (due to lower initial cr r? void fraction) . , <

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- v' <2 NEDC-31449

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.j'l 4, j

Hi (3);Ihe sensible heat in the vessel and internals could be slightly

p V ' .' , ' faser which could affect the overall system depressurization 4y ;

.g / .

response.

The MCA analysis of these effects demonstrated the insensitivity to

, ihmger.in feedwater temperature of this magnitude. The net impact of void W frhetion and other effects results in a LOCA PCT change of less than 10 deg,rees F, uhich is insignificant when compared to the conservatism in the i

, sterrlaid LOC A analysis . $

J

'M N 7 ,

i

)

I .; B,?.3 FEEDb'ATER N0ZZLE FATIGUE ANALYSIS . i s l

) l s

1An evaluation of the effect of a feedwater heater out-of-service on fedwater nozzle fatigue was performed for QCNPS Units 1 and '2 with the

, follovi'ag a'ssumptions : 1 1

.2' ,,

1 - . t

. 4

,(1) An 18 month fuel cycle. g 1

(i) Aisuming a feedwater heater out-of-service for various lengths of j 9:ime at the end of a fuel cycle, which causes a 100 degree F drop n

j j -

in the feedwater temperature.

.;i

A relationship was determined for incremental fatigue damage as a I imu; ion cf time spent at the lower feedwater temperature. As part of the

A.6), a feedwater nozzle d exumd4 operating domain analysis (Section

- I fitigue s ti. dy was performed for QCNPS operation with ICF/FWTR. Both ICf/FWTh an? feedwater heater out-of service operation involve the same phwie) f phenomena and fatigue mechanism to the feedwater no::le.

There'oro, the methods and assumptions previously described in Section tj/

4 A.6.1 r unain applicable to the feedwater heater out-of service condition.

- Also, t.he basis for the results of this analysis are identical to the 5 1CF/FfVTR analysis described in Section A.6.2.

c Table B.2-3 shows the modification of the indices for feedwater heater The percent.sge s.f time spent in the feedwater ll ,

out-of-service operation.

f 't'

j B.2-4 3?

4 4

_ _ = - _ _ _ - - - _ - _ _ - _ - _ _ _ _ _ _ _ _ _ _ . _ _ . __ _ _ . _ _ _ _

[:

t l

NEDC-31449 i

heater out of service mode was subtracted from the percentage of time spent in normal operation. A relationship of incremental fatigue damage as a

! function of time spent in the feedwater heater out of-service mode was developed from these results.

The analysis documented in Reference 10 indicated that refurbishment of the thermal sleeve seals after 11 years is necessary to keep the 40-year total fatigue usage (system cycling plus rapid cycling) below a value of 1.0. Keeping the refurbishment schedule constant for the analysis, the 40-year total fatigue usage was calculated as shown in Table B.2-4 The fatigue damage per cycle for FfVTR operation is conservatively estimated by taking the difference between the FFWTR fatigue and s.he normal operation fatigue and dividing that quantity by the number of cycles in 40 years.

If operation with a feedwater heater out-of-service were implemented for every cycle during the plant's life, the 40-year total fatigue usage factor would be greater than 1.0, assuming that the seals were replaced after every 11 years. Satisfactory fatigue usage can be achieved by reducing the refurbishment interval to 8 years, as noted in Table B.2-4 The impact on seal refurbishment was less severe for this mode of operation I than for ICF/FTWTR.

B.2.4 CONTAINMENT LOADS EVALUATION The containment analysis results for ICF with FFWTR (Section A.7) are applicable to the feedwater heater out of-service analysis because the resulting core subcooling increase following ICF and FFWTR bounds the case for a feedwater heater out of-service. Given that FFWTR and feedwater heater out-of-service both result in a temperature reduction of 100 degrees F, the addition of ICF will increase the core inlet subcooling and yield conservative results if applied to the feedwater heater out-of-service analysis.

Therefore, the containment evaluation performed to support ICF with ,

l.

FFWTR is applicable to this feedwater heater out-of-service analysis.

i B.2-5

NEDC 31449 Table B.2-1 TRANSIENT ANALYSIS RESULTS FOR QCNPS AT 100P/100F FEEDWATER HEATER OUT-OF-SERVICE Maximum Core Maximum Ave. Surface Maximum Maximum Transient Neutron Flux Heat Flux Dome Pressure Vessel Ires.

Description (9 NBR) (% NBR) (psig) (p s i c_)

LRNBP w/ FbH" 529.3 121.7 1193 1223 l LPSBP w/o Fkm 404.5 117.9 1174 1204 FWCF w/ TbH 267.8 118.0 1109 1141 FWCF w/o Fkm 224.0 120.7 1086 1117 l

FbH ; feedwater heater I

1 B.2-6 1

l l

I

_ _ _______ ..-. - a

NEDC-31449 Table B.2 2 OPERATING MCPR RESULTS FOR QCNPS AT 100P/100F FEEDWATER HEATER OUT-OF SERVICE Exposure '

Transient 1 - Bounding .

Description 2 - EOC10 A_CfB" OMCPR OMCPR b R LRNBP W/WH 1 0.23 1.37 1.32 WCF W/WH 1 0.18 1.31 1.26 i

LRNBP W/0 WH 1 0.21 1.34 1.29 FWCF W/0 WH 1 0.21 1.34 1.29 LRNBP W/WH 2 0.19 1.33 1.28 l

FWCF W/ FWH 2 0.14 1.27 1.22 Uncorrected for Option A and Option B.

FWH: feedwater heater.

B.2-7

NEDC 31449 Table B.2-4 N0ZZLE FATIGUE USAGE FOR A 11-YEAR SEAL REFURBISHMENT PERIOD FOR FEEDWATER HEATER OUT-OF-SERVICE (LOCATION 1) 18 Month Cycle FWH00S Operation for 1314 Hours Normal Fach Cycle Operation (10S ver year) 40-Year Total Fatigue Usage 0.5735 1.5305 i Additional Usage Due to FWHOOS --

0.9570 Additional Usage l

Per Cycle -- 0.0360 Note: The total 40-year usage factor for 1314 hours0.0152 days 0.365 hours 0.00217 weeks 4.99977e-4 months of feedwater heater out-of service operation during every cycle can be kept below 1.0 by refurbishing the seals after 8 years (at location D).

1 B.2-9

NEDC-31449 Table B.2-3 FEEDWATER DUTY MAP INDICES ADDED FOR FEEDWATER HEATER OUT-OF-SERVICE Cycle Feedwater Feedwater Region A Time Index Flow Temperature Temperature Year

(% rated) ( F) ( F) (%)

20 100 225 546 10 Notes: (1) The feedwater temperature is based on a lower value of a +3% variation on the nominal temperature.

(2) One case was arbitrarily analyzed:

104 time per year - 1314 hours0.0152 days 0.365 hours 0.00217 weeks 4.99977e-4 months / cycle.

(3) The time spent at this mode of operation (10%) was subtracted from normal operation (i.e., index 1 - 65.20 55.20).

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NEDC-31469 B.3 ONE SAFETY / RELIEF VALVE OUT-OF-SERVICE This analysis provides the technical bases for operation of QCNPS Units 1 and 2 with one safety / relief valve (S/RV) out-of-service. In particular, the accident and transient considerations for operation with one S/RV out-of-service are presented.

B.3.1 ABNORMAL TRANSIENT EVALUATIONS Operation of QCNPS Units 1 and 2 with one S/RV out of-service could affect the change in critical power ratio (ACPR) in the event of an abnormal operating transient. The decrease in relief capacity could lead to higher pressures during a pressurization event, which could lead to a larger ACPR. The failure of one S/RV could also result in a higher peak vessel pressure, thereby reducing the margin to the ASME upset code limit for a pressure vessel.

l The transients which yields the most limiting ACPR for QCNPS Unit 1 )

Cyc le 10 is the load rejection without bypass (LRNBP). This event was i

reanalyzed with the most limiting relief valve disabled (i.e., the relief valve with the lowest setpoint). The valve setpoints used in this analysis are given in Table B.3-1. For the overpressure criteria, the main 4 steamline isolation valve (MSIV) closure transient with high flux scram was l I

analyzed with the lowest setpoint spring safety valve out-of-service.

B.3.1.1 Irrnac t on Delta CPR Analysis )

l 1

The IRNBP transient was analyzed using the ODYN computer program with full relief capacity (as a base case) and with the lowest setpoint relief I

i N

I l

B.3-1

14EDC 31449 valve out-of-service. The results showed no change in the ACPR due.to the reduced relief capacity. Plots of typical transient responses are shown in Figures B.3-1 and B.3 2.

From the transient responses, it can be seen that the peak neutron !

flux occurs about 0.7 second before the relief valves open for both analyzed cases. Because QCNPS has two relief valves with the same low setpoint, disabling one relief valve affects only the pressure relief capacity and not the time of valve initiation. Because the neutron flux is decreasing rapidly at the time when the relief valves open, a change in the overall relief capacity will not affect the CPR result.

i In summary, with one relief valve out of-service there is negligible impact on the MCPR limit. The ACPR for this operating condition will be bounded by reload licensing calculations. This conclusion is valid for !

current General Electric fuel types and analysis methods as applied to QCNPS Units 1 and 2.

l l

l B.3.1.2 Irrna c t on overpressure CriterQ.

I Reference 9 documents the results of GE sensitivity studies which show the effect of a S/RV . out-of-service is a peak pressure increase of less )

than 20 psi.

The adequacy of the S/RV capacity based on ASME code requirements is demonstrated by the MSIV closure transient with high flux scram and without credit for relief valve operation. With the lowest setpoint spring safety )

valve out-of-service, this transient event still shows an adequate margin l of 54 psi to the ASME upset code limit of 1375 psig. The time response of j key variables for this transient is shown in Figure B.3-3.

i B.3-2

l:

l IX l NEDC 31449 i

B.3.2 LOCA ANALYSIS If the out-of service valve has an automatic depressurization function (ADS), there can be a potential impact on the calculated peak cladding 2

temperature (PCT) for small break sizes of less than approximately 0.2 ft With a worst case postulated single failure of the High Pressure Core f' Injection (HPCI) System, a small effect may be seen because the small break transient is dominated by the time required to depressurize the reactor to the operating pressure of the low pressure Emergency Core Cooling System .

l (ECCS).

The effect of one ADS valve out-of-service was accounted for in the Reference 8 LOCA analysis because only four of the five ADS valves were used for the break spectrum analysis. The results showed much lower PCT values for small breaks than for the large breaks, which are the most i

i limiting LOCA cases for this plant. For the large breaks cases, the reactor vessel is rapidly depressurized prior to the actuation of the ADS.

Therefore, one ADS valve out-of-service has no impact on the calculated PCT.

i l.

1 i

B.3-3 L_____________________

NEDC-31449 l

i Table B.3-1 VALVE SETPOINTS USED FOR TRANSIENT ANALYSIS Setroint (psig) Type No.

1105 + 1% RV 1 1125 + 1% RV 2 1125 + 1% S/RV 1 One S/RV assumed out of-service pl f

B.3-4

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B.3-7

(- B.4 ONE RECIRCULATION PUMP OUT-OF-SERVICE From a plant availability / outage planning standpoint, the capability of operating at reduced power with a single recirculation loop is highly desirable in the event that maintenance of a recirculation pump or other

! components render one loop inoperable. To justify the single-loop operation (SLO), accidents and abnormal operational transients associated ,

with power operation were reviewed for the single loop case with one pump in operation. This SLO analysis was previously performed for QCNPS Units 1 and 2 and documented in Reference 2.

To support the introduction of the GE8x8EB fuel design and the additional operating domain above the rated rod line (Figure A.1-1), the issues addressed by the referenced SLO analysis are reviewed to ensure their applicabilities with these operational changes. In addition, the impact on safety limits for SLO in the region above the rated rod line with one safety / relief valve (S/RV) out-of-service is also addressed here.

B.4.1 MINIMUM CPR FUEL C1. ADDING INTEGRITY SAFETY LIMIT Except for the total core flow and the traversing in-core probe (TIP) measurements, the uncertainties used in the statistical analyses to determine the fuel cladding integrity MCPR safety limit do not depend on whether coolant flow is provided by one or two recirculation pumps.

Since the core flow uncertainty and the TIP noise uncertainty are not affected by the proposed operational changes (i.e., the GE8x8EB fuel design and operation above the rated rod line with or without one S/RV out-of-service), the conclusions shown in the referenced SLO analysis are still applicable.

B.4-1

NEDC-31449 B.4.2 MINIMUM CPR OPERATANG LIMIT d

The referenced SLO analysis (Reference 2, Paragraph 6.B.3) demonstrated I that, within the normal operating domain, the consequences of abnormal l

l operation transients from one-loop operation will be considerably less severe than those analyzed for a two-loop operation mode. Operation with one recirculation loop results in a maximum power output significantly below (by 20 to 30%) that which is attainable with a two pump operation.

Thus, fer pressurization, flow decrease and cold water increase transients, the results for two-pump operation cases bound both the thermal and overpressure consequences of one-loop operation. The introduction of GE8x8EB fuel in the core is not expected to alter the above conclusion.

The observed transient performance trend (one-pump case bounded by two-pump case) remains applicable to the QCNPS cores with the new GE8x8EB fuel design.

The failure of the S/RV with the lowest setpoint for two-pump operation was previously shown to have no impact on the MCPR operatin5 limits and the vessel overpressure criteria (Section B.3) . For SLO, the same conclusion remains applicable because the peak neutron flux would j I

still occur before any S/RV actuotion. In addition, the lower initial j power level for SLO mode would reduce the severity of the vessel peak l pressure compared with the two-pump case.

I J

The above conclusions are also applicable for plant operation in the region above the rated rod line.

B.4.3 STABILITY ANALYSIS 1

1 The introduction of GE8x8EB fuel design to QCNPS Units 1 and 2 cores l J

will result in an insignificant impact on the core and channel decay ratio for reactor operation with one recirculation loop. Therefore, the conclusions stated in the Reference 2 regarding this subject are still applicable to QCNPS Units 1 and 2. 4 1

B.4-2 1

l l

NEDC-31449 B.4.4 LOCA ANALYSES The LOCA analysis documented in Reference 2, Paragraph 6.B.6 imposed a MAPLHGR reduction factor of 0.84 to GE 8x8 retrofit fuel type, l Based on analysis experiences in using the SAFER /GESTR LOCA evaluation l models, the CE8x8EB fuel design has been shown to have larger margins to the PCT limit than the 8x8R and BP/P8x8R fuel types. This is primarily due to the decrease in initial stored energy of the GE6x8EB fuel attributable to the increased initial pressurization level. The Reference 8 analysis concluded, with SAFER /GESTR, no MAPLHGR multiplier is required for SLO at QCNPS Units 1 and 2.

i I

l 1

I I

B.4-3

EEDC-31449 i

I REFERENCES 8 l

I4

,4

1. " General Electric Boiling Water Reactor Extended Load Line Limit 'j Analysis for Quad Cities Nuclear Power Station Unit 1 Cycle 7 and Unit 2 Cycle 6", General Electric Company, July 1982 (NEDO-22192).

2. "Dresden Nuclear Power Station Units 2 and 3 and Quad Cities Nuclear i i

Power Station Units 1 and 2 Single Loop Operation", General Electric ?

Con:pany , December 1980 (NEDO 24807) . I

3. R.A. Bolger., Commonwealth Edison Co., Letter to B.C. Rusche, Director

.I of Nuclear Reactor Regulation, USNRC, "QC-2 Proposed Amendment to Facility License No. DPR-30, Docket No. 50-265", June 11, 1976.

4. R.E. Engel, General Electric Company, Letter to T.A. Ippolito, USNRC, "End of Cycle Coastdown Analyzed with ODYN/TASC", September 1, 1981.

l S. " General Electric Boiling Water Reactor Supplemental Reload Licensing j

, Submittal for Quad Cities Nuclear Power Station Unit 1 Reload 9",

I l General Electric Company, June 1987. )

1

6. R.L. Gridley, General Electric Company, Letter to D.G. Eisenhut, USNRC, " Review of Low Core Flow Effects on LOCA Analysis for Operating !

BWRs", May 8, 1978.

7. D.G. Eisenhut, USNRC, Letter to R.L. Gridley, General Electric i Company, enclosing " Safety Evaluation Report Revision of Previously imposed MAPLHGR (ECCS LOCA) Restrictions for BWRs at Less Than Rated Core Flow", May 19 1978, i

8. " Quad Cities Nuclear Power Station Units 1 and 2 SAFER /GESTR Loss-of-I Coolant Accident Analysis", General Electric Company, June 1987 (NEDC-31345P). l I

i R-1

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NEDC-31449 l

9. " General Electric Standard Application for Reactor Fuel", General Electric Company, May 1986 (NEDE-24011-P-A-8-US).

10. G.L. Stevens, B.J. Cheek, " Economic Generation Control Fatigue Usage Evaluation for Dresden Units 2 and 3 and Quad Cities Unit 1 and 2",

General Electric Company, August 1984 (AE-78-0884).

11. " Mark I Containment Program Plant-Unique Load Definition Report Quad Cities Station Units 1 and 2", General Electric Company, April 1982 ,

(NEDO 24567 Rev.2). .

12. " Quad Cities Un',s 1 and 2 Plant-Unique Analysis Report", Nutech !

Report No. C^'.-02 039, May 1983.

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13. " ARTS Improvement Program Analysis for Quad Cities Nuclear Power Station Units 1 and 2", General Electric Company, June 1987 (NEDC-31448P).

l ACKNOWLEDGEMENTS The analyses included in this report were performed by the combined efforts of many individual contributors, inclucing:

P.F. Billig, T.G. Dunlap, G.D. Galloway,

! M.O. Lenz, L.K. Liu, H.X. Nghiem, l

G.L. Stevens, T.P. Shannon, J. Wallach and K. Jarhanian.

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t' m GENERAL ELECTRIC COMPANY AFFIDAV1T I, B. Wolfe, being duly sworn, depose and state as follows:

1. I am Vice President and Chief Scientist of Nuclear Systems Technology Operation, General Electric Company, and have been delegated the function of reviewing the information described in paragraph 2 which is sought to be withheld and have been authorized to apply for its withholding.

2. The information sought to be withheld is contained in " Quad Cities Nuclear Power Station Units 1 & 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31345P, June 1987.

3. In designating material as proprietary, General Electric utilizes the definition of proprietary information and trade secrets set forth in the American Law Institute's Restatement of Torts, Section 757. This definition provides:

"A trade secret may consist of any formula, pattern, device or compilation of information which is used in one's business and which gives him an opportunity to obtain an advantage over competitors who do not know or use it.... A substantial element of secrecy must exist, so that, except by the use of improper means, there would be difficulty in acquiring information.... Some factors to be considered in determining whether given information is one's trade secret are: (1) the extent to which the information is known outside of his business; (2) the extent to which it is known by employees and others involved in his business; (3) the extent of measures taken by him to guard the secrecy of the information; (4) the ,

value of the information to him and to his competitors; (5) {

the amount of effort or money expanded by him in developing !

the information; (6) the ease or difficulty with the which the information could be properly acquired or duplicated by others."

4. Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method or apparatus where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information consisting of supporting data and analyses, including test data, relative to a process, method or apparatus, the application of which provide a competitive economic advantage, e.g., by optimization or improved marketability; 1

__.._______________.______-_D

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c. 'Information which if used by a competitor, would reduce his :

. expenditure. of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality or, licensing of a similar products d, Information which reveals cost or price ~ -information, production capacities,-budget levels or commercial strategies of General Electric, its customers or suppliers;

e. Information which reveals aspects of past, present or future General Electric customer-funded development plans and programs of potential commercial ~value to General Electric;

f. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection;

g. Information which General Electric must treat as proprietary according to agreements with other parties.

5. Initial approval of proprietary treatment of a document is typically made by the Subsection manager of the originating

, omponent, who is most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge.

Access to such documents within the Company is limited on a "need to know" basis and such documents are clearly' identified as .

proprietary. 8

6. The procedure,for approval of external release of such a document typically requires review by the Subsection Manager, Project manager, Principal Scientist or other equivalent authority, by the Subsection Manager of the cognizant Marketing function (or delegate) and by the Legal Operation for technical content, competitive effect and determination of the accuracy of the proprietary designation in accordance with the standards enumerated above. Disclosures outside General Electric are. generally limited to . regulatory , bodies, customers and potential customers and their agents, suppliers and licensees then only with appropriate protection by applicable regulatory provisions or proprietary l agreements.

7. The document mentioned in paragraph 2 above has been evaluated in accordance with the above criteria and procedures and has been j found to contain information which is proprietary and which is j customarily held in' confidence by General Electric.

8. The document mentioned paragraph 2 above is classified as i proprietary because it contains important input parameters and analysis results of the SAFER /GESTR-LOCA analysis methodology, as i j well as details of current fuel designs that are not available to

j. other parties.

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.9.- -The information to the best of my knowledge and belief has

-consistently been held in confidence by the General Electric Company, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties have been made pursuant to regulatory provisions of proprietary agreements which provide for maintenance of the information in confidence.

10. Public disclosure of . the information sought to be withheld is likely to cause substantial harm to the competitive position of the General Electric Company and deprive or reduce the availability of profit making opportunities because it would provide other parties, including competitors, with valuable information regarding analysis inputs and rasults using the SAFER /GESTR-LOCA methodology, which were obtained at considerable cost to the General Electric Company.

In addition, this document contains details, not available to other parties, of the thermal limits for the current General Electric multiple-lattice fuel designs. The lattice-specific thermal limits are proprietary to the General Electric Company since valuable design information could be derived by competitors with knowledge of the lattice specific thermal limits. Although it is necessary to include some thermal limits information in plant Technical Specifications, thus making this information available to j competitors, it is important to limit the amount of information available to the maximum extent possible. i STATE OF CALIFORNIA )

COUNTY OF SANTA CLARA ) ss:

B. Wolfe, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

1 Exe:uted at San Jose, Ca'.ifornia, this /$ day of du 4<a.I , 1987. 1 e

B. Wolfe F General Electric Company Subscribed and sworn before me this y of[ 1987.

>~ x' &

NOTARY YUBLIC, STATE OF CALIFORNIA i 4

OFFICIAL SEAL" f MARY L KENDALL '

@ NOTARY PUBUC SANTA CLARA COUNIT My cett41. expires MAR 13, 1339

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ML20235C487

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SUMMARY

SUMMARY

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