8605020194 860430PDR ADOCK 05000397P PDR

NEDC-31107DRF L12-00737

Class IIMarch 1986

~ 'AC 310

SAFETY REVIEW OF

WPPSS NUCLEAR PROJECT NO. 2

AT CORE FLOW CONDITIONS ABOVE RATED FLOW THROUGHOUT CYCLE 1

AND FINAL FEEDWATER TEMPERATURE REDUCTION

S. WolfTechnical Project Engineer

Approved:A.E. Rogers, ManagerPlant Performance Engineering

Approved:R. Art gas, ManagerLicensing Services

g tP

NEDC-31107

IMPORTANT NOTICE REGARDING

CONTENTS OF THIS REPORT

Please Read Carefully

The only undertakings of General Electric Company respecting informa-tion in this document are contained in the contract between Washington

Public Power Supply System (MPPSS) and General Electric Company, as

identified in the purchase order for this report and nothing containedin this document shall be construed as changing the contract. The use

of this information by anyone other than WPPSS or for any purposeother than that for which it is intended, is not authorized; and withrespect to any unauthorized use, Gereral Electric Company makes no

representation or warranty, and assumes no liability as to thecompleteness, accuracy, or usefulness of the information contained inthis document.

N EDC-31107

CONTENTS

~Pa e

ABSTRACT

ACKNOWLEDGMENTS

l. INTRODUCTION AND SUMMARY

2. SAFETY ANALYSIS2.1 Abnormal Operational Transients

2.1.1 Limiting Transients2.1.2 Overpressurization Analysis2.1.3 Rod Withdrawal Error

2.2 Fuel Loading Error2.3 Rod Drop Accident2.4 Loss-of-Coolant Accident Analysis2.5 Thermal-Hydraulic Stability

3. MECHANICAL EVALUATION OF REACTOR INTERNALS ANDFUEL ASSEMBLY3. 1 Loads Evaluation3.2 Loads Impact

3.2. 1 Reactor Internals3.2.2 Fuel Assemblies

4. FLOW-INDUCED VIBRATION

5. FEEDWATER NOZZLE AND FEEDWATER SPARGER FATIGUE USAGE5. 1 Method and Assumption5.2 Feedwater Nozzle Fatigue5.3 Feedwater Sparger Fatigue

6. CONTAINMENT ANALYSIS

7 .. OPERATING LIMITATIONS

8. REFERENCES

vi

2-12-12-12-22-32-32-32-32-4

3-1

3-13-23-23-2

4-1

5-15-15-25-3

6-1

7-1

8-1

NEDC-31107

TABLES

TABLE

2-1

2-2

2-3

5-1

5-2

Title

Core-Wide Transient Analysis Results at ICF and/or FFWTR

Required MCPR Operating Limits at ICF and/or FFWTR

Overpressurization Analysis Results

Feedwater Nozzle Fatigue Usage

Feedwater Sparger Fatigue Usage

~Pa e

2-6

2-7

2-8

5-5

ILLUSTRATIONS

~Fi ure

Operating Map

Title Paae

1-3

2-1

2 2

2-3

2-4

2-5

Generator Load Rejection with Bypass Failure at 104.2/Power, 106% Flow and Normal Feedwater Temperature

Generator Load Rejection with Bypass Failure at 104.5%Power, 106K Flow and Reduced Feedwater Temperature

Feedwater Controller Failure, Maximum Demand, at 104.2~Power, 106% Flow and Normal Feedwater Temperature

Feedwater Controller Failure, Maximum Demand, at 104.5/Power, 106'A Flow and Reduced Feedwater Temperature

MSIV Closure, Flux Scram, at 104.2~ Power, 106Ã Flow andNormal Feedwater Temperature

2-9

2-10

2-11

2-12

2-16

~NEDC-31107

ABSTRACT

A safety evaluation has been performed to show that Washington Public

Power Supply System (WPPSS) Nuclear Project No. 2 (WNP-2 or Hanford 2)

can increase core flow to operate within the region of the operating

map bounded by the line between 100% power, 100/ core flow (100,100)

and 100% power, 106% core flow (100, 106) throughout Cycle 1. WNP-2,

after reaching End-of-Cycle 1 (EOCl) exposure (depletion of full-powerreactivity under standard feedwater conditions) with all control rods

out, can continue to operate in the region of the operating map

bounded by the 106% core flow line between 100/ power and the

cavitation interlock power with or without the last-stage feedwater

heaters valved out-of-service (Final Feedwater Temperature Reduction

of < 65'F at rated power).

The minimum critical power ratio (MCPR) operating limits will be

changed from the values established by the Final Safety Analysis

Report licensing submittal, to the appropriate values (Table 2-2) forIncreased Core Flow (ICF) and Final Feedwater Temperature Reduction

(FFWTR) operating conditions. All other operating limits established

in the Cycle 1 licensing basis have been found to be bounding for the

ICF and FFWTR operations as defined above.

~ ~

~ ~

NEDC-31107

ACKNOWLEDGMENTS

The analyses reported in this repor t were performed by thecombined efforts of many individual contributors, including:

C. S.

M. L.

M. 0.G. L.

Chen, G. G. Chen, D. A. Copinger, S. K. Dhar,Gensterblum, J. K. Garrett, B. Haaberg, B. H. Koepke,Lenz, H. X. Nghiem, J. R. Pallette, R. Seetharaman,Stevens, M. W. Thompson, S. Wolf and C. T. Young

NEDC-31107

1. INTRODUCTION AND SUMMARY

This evaluation supports the operation of the Washington Public Power

Supply System (WPPSS) Nuclear Project No. 2 (WNP-2 or Hanford 2), within theincreased core flow ( ICF) region of the operating map as illustrated in Figure1-1. This report presents the results of a safety evaluation for operation withICF for Cycle 1 [up to and including End-of-Cycle 1 (EOC1) exposure]. The

safety evaluation also covers operation for exposure beyond standard EOC1* withICF and/or last-stage feedwater heaters valved out, followed by a naturalreactivity coastdown bounded by 106 core flow. Final feedwater temperaturereduction (FFWTR) from a normal rated power temperature of 420'F to a feedwatertemperature of 355'F at 100% power and reactivity coastdown to a minimum

feedwater temperature of approximately 321'F (about 65/ power) should occur onlyat the end-of-cycle. The extended region of operation with increased core flowfollowed by FFWTR at end-of-cycle is bounded by the ICF region marked on theoperating map in Figure 1-1.

In order to evaluate operation with ICF and FFWTR, the limiting abnormal

operational transients reported in the Final Safety Analysis Report (FSAR),Reference 1, for rated flow operation were reevaluated at EOC1 at 106% core flowwith and without FFWTR. The loss-of-coolant accident (LOCA), fuel loading erroraccident, rod drop accident, and rod withdrawal error event were alsoreevaluated for increased core flow operation.'hese events were alsoreevaluated for end-of-cycle operation with ICF and the last-stage feedwaterheaters valved out.

*EOC1 is defined as the core average exposure at which there is no longersufficient reactivity to achieve rated thermal power with rated core flow, allcontrol rods withdrawn (beyond Rod Position 24), all feedwater heaters inservice and equilibrium xenon.

1-1

k~ ~

NEO C-31107

In addition, the effect of the increased pressure differences (due to the

increased core flow) on the reactor internals components, fuel channels, and

fuel bundles was also analyzed to show that the design limits will not be

exceeded. The effect of the increased core flow rate on the flow-induced

vibration response of the reactor internals was also evaluated to ensure thatthe response is within acceptable limits. The thermal-hydraulic stability was

evaluated for ICF/FFWTR operation, and the increase in the feedwater nozzle and

feedwater sparger usage factors due to the feedwater temperature reduction was

determined. The impact of feedwater temperature reduction and increased core

flow on the containment LOCA response was also analyzed.

The results of the safety evaluation show that the current technicalspecifications with incorporation of the MCPR limits of Table 2-2 are adequate

to preclude the violation of any safety limits during operation of WNP-2 withinthe increased core flow region of the operating map as illustrated in Figure l-lfor Cycle 1 and for exposures beyond EOC1 with the conditions assumed in the

analysis. The LCPRs and the minimum critical power ratio (MCPR) operatinglimits for plant operation are given in Tables 2-1 and 2-2. The EOCl Option A

and Option 8 MCPR limits (Reference 1) will be increased to the appropriatevalues as shown in Table 2-2.

1-2

0 '

~ ~

~ ~

130

120

FLOW CONTROL PUMPVALVEPOSITION SPEED

CURVE )% OF FULLSTROKE) )% RATED)APRM STP SCRAM

APRM ROD BLOCK

INCAEASEDCORE FLOWREGION

110

TO

Iw 80

01

23468789

10

0.68 W~ s 61'%

0100

014212838466888

NATCIRC

I 26

ROD BLOCK MONITOR

~ 100

RATED ROD LINE

ALLOWABLEOPERATINGDOMAINr2

80%

0'108,'I 00)

mtDI

C)

40

30

10

0.68 Ws+ 42%

0.88' 40%

CAVITATIONLINES:

JET PUMP NOZZLE

JET PUMP SUCTION

RECIRC PUMP

60%

CAVITATIONINTERLOCK

10 30 40 50 60 70 80 90 100 ''IO

CORE F LOW (psrcsntl

Figure l-l. Operating t)ap

NEDC-31107

2. SAFETY ANALYSIS

2.1 ABNORMAL OPERATIONAL TRANSIENTS

2.1.1 Limitin Transients

The limiting abnormal operational transients analyzed in the Cycle 1 FSAR

licensing submittal (Reference 1) were reevaluated for increased core flow

and/or FFMTR.

Nuclear transient data for 104.5% power*, 106% core flow (104.5, 106) withand without the last-stage feedwater heaters out were developed based on the

Haling method at rated power for EOC1. The nuclear data was then used to

analyze the load rejection with bypass failure (LRNBP) event and the feedwater

controller failure to maximum demand (FWCF) event at the (104.5, 106)

conditions.

The results of the transient analyses are presented in Tables 2-1 and 2-2

with the limiting transient results previously submitted in the FSAR licensingsubmittal (Reference 1). The transient performance responses are presented inFigures 2-1 through 2-4. The results demonstrate that the hCPR values and the

critical power ratio operating limits for the LRNBP and FMCF events increase

compared with the corresponding FSAR values. However, the FSAR licensingsubmittal (Reference 1) OLCPR = 1.24 for either Option A or Option B based on

the rod withdrawal error (RWE) transient is bounding for both the LRNBP and FWCF

events for ICF with or without FFWTR. The current evaluation of the RWEevent's

presented in Section 2. 1.3.

*All transients were analyzed using 105% steam flow. The power level corre-sponding to this condition will vary from 104.5X to 104.2%, depending onwhether final feedwater heaters are in service. The 104.5 power level providesa 5X steam flow margin to the 100% power operating conditions to simulateeventual stretch power operation, similar to the original FSAR analyses.

2-1

~ ~ eNEOC-31107

Oecreasing the power from the 1005 rated condition along the 106% core flowline will result in an increase in transient sCPR for some events. Thisincrease is less than the increase in operating CPR due to the power-decrease,and, henCe, such operation will not result in violation of the safety limit MCPR

due to a transient (Reference 2, p. 2-12).

2.1.2 Over ressurization Anal sis

The limiting transient for ASME code overpressurization analysis, mainsteam isolation valve (MSIV) closure with flux scram (direct scram failure), was

evaluated for the extended EOC1 conditions with ICF without FFWTR (Table 2-3 and

Figure 2-5). For this evaluation ICF without FFWTR is more severe than ICF withFFWTR. The ICF for the LRNBP event results in a less severe overpressuretransient than MSIV closure with flux scram. The overpressurization analysis(Table 2-3) for the ICF region produced a peak vessel pressure of 1264 psig,which is below the upset code limit of 1375 psig and is, therefore, acceptable.

2.1.3 Rod Withdrawal Error

The rod withdrawal error transient was evaluated under ICF and/or FFWTR

conditions. When ICF is employed, the rod block monitor (RBM) setpoint (whichis flow biased) increases, giving an unacceptably high MCPR limit. Thus, theRBM should be clipped at flows greater than 1005 of rated so that the aCPR

values (Reference 1) determined wi thout ICF apply.

2.2 FUEL LOAOING ERROR

This event is not adversely affected by the increased core flow mode ofoperation with the last-stage feedwater heaters removed from service. The

impact of ICF and/or FFWTR on aCPR is expected to be very small compared withthe margin to the OLCPR. Thus, the FSAR bCPR would not be affected by thisevent under ICF and/or FFWTR conditions.

2-2

NEDC-31107

2.3 ROD DROP ACCIDENT

WNP-2 uses banked position withdrawal sequence (BPWS) for control rod

movement. Control Rod Drop Accident (CRDA) results from BPWS plants have been

statistically analyzed. The results show that, in all cases, the peak fuelenthalpy in an RDS would be much less than the corresponding design limit even

with a maximum incremental rod worth corresponding to 9N probability at the 95K

confidence level. Based on these results, it was proposed to the US NRC, and

subsequently found acceptable, to delete the CRDA from the standard GE-BWR

reload package for the BPWS plants (Reference 2, Section S.2.5.1.3 (1), Page

2-53). Hence, the CRDA is not specifically analyzed for WNP-2.

2.4 LOSS-OF-COOLANT ACCIDENT (LOCA) ANALYSIS

LOCA analysis performed for WNP-2 shows that operation with ICF withoutFFWTR bounds operation with ICF and FFWTR.

The effect of increased core flow on LOCA analyses is not significantbecause the parameters which most strongly affect the calculated peak claddingtemperature (PCT), i.e., high power node boiling transition time and coreref looding time, have been shown to be relatively insensitive to increased core

flow.

Results of the LOCA analysis performed show that the PCT for ICF increases

by less than O'F throughout the break spectrum compared to the rated core flowcondition.

Therefore, it is concluded that the LOCA PCT is acceptable and that thecurrent maximum average planar linear heat generation rates (MAPLHGRs) for WNP-2

are applicable for ICF.

2-3

0 ~ ~

,I

NEOC-31107

2. 5 TMERMAL-HYORAULIC STABILITY

'he

General, Electric Company has established stability criteria-to demonstrate

compliande to requirements set forth in 10CFR50 Appendix A, General Design Criteria(GDC). 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 fueldesign limits being exceeded. Furthermore, the onset of power oscillations'forwhich corrective actions are necessary is reliably and readily detected and sup-

„ pressed by operator actions and/or automatic system functions. The stabilitycompliance of all licensed GE BWR fuel designs including those fuels containedin the General Electric Standard Application for Reactor Fuel (GESTAR, Reference

2) is demonstrated on. a generic basi's in Reference 3 (for operation in thenormal as well as the extended operating domain with ICF and FFWTR). The NRC

has reviewed and approved this in Reference 4; therefore, a specific analysisfor each cycle is not required. The WNP-2 Cycle 1 core contains licensed GE BWR

initial core and, hence, the generic evaluation in Reference 3 is applicable toWNP-2.

For operation in the ICF region, the stability margin (defined by the coredecay ratio) is increased as flow increases for a given power. ICF operation isbounded by the fuel integrity analyses in Reference 3.

Similarly, operation in the FFWTR mode is bounded by the fuel integrityanalyses in Reference 3. In general, the effect of reduced feedwater tempera-

ture results in a higher initial CPR which yields even larger margins than those, reported in Reference 3. The fuel integrity analyses are independent of the

stability margin, since the reactor is already assumed to be in limit cycleoscillations. Reference 3 also demonstrates that even if neutron flux limitcycle oscillations did occur just below the neutron flux scram setpoint, fueldesign limits are not exceeded for those GE BWR fuel designs contained inGeneral Electric Standard Application for Reactor Fuel (GESTAR, Reference 2).These evaluations demonstrate that substantial thermal/mechanical marg'in .is

available for the GE BWR fuel designs even in the unlikely event of very largeoscillations.

2-4

t

NEDC-31107

To provide assurance that acceptable plant performance is achieved during

operation in the least stable region of the power/flow map, as well as duringall plant maneuvering and operating states, a generic set of operating recom-

mendations has been developed as set forth in Reference 5 and communicated toall 6E BWRs. These recoranendations instruct the operator on how to reliablydetect and suppress limit cycle neutron flux oscillations should they occur.

The recommendations were developed to conservatively, bound the expected per-formance of all current product lines and are applicable to operation with FFWTR

(feedwater temperature of approximately 355'F at rated power).

2-5

Table 2-1

CORE-WIDE TRANSIENT ANALYSIS RESULTS AT ICF AND/OR FFMTR

TransientDescription

FigureNumber

Power(I NBR)

Flow(I NBR)

RatedFeedwater

TemperatureReduction

('F)

HaximumNeutron

F lux('X NBR)

HaximumCore Ave. Haximum

Surface DomeHeat Flux Press

('l Initial) (psig)

HaximumYessel

Press(psig)

Hax imumSteamL inePress

(psig) aCPR

LRNBPLRNBPLRNBPFMCFFMCFFMCF

Ref. 1

2.12.2

Ref. 1

2.32,4

104.4104.2104.5104.4104.2104.5

100106106100106106

00

-6500

65

236.4252.4243.2154. 3163. 7174,7

107.8108. 8108.8108. 7

109. I113. 9

117311721160114811451138

120212031191117711771166

1168 0.091168 0. 11

1157 0. 111140 0.081141 <0. 131135 0. 13

a. oa rebec on w ypass failure, FMCF * feedwater controller failure to maximum demand,

b. Reduction of feedwater temperature from nominal rated feedwater temperature (420"F) and at ratedconditions.

c. ACPR based on initial CPR which yields HCPR = 1.06; uncorrected for Options A and B.

NEDC-31107

Table 2-2

REQUIRED HCPR OPERATING LIMITS AT ICF AND/OR FFWTR

aTransientDescription

InitialCore

Power(X NBR)

InitialCoreFlow

(X NBR) aCPR OLCPRA

OLCPR8

LRNBPf(FSAR)LRNBPFWCF (FSAR)FMCFg

RWE (FSAR)

104.4104.2104.4104.5

104.4

100106100106

100

0.090.110.080.13

aCPR

0.18

1.201.221.191.24

OLCPR

1.24

1.121.141.161.21

a. LRNBP = Load rejection with bypass failure, FWCF = feedwater controllerfailure at maximum demand, RME = rod withdrawal error.

b. ODYN results without adjustment factors, based on initial CPR which yieldsan MCPR = 1.06.

c. Includes Option A adjustment factors.

d. Includes Option 8 adjustmentfactors.'.

Option A and 8 adjustment factors are specified in the NRC safetyevaluation report on ODYN (NEDO-24154 and NEDE-24154P).

f. For load rejection with bypass failure, ICF w/o FFWTR bounds ICF withFFWTR.

g. For feedwater controller failure to maximum flow demand, ICF with FFWTRbounds ICF w/o FFWTR.

h.. Required OLCPR using either Option A or Option 8 adjustment factor withrod block monitor of 106Ã at rated flow

2-7

Ie

EEOC-31107

Table 2-3

OVERPRESSURIZATION ANALYSIS RESULTS

Transient

InitialPower

(%)

InitialFlow

(~)

MaximumVessel

Pressure(psig) Figure No.

MSIV Closure - Flux Scram

(FSAR)

104.3 100 1266 Reference 1

MSIV Closure - Flux Scram

( ICF w/o FFWTR)

104. 2 106 1264 Figure 2-5

2-8

0

', )i

l

150.

I NEUTRON2 PEAK FUEl3 AVL SUAF(k FfE044IL(5 VESSEL 5

LUXCENTER TEMP

CF. HEAT FLUXFLOH

EAH FLOH

I VESSEL P ES AISE (PS I)2 STH Llt(E PRES RISE (PSI)3 TURBINE f RES RISE (PSI)<I COAE It(L I SUB (BTU/LB)5 RELIEF V LVE FLOH (PC'f)6 TURB STE 4 FLOH (PCT)

g 100.

(5 50. 0.

0.0 20 6.

TIME (SEC) .

-I00.0 2. 6.

TIME (SEC)8.

200.

I LEVEL ( I H-AEF-SEP-SKIRT2 4 A SENS 0 LEVEL(INCHES)3 N 4 SENS 0 LEVEL(INCHES)

L TVD~Xtf)5 OAIVE FL 4 I (PCT)

I VOIO AEA2 DOPPLEA3 SCRAH RE

TIVITTEACT IV ITTCTI VITT

IOO.

0. -I.

I

-I00.0 LJ. 6.

TIME (SEC)8.

-20. 2. 3.

TIME (SEC)

I

Figure 2-1. Generator Load Rejection with Bypass Failure at 104.2X Power, 106% Flow andNormal Feedwater Temperature

l

~ ~ ~

150.

100.I

I NEUTRON2 PEAK FUE(3 AVF SURF(0 FEEE)HATE(5 VESSEL S

LUXCENTER TEHP

CE HFAT FLUXFLOW

EA(( FLOH

200.

100.

I VESSEL P ES RISE (PS I)2 STN LINE PRES RISE (PSI)3 T()RBINE I RES f((SE (PSI)4 Ri:LIEF V(LVCKOH (PCT)5 RELIEF V LVE FLOH (PCT)6 TURB S'(E H FLOH (PCT)

3

0.

0.0 20 6.

TINE (SEC)8.

-100,0. 20 6.

TI((E (SEC)8.

I LEVEL(I, 2 H A SENS

3 N R SENS

5 ORIVE FL

H-REF-SEP-SKIRT0 LEVEL(INCHES)0 LEVEL(INCHES)fVCG~T)W I (PCT)

I VOID RER2 OOPPLER3 SCRAH RE

TIVITTEACT IV ITTCTIVI7Y

C7DlI

C)

100. 0.

0.

-100.0. 2. 6.

TINE (SEC)

-20. 2. 3.

TINE (SEC)4.

Figure 2.2. Generator Load Rejection With Bypass Failure at 104.5X Power, 106% Flow andReduced Feedwater Temperature of 65 F at Rated Power.

150.

I NEUTRON I0 Pffl( FUEI3 AVE SURf'(

l.C(:OMiI.lVESSEL 5

LUXCENTER YEHP

CF. HEAT FLUXFLON

EAH FLON

I VESSEL P2 STH LINC3 YUAOfNE fi) C(iAE INA5 '( IEF V

6 U Sf( f

ES RISE (PS I)PACS f)ISE (PSI)AE~ fl(Sf.'PSI )

I SUA lBlllfLB)LVE FLO)l(PCT)H FLON (PC'f)

a 100.I

Ki 50.

W

100.

0.

0. 10. )5.TIHE (SEC)

20.-100.

0. 10. 15.TIHE (SEC)

20.

150,

I 'LEVEL ( IN H-AEF-SEP-SK lRT2 N A SENQO LEVEL(INCHES)3 N A SEN 0 LEVEL(INCHES)~l5%1NL ~LAN (((T)5 BYPASS 5 EAH FLO)I(PCT)

I VOIO BE2 OOPPLER3 AAH RE

TIVITTEACTI V ITCTI VITTC

mnI

C)

100. 0.

0.~

~ 5. 10. 15.TINE (SEC)

-20. 5. 10. 15.

TINE (SEC)20.

Figure 2-3. Feedwater Controller Failure, Maximum Demand at 104.2X Power, 106% Flowand Normal Feedwater Temperature

f

I IJF(ITA()JI I

2 ('I.t(K I (If'I:) (JVI p()f(I J

JI I I I (I(JAII I

5 VLSSLL s

((IrI,f,Nfff1 (FHP

~c(: Nfl(I f(,UXI I OH

L'(JH f LOH

2(JO.

I Vl Ssf.l. f'fc 'i(It I INI'( l(i(IH)lll Iil LJ)I<1 INII!~ (0) It.l V(( IJ)ftn Sll(

f") Afsf lf "I)I I<t i AISI (I'Sl)Jg,s .I(/sf (f".ilII:I(IFJ ((J)(VI,A)I Vi' LOJJ JI 0 r')

H I I.t)JI tl'(.I)

a IM.UJ

5hII so. 0.

0.0. 5. 10. 15.

TIME (SECI20.

-100.0. 10. 15.

TIME (SEC)20.

150.

I LEVEL(I2 H A SO(St3 N A SENcltl CdhE TN(.iS 0)I'ASS S

H-AFF-SEP-SKIAT0 LI.VFL()NCHCS)0 I.FVEI (INC((FS)T Fiof( (ICI)EAH f LOH(f'Cf)

I VOID AE TIVITT2 OOPPLEA I'ACIJVTlT3 SCAAH AF MVITTWDI CTivTTV

0.

0.0. 5.

Figure 2-4.

-I

ILJCIUICC

-21510. 20. 0. 5. 10. 15.

TIHE ISEC) TIHE (SEC)

Feedwater Controller Failure, Maximum Demand, at 104.5% Power, 106K Flowand Reduced Feedwater Temperature of 65 F at Rated Power

20.

I NEUTRON f2 PFAK FUEI3 AVF. SUnf(4 FEEOHAI):(S VESSEL S

LUXCENTER TEMP

CE MCAT FLUXFLOH

EAM FLOH

300.

I VESSEL P2 SIM LINE3 S(lFETV Vf<I CORE l(H.S CORE AV(:.6 1URO SIE

FS AISE (PSI)PRES RISE (f'SI)LVE FLOH (PC')I STD (fi(U/LO)YOIO FAAC (PCT)H FLOH (PCT)

ci 100. 200.

100.

0.0 20 LI, 6.

TIME (SEC)

0.0 20 6.

TIME (SEC)B.

I LEVEL(l2 H A SENS3NA SENS

5 OAIVE FL

H-REF-SEP-SHIRT0 LEVEL(INCHES)D LEVEL( INCHES)TVWHHKT)H I (PCI)

I VOIO AERLEA

3 SC RE

TIVITTEACT I V IT T

CTIVITY

0.

0.

-IM.0 20 4. 6.

TIME (SEC)8.

-20 2. 3.

TIME lSEC)4,

Figure 2-5. HSIV Closure, Flux Scram, at 104.2X Power, 106% Flow and Normal Feedwater Temperature

NEDC-31107

3. MECHANICAL EVALUATION OF REACTOR INTERNALS AND FUEL ASSEMBLY

3.1 LOADS EVALUATION

Evaluations were performed to determine bounding acoustic and flow-inducedloads, reactor internal pressure difference loads and fuel-support loads for ICF

and/or FFWTR operation.

Acoustic loads are lateral loads on the vessel internals that result frompropagation of the decompression wave created by a sudden recirculation suction

,line break. The acoustic loading on vessel internals is proportional to thetotal pressure wave amplitude in the vessel recirculation outlet nozzle. The

total pressure amplitude is the sum of the initial pressure subcooling plus theexperimentally determined pressure undershoot below saturation pressure. FFWTR

operation increases the expected acoustic loads because this downcomer sub-cooling increases and, therefore, the total pressure wave amplitude increases.The high velocity flow patterns in the downcomer resulting from a recirculationsuction 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 ofthe break. The additional subcooling in the downcomer resulting from FFWTR

operation leads to an increase in the critical flow and, therefore, to a corres-ponding increase in the flow-induced loads. The reactor internals most impactedby acoustic and flow-induced loads are the shroud, shroud support and jet pumps.

A reactor internals pressure difference analysis was performed for the ICF

region. The increased reactor internal pressure differences across the reactorinternals were generated for the maximum core flow at normal, upset, and faultedconditions for the reactor internal impact evaluation.

Fuel-support loads and fuel bundle lift for WNP-2 were evaluated based on

results from probabilistic fuel lift analyses pe) formed at 106% of rated coreflow following the procedures of Reference 6. Fuel-support loads and fuelbundle liftwere evaluated for upset, faulted and fatigue load combinations.

3-1

NEOC-31107

It was shown that the fuel bundle lift is a small fraction of the applicabledesign criteria (established in the NRC Safety Evaluation Report to Reference 6)for the faulted event.

3.2 LOADS IMPACT

'3.2.1 Reactor Internals

The reactor internals most affected by ICF and/or FFWTR operation are thecore plate, shroud support, shroud, top guide, shroud head, steam dryer, controlrod guide tube, control rod drive housing and jet pump. These and othercomponents were evaluated using the bounding loads, discussed in Section 3. 1,under normal, upset, emergency and faulted conditions. It is concluded that thestresses produced in these and other components are within the allowable designlimits given in the Final Safety Analysis Report (Chapter 3 and 4) or the ASME

Code, Section III, Subsection NG.

3.2.2 Fuel Assemblies

The fuel assemblies, including fuel bundles and channels, were evaluatedfor increased core flow operation considering the effects of loads discussed inSection 3. 1 under normal, upset, faulted and fatigue load combinations. Resultsof the evaluation demonstrate that the fuel assemblies are adequate to withstandICF effects to 1064 rated flow.

The fuel channels were also evaluated under normal, upset, emergency and

faulted conditions for increased core flow (Reference 7). The channel wallpressure diffewentials were found to be within the allowable design values .

3-2

It

NEDC-31107

'.

fLOW-INDUCED VIBRATION

To ensure that the flow-induced vibration response of the reactor internalsis acceptable, a single reactor of each product line and size undergoes an

extensive vibration test during initial plant startup. After analyzing theresults of such tests and assuring that all responses fall within acceptablelimits of the established criteria,'he reactor is classified as a valid proto-type in accordance with Regulatory Guide 1.20. All other reactors of the same

product line and size undergo a less rigorous confirmatory test to assuresimilarity to the base test. The acceptance criteria used for vibration assess-ment is based on a maximum allowable alternating stress intensity of 10,000 psi.

The increased core flow vibration analysis was performed by analyzing thestartup test vibration data for-the valid prototype plant (BWR/5-251 Tokai 2).Based on the results of the analysis and a review of the test data, the reactorinternals response to flow-induced vibration is expected to be within acceptablelimits for plant operation in the ICF region (region bounded as shown on thepower flow map, Figure 1-1).

4-1

NEDC-31107

5. FEEDWATER NOZZLE AND FEEDWATER SPARGER FATIGUE USAGE

5.1 METHOD AND ASSUMPTIONS

The fatigue experienced by the feedwater nozzle and feedwater spargerresults from two phenomena: system cycling and rapid cycling. System cycling iscaused by major temperature changes associated with system transients. The

system cycle stresses are based on limiting cycles that use the maximum temper-ature range possible to show expected worst conditions. These transients areidentified on thermal cycle diagrams. Thermal stresses due to these transientsare calculated by determining inner and outer metal surface temperatures usingfinite element analysis. Fatigue usage is determined by dividing the number ofdesign cycles for each transient by the number of allowable cycles for each

stress calculated. Cumulative system fatigue usage is determined by summing allof the respective transient fatigue usage factors.

Rapid cycling is caused by small, high frequency temperature fluctuationscaused by mixing of relatively colder nozzle annulus water with the reactorcoolant. The colder water impinging the nozzle bore originates from theboundary layer of colder water formed by heat transfer through the thermalsleeve. The mixing region extends from the feedwater nozzle surface region tothe feedwater sparger surface; therefore, rapid cycling applies to both of thesecomponents. Once thermal stress due to rapid cycling-is determined, fatigueusage is calculated and the results are added to the cumulative system cyclingusage factor to obtain the total usage factor.

The introduction of FFWTR will cause a change in calculated rapid cyclingfatigue only.. This is because the system transient is very mild (smalltemperature change and relatively long duration) and is bounded by the originaldesign basis thermal stress analysis. General Electric has developedstandardized rapid 'cycling duty maps for each BWR plant that cover the designbasis rapid cycles in the same manner that thermal cycle diagrams cover thedesign basis thermal transients (system cycling). The methodology used todevelop the duty maps is based on the results of extensive testing of feedwater

5-1

I

NEDC-31107

nozzles by General Electric. FFWTR is analyzed by modifying the design cyclesin order to gauge its effect on fatigue usage. The reduced feedwater

temperature will tend to increase fatigue usage due to an increase ia thermal

stress.

An evaluation of the effect of FFWTR on the feedwater nozzle and feedwater

'parger fatigue was performed for the following conditions:

As the last step in a 12-month fuel cycle, FFWTR to a feedwater temperature

of 355'F (65'F reduction from nominal rated feedwater temperature) at ratedpower for 18 days was followed by a 3X per week coastdown over 12 weeks toa final power of 65K. The coastdown was initiated from a reduced feedwater

temperature of 55'F. The associated feedwater temperature at the end ofthe coastdown was 321'F.

The analysis was performed by simulating the feedwater temperature reduc-

tion during the coastdown period in four equal increments. An appropriatemaximum feedwater flow rate was assumed for each of the four increments to

provide conservative results.

5.2 FEEDWATER NOZZLE FATIGUE

The original stress analysis of the feedwater nozzle showed that the

maximum system cycling fatigue usage factor for the nozzle blend radius regionwas 0.6524 for emergency and faulted conditions (Reference 8). The usage factorfor rapid cycling using the design basis (unmodified) duty map is 0.2047,

providing a total 40-year usage factor of 0.8571. The usage factor for rapidcycling including FFWTR operation is 0.2796 providing a total 40-year usage

factor of 0.9320. This result is based on FFMTR operation during every 12 month

cycle for the life of the plant. This is equivalent to 0.0019 fatigue damage

per cycle of FFMTR operation. The 40-year total usage factor remains below the

AStlE Code Limit of 1.0 with FFWTR operation and is thus considered acceptable.

The results are summarized in Table 5-1.

5-2

HEDC-31107

The results of this analysis are based on cycling correlations developed

during testing of various nozzle configurations. The fatigue results"areintended to be a conservative best-estimate for the expected plant operation. A

more accurate e'valuation 'of fatigue usage could be made by considering actual

plant performance.

5.3 FEEOWATER SPARGER FATIGUE

Feedwater sparger fatigue usage is calculated in the same manner as

feedwater nozzle fatigue usage. However, since the feedwater sparger is not an

ASME Boiler and Pressure Vessel Class I Code component, a fatigue analysis was

not originally required. WNP-2 has a welded single thermal sleeve design which

does not allow leakage of feedwater flow to occur at the safe end as do otherthermal sleeve designs. This leakage flow is the primary contributor to spargerfatigue usage. Therefore, sparger fatigue usage is much less affected by

changes in feedater flow and temperarture for the welded single sleeve design.The sparger is made from stainless steel material which is less susceptible tohigh cycle fatigue than the low alloy steel of the nozzle as evidenced by thedifferences in their respective fatigue curves. Small changes in flat the (highcycle) portion of the fatigue curve can cause very significant changes infatigue usage (i.e., a relatively small

changers

in stress can cause a verysignificant change in the allowable number of cycles). Thus, it becomes evidentthat the sparger fatigue damage is much less severe than nozzle fatigue damage

during feedwater condition changes like FFWTR for the welded single sleeve

design. Since the nozzle fatigue damage is so low (0.0019 per cycle), the

sparger damage will be insignificant and, therefore, can be neglected.

5-3

HEDC-31107

Table 5-1

FEEDWATER NOZZLE FATIGUE USAGE

Condition

Fatigue Usage

Due to FFWTR

(Over Normal Operation)*Per Cycle

40-Year FatigueUsage Factor*

Normal Operation 0.8571

FFWTR 0.0019 0.9320

*The total fatigue usage factor includes a system cycling usage factor of0.6524 due to emergency and faulted conditions as given in the originalstress analysis of the nozzle (Reference 8).

5-4

NEDC-31107

6. CONTAINMENT ANALYSIS

The impact of feedwater temperature reduction and increased core flowoperation on the containment LOCA response was evaluated.

The results show that the containment LOCA response for ICF operation alone

is bounded by the corresponding FSAR results (Reference I). Operation withFFWTR causes a slight increase in the initial drywell pressurization rate over

the rate reported in the FSAR. The calculated peak values for drywell pressureand wetwell pressure under ICF and/or FFWTR are bounded by the correspondingvalues for the FSAR (Chapter 6) conditions. The peak value for drywell floordifferential presure (download) is bounded by the appropriate design limit of 25

psid. All other containment parameters are bounded by the results reported inthe FSAR.

The LOCA-related pool swell, condensation oscillation and chugging loads

were evaluated at the worst power/flow conditions during ICF/FFWTR operation.Pool boundary pressure load during pool swell under ICF/FFWTR conditions exceeds

the load calculated based on FSAR conditions by less than 2.2W. However, thisload and all other pool swell loads are bounded by the appropriate design loads.The condensation oscillation and chugging loads with ICF/FFWTR conditions are

also bounded by the appropriate design loads.

6-1

NEDC»31107

7. OPERATING L IMITATION

Restrictions/limitations which are unique to ICF/FFWTR operation are

identified below.

7. 1 FEEDWATER HEATERS

The FFWTR analyses have assumed that the last-stage feedwater heater isvalved out-of-service in each string of feedwater heaters (Final Feedwater

Temperature Reduction < 65'F at rated power) for exposures beyond EOC1. This

may be done at any time after EOC1 whether or not ICF is used. This is done tohelp increase or maintaine rated power after all'control rods have been with-drawn at EOC1 and was accounted for in the safety analyses in Sections 2.

7.2 OPERATING NAP

The allowable operating domain of the normal power-flow map has been

increased to allow operation at lOOX power up to 106% core flow. The minimum

allowable power in this increased core flow region is bounded by the jet pump

cavitation protection interlock as shown in Figure 1-1. The increased core flowreactor internal pressure differences and fuel bundle lift calculations were

analyzed and are applicable only for reactor operation within the ICF regionshown on the power flow map in Figure 1-1.

7.3 MCPR OPERATING LIMITS

Required NCPR operating limits applicable to ICF/FFWTR have been determined

for WNP-2 as given in Table 2-2.

7.4 Kf FACTOR

For core flows greater than or equal to rated core flow, the Kf factor isequal to 1.0.

7-1

NEOC-31107

7.5 CONTROL ROOS

The safety evaluation for ICF with FFWTR operation was performed with theassumption of an all-rods-out condition. This is defined as the condition ofoperation in which all control rods are fully withdrawn from the core orinserted no deeper than rod position 24.

7-2

AEDC-31107

8. REFERENCES

1. "Final Safety Analysis Report, WPPSS Nuclear Project No. 2,"as revised through Amendment 35, November 1984.

2. "General Electric Standard Application for Reactor Fuel (Supplement forUnited States)," August 1985 (NEDE-24011-P-A-7-US, as amended).

3. "Compliance of the General Electric Boiling Water Reactor Fuel Designs toStability Licensing Criteria," October 1984 (NEDE-22277-P-1).

4. Letter, C. 0. Thomas (NRC) to H. C. Pfefferlen (GE), "Acceptance forReferencing of Licensing Topical Report NEDE-24011, Revision 6, Amendment

8, Thermal Hydraulic Stability Amendment to GESTAR II," April 24, 1985.

5. "BWR Core Thermal Hydraulic Stability," SIL No. 380 Revision 1, February10, 1984.

6. "BWR Fuel Assembly Evaluation of Combined SSE and LOCA Loadings," LicensingTopical Report, Amendment No. 3, October 1984 (NEDE-21175-3-P-A and

NEDO-21175-3-A) .

7. "BWR Fuel Channel Mechanical Design and Deflection," General ElectricCompany, September 1976 (NEDE-21354-P).

8. "Hanford 2 - 251 BWR-5 Stress Report for Feedwater Nozzle," Section E4,Contract 72-2647, Chicago Bridge and Iron Nuclear Company, 1973.

8-1

N EDC-31107

DISTRIBUTION

'ailCode

R. J. Brandon

C. S. Chen

G. A. Deaver

S. S. Dua

T. D. Dunlap

E. C. Eckert

W. G. Edmonds(6)

J. K. GarrettD. A. Hamon

E. C. Hansen

G. V. Kumar (3)L. K. LiuW. Harquino

J. R. PalletteA. E. Rogers

R. Seetharaman

G. L. Stevens

J. T. Teng

H. W. Thompson

J. Wallach

S. Wolf (2)C. T. Young

NEBO Library (3)

779

147

743

769

155

763

WPPS

755

769

~ 156

770

743

763

763

763

769

747

769

156

775

763

269

528

l

NE DC-31107DR F L12-00737

CLASS IIMARCH 1986

DAC 310

SAFETY REYIEW OF WPPSSNUCLEAR PROJECT NO. 2

AT CORE FLOW CONDITIONSABOYE RATED FLOW THROUGHOUT CYCLE 1

AND FINAL FEEDWATER TEMPERATUREREDUCTION

S. WOLF

GENERAL '' ELECTRIC

NEDC-31107DRF L12-00737

Class IIMarch 1986

DAC 310

SAFETY REVIEW OF

WPPSS NUCLEAR PROJECT NO. 2

AT CORE FLOW CONDITIONS ABOVE RATED FLOW THROUGHOUT CYCLE 1

AND FINAL FEEDWATER TEMPERATURE REDUCTION

S. WolfTechnical Project Engineer

Approved:A.E. Rogers, ManagerPlant Performance Engineering

Approved:R. Art gas, ManagerLicensing Services

NUCLEAR ENERGY BUSINESS OPERATIONS ~ GENERAL ELECTRIC COMPANYSAN JOSE. CALIFORNIA95125

GENERAL e ELECTRIC

NEDC-31107

IMPORTANT NOTICE REGARDING

CONTENTS OF THIS REPORT

Please Read Carefully

The only undertakings nf General Electric Company respecting informa-

tion in this document are contained in the contract between Washington

Public Power Supply System (WPPSS) 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'PPSS or for any purpose

other than that for which it is intended, is not authorized; and with

respect to any unauthorized use,'eneral Electr'ic Company makes no

representation or warranty, and assumes no liability as to the

completeness, accuracy, or usefulness of the information contained in

this document.

NEDC-31107

CONTENTS

~Pa e

ABSTRACT

ACKNOWLEDGMENTS

l. INTRODUCTION AND SUMMARY

2. SAFETY ANALYSIS2.1 Abnormal Operational Transients

2.1.1 Limiting Transients2.1.2 Overpressurization Analysis2.1.3 Rod Withdrawal Error

2.2 Fuel Loading Error2.3 Rod Drop Accident2.4 Loss-of-Coolant Accident Analysis2.5 Thermal-Hydraulic Stability

3. MECHANICAL EVALUATION OF REACTOR INTERNALS ANDFUEL ASSEMBLY3.1 Loads Evaluation3.2 Loads Impact

3.2.1 Reactor Internals3.2.2 Fuel Assemblies

4. FLOW-INDUCED VIBRATION

5. FEEDWATER NOZZLE AND FEEDWATER SPARGER FATIGUE USAGE5.1 Method and Assumption5.2 Feedwater Nozzle Fatigue5.3 Feedwater Sparger Fatigue

6. CONTAINMENT ANALYSIS

7 .. OPERATING LIMITATIONS

8. REFERENCES

vi

2-12-12-12-22~32~32~32-32-4

3-1

3-1' 2

3~23-2

4-1

5-15-15-2'5-3

6-1

7-1

8-1

1 11

NEDC-31107

TABLES

TABLE

2-1

2-2

2-3

5-1

5-2

T it 1 e

Core-Wide Transient Analysis Results at ICF and/or FFWTR

Required MCPR Operating Limits at ICF and/or FFWTR

Overpressurization Analvsis Results

Feedwater Nozzle Fatigue Usage

Feedwater Sparger Fatigue Usage

~Pa e

2-6

2-7

2-8

5-5

ILLUSTRATIONS

~Fi ere

Operating Map

Title Paae

1-3

2-1

2-2

2-3

2-4

2-5

Generator Load Rejection with Bypass Failure at 104.2%Power, 106% Flow and Normal Feedwater Temperature

Generator Load Rejection with Bypass Failure at 104.5%Power, 106% Flow and Reduced Feedwater Temperature

Feedwater Controller Failure, Maximum Demand, at 104.2%Power, 106% Flow and Normal Feedwater Temperature

Feedwater Controller Failure, Maximum Demand, at 104.5%Power, 106% Flow and Reduced Feedwater Temperature

MSIV Closure, Flux Scram, at 104.2% Power, 106% Flow andNormal Feedwater Temperature

2-9

2-10

2-11

2-12

2.-16

NEDC-31107

ABSTRACT

A safety evaluation has been performed to show that Washington PublicPower Supply System (WPPSS) Nuclear Project No. 2 (WNP-2 or Hanford P)

can increase core flow to operate within the region of the operatingmap bounded by the line between 100% power, lOOX core flow ( 100, 100)and 100% power, 106'A core flow ( 100, 106) throughout Cycle 1. WNP-2,

after reaching End-of-Cycle 1 (EOCl) exposure (depletion of full-powerC

reactivity under standard feedwater conditions) with all control'odsout, can continue to operate in the region of the operating map

bounded by the 106% core flow line between 100'A power and thecavitation interlock power with or without the last-stage feedwaterheaters valved out-of-service (Final Feedwater Temperature Reductionof < 65'F at rated power).

The minimum critical power 'atio (MCPR) operating limits will be

changed from the values established by the Final Safety AnalysisReport licensing submit'tal, to the appropriate values (Table 2-2) forIncreased Core Flow (ICF) and Final Feedwater Temperature Reduction(FFWTR) operating conditions. All other operating limits establishedin the Cycle 1 licensino basis have been found to be bounding for theICF and FFWTR operations as defined above.

NEDC-31107

ACKNOWLEDGMENTS

The analyses reported in this report were performed by the

combined efforts of many individual contributors, including:

C. S.

M. L.

M. 0.

G. L.

Chen, G. G. Chen, 0. A. Copinger, S, K. Dhar,

Gensterblum, D. K. Garrett, B. Haaberg, B. H. Koepke,

Lenz, H. X. Nghiem, J. R. Pallette, R. Seetharaman,

Stevens, M. W. Thompson, S. Wolf and C. T. Young

NE DC-31107

1. INTRODUCTION AND SUMMARY

This evaluation supports the operation of the Washington Public Power

Supply System (WPPSS) Nuclear Project No. 2 (WNP-2 or Hanford 2), within the

increased core flow (ICF) region of the operating map as illustrated in Figure1-1. This report presents the results of a safety evaluation. for operation withICF for Cycle 1 [up to and including End-of-Cycle 1 (EOC1) exposure]. The

safety evaluation also covers operation for exposure beyond standard EOC1* withICF and/or last-stage feedwater heaters valved out, followed by a naturalreactivity coastdown bounded by 106K core flow. Final feedwater temperature

reduction (FFWTR) from a normal rated power temperature of 420'F to a feedwater

temperature of 355'F at 100% power and reactivity coastdown to a minimum

feedwater temperature of approximately 321'F (about 65K power) should occur onlyat the end-of-cycle. The extended region of operation with increased core flowfollowed'by FFWTR at end-of-cycle is bounded by the ICF region marked on the

operating map in Figure 1-1.

In order to evaluate operation with ICF and FFWTR, the limiting abnormal

operational transients reported in the Final Safety Analysis Report (FSAR),

Reference 1, for rated flow operation were reevaluated at EOC1 at 106% core flowwith and without FFWTR. The loss-of-coolant accident (LOCA), fuel loading erroraccident, rod drop accident, and rod withdrawal error event were alsoreevaluated for increased core flow operation. These events 'were alsoreevaluated for end-of-cycle operation with ICF and the last-stage feedwater

heaters valved out.

k

*EOCl is defined as the core average exposure at which there is no longersufficient reactivity to achieve rated thermal power with rated core flow, allcontrol rods withdrawn (beyond Rod Position 24), all feedwater heaters inservice and equilibrium xenon.

1-1

NEDC-31107

In addition, the effect of the increased pressure differences (due to the

increased core flow) on the reactor internals components, fuel channels, and

fuel bundles was also analyzed to show that the design limits will not be

exceeded. The effect of the increased core flow rate on the flow-inducedvibration response of the reactor internals was also evaluated to ensure thatthe response is within acceptable limits. The thermal-hydraulic stability was

evaluated for ICF/FFMTR operation, and the increase in the feedwater nozzle and

feedwater sparger. usage factors due to the feedwater temperature reduction was

determined. The impact of feedwater temperature reduction and increased core

flow on the containment LOCA response was also analyzed.

The results of the safety evaluation show that the current technicalspecifications with incorporation of the NCPR limits of Table 2-2 are adequate

to preclude the violation of any safety limits during operation of WNP-2 withinthe increased core flow region of the operating map as illustrated in Figure 1-1

for Cycle 1 and for exposures beyond EOC1 with the conditions assumed in the

analysis. The ACPRs and the minimum critical power ratio (MCPR) operatinglimits for plant 'operation are given in Tables 2-1 and 2-2. The EOC1 Option A

and Option B MCPR limits (Reference 1) will be increased to the appropriatevalues as shown in Table 2-2.

1-2

130

120FLOW CONTROL PUMP

VALVEPOSITION SPEEDCURVE l% OF FULLSTROKEI OC RATED)

APRM STP SCRAM

APRM ROD BLOCK

INCREASEDCORE FLOW

13.6 REGION

110

70

I

6o

01

23466789

10

0.68 Wr t 611'

0100

014212838466688

NATCIRC ROD BLOCK MONITOR

~ 100 ~ ~~ RATED ROD LINE

10NC ALLOWABLE4 OPERATING

DOMAIN/2

80%

l108,100)

40

30

10

0.68 Wr + 42%

0.68 Wr i 40m

CAVITATIONLINES:

JET PUMP NOZZLE

JET PUMP SUCTION

RECIRC PUMP

CAVITATIONINTERLOCK

00 10 20 30 40 50 60 70 80 110

Figure 1-1. Operating Map

CORE FLOW (percent)

NEDC-31107

2. SAFETY ANALYSIS

2.1 ABNORMAL OPERATIONAL TRANSIENTS

2.1. 1 Limitin Trans ients

The limiting abnormal operational transients analyzed in the Cycle 1 FSAR

licensing submittal (Reference 1) were reevaluated for increased core flowand/or FFWTR.

Nuclear transient data for 104.5% power*, 106% core flow (104.5, 106) withand without the last-stage feedwater, heaters out were developed based on the

Haling method at rated power for EOC1. The nuclear data was then used toanalyze the load rejection with bypass failure (LRNBP) event and the feedwater

controller failure to maximum demand (FMCF) event at the (104.5, 106)

conditions.

The results of the transient analyses are presented in Tables 2-1 and 2-2

with the limiting transient results previously submitted in the FSAR licensingsubmittal (Reference 1). The transient performance responses are presented inFigures 2-1 through 2-,4. The results demonstrate that the ACPR values and thecritical power ratio operating limits for the LRNBP and FMCF events increasecompared with the corresponding FSAR values. However, the FSAR licensingsubmittal (Reference 1) OLCPR = 1.24 for either Option A or Option 8 based on

the rod withdrawal error (RWE) transient is bounding for both the LRNBP and FWCF

events for ICF with or without FFWTR. The current evaluation of the RME event

is presented in Section 2.1.3.

*All transients were analyzed using 105% steam flow. The power level corre-sponding to this condition will vary from 104.5% to 104.2~, depending onwhether final feedwater heaters are in service. The 104;5 power level providesa 5X steam flow margin to the 100% power operating conditions to simulateeventual stretch power operation, similar to the original FSAR analyses.

2-1

NEDC-31107

Decreasing the power from the 100Ã rated condition along the 106% core flowline will result in an increase in transient hCPR for some events. This

increase is less than the increase in operating CPR due to the power decrease,

and, hence, such operation will not result in violation of the safety limitMCPR'ue

to a transient (Reference 2, p. 2-12).

2.1.2 Over ressurization Anal sis

The limiting transient for ASME code overpressurization analysis, main

steam isolation valve (MSIV) closure with flux scram (direct scram failure), was

evaluated for the extended EOC1 conditions with ICF without FFWTR (Table 2-3 and

Figure 2-5). For this evaluation ICF without FFWTR is more severe than ICF withFFWTR. The ICF for the LRNBP event results in a less severe overpressuretransient than MSIV closure with flux scram. The overpressurization analysis(Table 2-3) for the ICF region produced a peak vessel pressure of 1264 psig,which is below the upset code limit of 1375 psig and is, therefore, acceptable.

2.1.3 Rod Withdrawal Er ror

The rod withdrawal error transient was evaluated under ICF and/or FFWTR

conditions. When ICF is employed, the rod block monitor (RBM) setpoint (which

is flow biased) increases, giving an unacceptably high MCPR limit. Thus, the

RBM should be clipped at flows greater than 100K of rated so that the ACPR

values (Reference 1) determined without ICF apply.

2.2 FUEL LOADING ERROR

This event is not adversely affected by the increased core flow mode ofoperation with the last-stage feedwater heaters removed from service. The

impact of ICF and/or FFWTR on hCPR is expected to be very small compared withthe margin to the OLCPR. Thus, the FSAR hCPR would not be affected by thisevent under ICF and/or FFWTR conditions.

2-2

NEDC-31107

2.3 ROD DROP ACCIDENT

WNP-2 uses banked position withdrawal sequence (BPWS) for control rod

movement. Control Rod Drop Accident (CRDA) results from BPWS plants have been

statistically analyzed. The results show that, in all cases, the peak fuelenthalpy in an RDS would be much less than the corresponding design limit even

with a maximum incremental rod worth corresponding to 95K probability at the 95K

confidence level. Based on these results, it was proposed to the US NRC, and

subsequently found acceptable, to delete the CRDA from the standard GE-BWR

reload package for the BPWS plants (Reference 2, Section S.2.5.1.3 (I), Page

2-53). Hence, the CRDA is not specifically analyzed for WNP-2.

2.4 LOSS-OF-COOLANT ACCIDENT (LOCA) ANALYSIS

LOCA analysis performed for WNP-2 shows that operation with ICF withoutFFWTR bounds operation with ICF and FFWTR.

The effect of increased core flow on LOCA analyses is not significantbecause the parameters which most strongly affect the calculated peak claddingtemperature (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 show that the PCT for ICF increases ,

by less than 5'F throughout the break spectrum compared to the rated core flowcondition.

E

Therefore, it is concluded that the LOCA PCT is acceptable and that thecurrent maximum average planar linear heat generation rates (MAPLHGRs) for WNP-2

are applicable for ICF.

2-3

NEDC-31107

2. 5 THERMAL-HYDRAULICSTABILITY

The General Electric Company has established stability criteria to demonstrate

compliance to requirements set forth in 10CFR50 Appendix A, General Design Criteria(GDC). 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 fueldesign limits being exceeded. Furthermore, the onset of power oscillations forwhich corrective actions are necessary is reliably and readily detected and sup-

pressed by operator actions and/or automatic system functions. The stabilitycompliance of all licensed GE BWR fuel designs including those fuels contained

in the General Electric Standard Application for Reactor Fuel (GESTAR, Reference

2) is demonstrated on a generic basis in Reference 3 (for operation in the

normal as well as the extended operating domain with ICF and FFWTR). The NRC

has reviewed and approved this in Reference 4; therefore, a specific analysisfor each cycle is not required. The WNP-2 Cycle 1 core contains licensed GE BWR

initial core and, hence, the generic evaluation in Reference 3 is applicable toWNP-2.

For operation in the ICF region, the stability margin (defined by the core

decay ratio) is increased as flow increases for a given power. ICF operation isbounded by the fuel integrity analyses in Reference 3.

Similarly, operation in the FFWTR mode is bounded by the fuel integrityanalyses in Reference 3. In general, the effect of reduced feedwater tempera-

ture results in a higher initial CPR which yields even larger margins than those

reported in Reference 3. The fuel integrity analyses are independent of the

stability margin, since the reactor is already assumed to be in limit cycleoscillations. Reference 3 also demonstrates that even if neutron flux limitcycle oscillations did occur just below the neutron flux scram setpoint, fueldesign limits are not exceeded for those GE B'WR fuel designs contained inGeneral Electric Standard Application for Reactor Fuel (GESTAR, Reference 2).These evaluations demonstrate that substantial thermallmechanicaI margin isavailable for the GE BWR fuel designs even in the unlikely event of very largeoscillations.

2-4

NEDC-31107

To provide assurance that acceptable plant performance is achieved duringoperation in the least stable region of the power/flow map, as well as duringall plant maneuvering and operating states, a generic set of operating recom-

mendations has been developed as set forth in Reference 5 and communicated toall GE BWRs. These recomnendations instruct the operator on how to reliablydetect and suppress limit cycle neutron flux oscillations should they occur.The recommendations were developed to conservatively bound the expected per-formance of all current product lines and are applicable to operation with FFWTR

(feedwater temperature of approximately 355'F at rated power).

2-5

l

Table 2-1

CORE-KIDE TRANSIENT ANALYSIS RESULTS AT ICF AND/OR FFMTR

TransientDescr1ption

FigureNumber

Power(X NBR)

RatedFeedwater

TemperatureFlow Reduction

(X NBR) (oF)

MaximumNeutronFlux

(X NBR)

HaximumCore Ave. Haximum

Surface OomeNeat Flux Press

('l Initia 1 ) (ps ig)

Hax1mumVesselPress

(psig)

'aximumSteamLinePress

(psig) aCPR

LRNBPLRNBPLRNBPFMCFFWCF

FMCF

Ref. 1

2.12.2

Ref. 1

2.32.4

104.4104.2104.5104.4104.2104.5

100106106100106106

00

6500

$ 5

236.4252.4243.2154.3163. 7174.7

107.8108.8108.8108.7109. 1

113.9

1173 12021172 1203-1160 11911148 11771145 11771138 1166

1168 0.091168 0.111157 0.111140 0.081141 <0. 131135 0.13

a. oa re ec on w ypass failure, FMCF * feedwater controller failure to maximum demand,

b. Reduct1on of feedwater temperature from nominal rated feedwater temperature (420'F) and at ratedconditions.

c. aCPR based on 1nitial CPR which yields HCPR = 1.06; uncorrected for Options A and B.

NEDC-31107

Table 2-2

REQUIRED MCPR OPERATING LIMITS AT ICF AND/OR FFWTR

aTransientDescription

Initial

CorePower

(X NBR)

InitialCoreFlow

(X NBR) aCPR OLCPRA

OLCPRB

LRNBPf(FSAR)LRNBPFWCF (FSAR)FWCFg

RWE (FSAR)

104.4104.2104.4104.5

104.4

100106100106

100

0.090.110.080.13

a,CPR

0;18

1.201'. 221.191.24

OLCPR

1.24

1.121.141.161.21

a. LRNBP = Load rejection with bypass failure, FWCF = feedwater controllerfailure at maximum demand, RWE = rod withdrawal error.b. ODYN results without adjustment factors, based on initial CPR which yields

an MCPR = 1.06.

c. Includes Option A'djustment factors.

d. Includes Option B adjustment factors.

e. Option A and B adjustment factors are specified in the NRC safetyevaluation report on ODYN (NEDO-24154 and NEDE-24154P).

f. For load rejection with bypass failure, ICF w/o FFWTR bounds ICF withFFWTR.

g. For feedwater controller failure to maximum flow demand, ICF with FFWTRbounds ICF w/o FFWTR.

h. Required OLCPR using either Option A or Option B adjustment factor withrod block monitor of 106% at rated flow

2-7

NEOC-31107

Table 2-3

OVERPRESSURI ZATION ANALYS!S RESULTS

Transient

InitialPower

(X)

InitialFlow(l)

MaximumVessel

Pressure(psig) Figure No.

MSIV Closure - Flux Scram.

(FSAR)

104.3 100 1266 Reference 1

MSIV Closure - Flux Scram

(ICF w/o FFWTR)

104.2 106 1264 Figure 2-5

2-8

150.

I NEUTRON LUX2 PEAK FUE CENTER TEHP3 AVE SURF CE HEAT FLUX4 FEEOHAIT) FLON5 VESSEL S EAH FLOH

200.

I VESSEL P2 SIH LINE3 TU88)NC4 CORE INL5 RELIEF V

6 TURB SIE

ES AISE (PSI)PRES RISE IPSI)RES RISE (PSI)I SUI) IBTU/LB)LVE FLOH (PCT)H FLOH IPCT)

p100.

lh

5

100.

0.

0.0. 2. 4. 6.

TIHE lSEC)8.

-100.0. 20 4. 6.

TINE (SEC)8.

200.

I LEVEL(12 II R SENS3 N A SENS

5:DRIVE FL

H-REF-SEP-SKIRTD LEVELI INCHES)D LEVEL(INCHES)

T)4 I (PCT)

I VOIO BE2 ODPPLEA3 SCRAH RE

TIVITTEACTIVITTCT IVITT

100. 0.

0.

-100.0 2. 4. 6.

TIHE )SEC)B.

200 2. 3.

TIHE (SEC)4.

Figure 2-1. Generator Load Rejection with Bypass Failure at 104.2X Power, 106K Flow andNormal Feedwater Temperature

1

150.

I NEUTRON2 PEAK FUEl3 AVE SuflFfW FEEOWA(f:f5 VESSEL 5

LUXCFNTER TEHP

CE t)EAT FLUXFLOW

EAH FLOW

200.

I VESSEL P ES AISE (PSI)2 STH LINE PRES RISE (PSI)3 T()ABIDE f A)S R/SE (P51)'I AELfEF Vf LVE FLOW (PCTl5 BELIEF V LVE FLOW (PCT)6 1UAB STEf H FLOW (PCT)

100.

0.

0.o. 2. 6.

TIHE (SEC)8.

-100.0 20 Q. 6.

TINE (SEC)8.

200.

I LEVEL(I2 W R SENS3 N A SENS

5 ORIVE FL

H-AEF-SEP-SKIRTD LEVE(.(INCHES)D LEVEL(INCHES)

I)ff I (PCT)

I VOIO REA TIVITT2 OOPPLEA EACT IV ITT3 SCAAH RE CTIVITT

100. 0.

0.

-100.0 2. 6.

TIHE (SEC)8.

200. 2. 3.

TIHE (SEC)

Figure 2.2. Generator Load Rejection With Bypass Failure at l04.5f. Power, 1061, Flow andReduced Feedwater Temperature of 65 F at Rated Power.

150.

I NEUTRON9 PEAK FUF)

AVE 5URFfFEI.OWAIE(VESSL'L S

LUXCENTER TEHP

CE f(EAT FLUXFLOW

EflH FLOW

200.

I VESSEL P2 STH L INE3 TURBINE f<I CIIRE INll5 'LIEF V6 U STEf

ES RISE (PSI)PRES RISE (PSI)RES fl)SL (PS)iI 5Uh (RIU/LB)lVE f'LOWIPCT)H FLOW (PCT)

100.

h

5 50.

4J

100.

0.

0.0. 5. 10. 15.

TIME lSEC)20.

-100.0 10. 15.

TIHE (SEC)20.

150.

I LEVEL(IN2 W 8 SENS3 N 8 SENS

N5 BTPASS 5

H-REF-SEP-SKIRT0 LEVEL(INCHES)0 LEVEL(INCHES)~W l((.T)EAH FLOW(PCT)

I VOIO REA2 DOPPLER3 RAH RE

TIVITYEACT IV ITCT IVITY

'100. 0.

0.0. 5. 10. 15.

TIME (SEC)20.

-20. 10. 15.

TIHE (SEC)20.

Figure 2-3. Feedwater Controller Failure, Maximum Demand at 104.2% Power, 106% Flowand Normal Feedwater Temperature

150.

I NF.UTAON2 PEAK FU(I3 AVf S()AI (

4 Ff'I OHA)(.l5 VLSSLL 5

LUXCCNTFA TEHP

CF HFI)I~FUXFIOH

CAH PLOH

200.

I VrSSCL If2 5(H I, if(Il ll)(8)INI" I~I Ci)HL IN)I'.I IN I'(CF(~ IUAI) 51(I

rs AISE (rsl)I'IiCS A I SC I('S I IA(,S A IS[ (P:il )I:il)[i I

All�)/I.B)LVf'LOH(I'Cf)H FL(IH (I'lI

a 100.W

hI

g 50.

100.

0.

0.0 5. 10. 15.

TINE (SEC)20.

-100.0. 5. 10. 15.

TIHE (SEC)20.

150.

I LEVEL ( IN2 H A SENS(3 N A SENSI

SIE fNI.5 OT('ASS 5

H-AFF-SEP-SKIAT0 LI.VEL(INCHCS)0 I.FVCL( INCIIFS)WLOH AT)CAH F LOH(f'CT)

I VOIO AEA TIVITT2 OOPPLEA CAC3 SCAAH AF VITT

61 CffvTTV

0.

50.

0.0. 5

Figure 2-4.

-210 15 20. 0. 5. 10. 15

TINE (SEC) TINE (SEC)

Feedwater Controller Failure, Maximum Demand, at 104.5X Power, 106K Flowand Reduced Feedwater Temperature of 65"F at Rated Power

20.

150.

I NEUTRON2 PEAK FUEI3 AVg SURF)u FEEOWAII'.i5 VESSEL 5

LUXCENTER TEHP

CE Hl;RT FLUXFLOW

EAH FLOW .

300.

I VESSEL P2 STH LINE3 SAFETT VI<I CORE INL5 COIIE AVF.6 1URB S)E

ES AISE IPSI)PRES RISE II'51)LVC FLOW IPCT)I Sls IOIU/I.O)VOIO FRAC IPCT)N FLOW IPCT)

g100.

50.

m

200.

100.

0.0. 20 6.

TIHE ISEC)8.

0.0 20 Q. 6.

TIHE ISEC)8.

I LEVELI IN2 H A SENS3NR SENS

5 ORIVE FL

H-REF-SEP-SKIRT0 LEVELI INCHE5)0 LEVEL(INCHES)

T)W 1 (PCT)

I VOIO RER TIVITTLEA EACTIY IT T

3 SC RE CTIVITT

fhC7nI

CA

CO

100. 0.

0.

-100.0. 20 u. 6.

TIHE )SEC)

-20. 20

TIHE ISEC)

Figure 2-5. NSIV Closure, Flux Scram, at 104.2%%d Power, 106%%d Flow and Normal Feedwater Temperature

NE DC-31107

3. MECHANICAL EVALUATION OF REACTOR INTERNALS AND FUEL ASSEMBLY

3. 1 LOADS EVALUATION

Evaluations were performed to determine bounding acoustic and flow-inducedloads, reactor internal pressure difference loads and fuel-support loads for ICFand/or FFWTR operation.

Acoustic loads are lateral loads on the vessel internals that result frompropagation of the decompression wave created by a sudden recirculation suctionline break. The acoustic loading on vessel internals is proportional to thetotal pressure wave amplitude in the vessel recirculation outlet nozzle. Thetotal pressure amplitude is the sum of the initial pressure subcooling plus theexperimentally determined pressure undershoot below saturation pressure. FFWTR

operation increases the expected acoustic loads because this downcomer sub-cooling increases and, therefore, the total pressure wave amplitude increases.The high velocity flow patterns in the downcomer resulting from a recirculationsuction 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 ofthe break. The additional subcooling in the downcomer resulting from FFWTR

operation leads to an increase in the critical flow and, therefore, to a corres-ponding increase in the flow-induced loads. The reactor internals most impactedby acoustic and flow-induced loads are the shroud, shroud support and jet pumps.

A reactor internals pressure difference analysis was performed for the 'ICFregion. The increased reactor internal pressure differences across the reactorinternals were generated for the maximum core flow at normal, upset, and faultedconditions for the reactor internal impact evaluation.

Fuel-support loads and fuel bundle lift for WNP-2 were evaluated based onresults from probabilistic fuel lift analyses performed at 106% of rated coreflow following the procedures of Reference 6. Fuel-support loads and fuelbundle liftwere evaluated for upset, faulted and fatigue load combinations.

3-1

I

NEDC-31107

It was shown that the fuel bundle lift is a small fraction of the applicabledesign. criteria (established in the NRC Safety Evaluation Report to Reference 6)

for the faulted event.

3.2 LOADS IMPACT

3.2.1 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, controlrod guide tube, control rod drive housing and jet pump. These and othercomponents were evaluated using the bounding loads, discussed in Section 3. 1,

under normal, upset, emergency and faulted conditions. It is concluded that the

stresses produced in these and other components are within the allowable design

limits given in the Final Safety Analysis Report (Chapter 3 and 4) or the ASME

Code, Section III, Subsection NG.

3.2.2 Fuel Assemblies

The fuel assemblies, including fuel bundles and channels, were evaluatedfor increased core flow operation considering the effects of loads discussed inSection 3.1 under normal, upset, faulted and fatigue load combinations. Results

of the evaluation demonstrate that the fuel assemblies are adequate to withstandICF effects to 106% rated flow.

The fuel channels were also evaluated under normal, upset, emergency and

faulted conditions for increased core flow (Reference 7). The channel wallpressure differentials were found to be within the allowable design values .

3-2

NEO C-31107

4. FLOW-INDUCED VIBRATION

To ensure that the flow-induced vibration response of the reactor internalsis 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 reactor is classified as a valid proto-

type 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. The acceptance criteria used for vibration'ssess-ment is based on a maximum allowable alternating stress intensity of 10,000 psi,

The increased core flow vibration analysis was performed by analyzing thestar tup test vibration data for the valid prototype plant (BWR/5-251 Tokai 2).Based on the results of the analysis and a review of the test data, the reactorinternals response to flow-induced vibration is expected to be within acceptablelimits for plant operation in the ICF region (region bounded as shown on the

power flow map, Figure 1-1).

4-1

y, ~

NEO C-3110?

5. FEEDWATER NOZZLE AND FEEDWATER SPARGER FATIGUE USAGE

5.1 METHOD AND ASSUMPTIONS

The fatigue experienced'by the feedwater nozzle and feedwater spargerresults from two phenomena: system cycling and rapid cycling. System cycling iscaused by major temperature changes associated with system transients. The

system cycle stresses are based on limiting cycles that use the maximum temper-

ature range possible to show expected worst conditions. These transients are

identified on thermal cycle diagrams. Thermal stresses due to these transientsare calculated by determining inner and outer metal surface temperatures usingfinite element analysis. Fatigue usage is determined by dividing the number ofdesign cycles for each transient by the number of allowable cycles for each

stress calculated. Cumulative system fatigue usage is determined by summing allof the respective transient fatigue usage factors.

Rapid cycling is caused by small, high frequency temperature fluctuationscaused by mixing of relatively colder nozzle annulus water with the reactorcoolant. The colder water impinging the nozzle bore originates from the

boundary layer of colder water formed by heat transfer through the thermal

sleeve. The mixing region extends from the feedwater nozzle surface region tothe feedwater sparger surface; therefore, rapid cycling applies to both of these

components. Once thermal stress due to rapid cycling is determined, fatigueusage is calculated and the results are added to the cumulative system cyclingusage factor to obtain the total usage factor.

The introduction of FFWTR will cause a change in calculated rapid cyclingfatigue only. This is because the system transient is very mild (smalltemperature change and relatively long duration) and is bounded by the originaldesign basis thermal stress analysis. General Electric has developed

standardized rapid cycling duty maps for each 8WR plant that cover the design

basis rapid cycles in the same manner that thermal cycle diagrams cover the

design basis thermal transients (system cycling). The methodology used todevelop the duty maps is based on the results of extensive testing of feedwater

P I j

NEDC-31107

nozzles by General Electric. FFWTR is analyzed by modifying the design cyclesin order to gauge its effect on fatigue usage.. The reduced feedwater

temperature will tend to increase fatigue usage due to an increase in thermal

stress.

An evaluation of the effect of FFWTR on the feedwater nozzle and feedwater

sparger fatigue was performed for the following conditions:

As the last step in a 12-month fuel cycle, FFWTR to a feedwater temperatureof 355'F (65'F reduction from nominal rated feedwater temperature) at ratedpower for 18 days was followed by a 3Ã per week coastdown over 12 weeks toa final power of 65K. The coastdown was initiated from a reduced feedwater

temperature of 55'F. The associated feedwater temperature at the end ofthe coastdown was 321'F.

The analysis was performed by simulating the feedwater temperature reduc-

tion during the coastdown period in four equal increments. An appropriatemaximum feedwater flow rate was assumed for each of the four increments toprovide conservative results.

5.2 FEEDWATER NOZZLE FATIGUE

The original stress analysis of the feedwater nozzle showed that themaximum system cycling fatigue usage factor for the nozzle blend radius regionwas 0.6524 for emergency and faulted conditions (Reference 8). The usage factorfor rapid cycling using the design basis (unmodified) duty map is 0.2047,

providing a total 40-year usage factor of 0.8571. The usage factor for rapidcycling including FFWTR operation is 0.2796 providing a total 40-year usage

factor of 0.9320. This result is based on FFWTR operation during every 12 month

cycle for the life of the plant. This is equivalent to 0.0019 fatigue damage

per cycle of FFWTR operation. The 40-year total usage factor remains below the

ASME Code Limit of 1.0 with FFWTR operation and is thus considered acceptable.The results are summarized in Table 5-1.

5-2

NE DC-31107

The results of this analysis are based on cycling correlations developedduring testing of various nozzle configurations. The fatigue results areintended to be a conservative best-estimate for the expected plant operation. A

more'accurate evaluation of fatigue usage could be made by considering actualplant performance.

5.3 FEEDWATER SPARGER FATIGUE

Feedwater sparger fatigue usage is calculated in the same manner as

feedwater nozzle fatigue usage. However, since the feedwater sparger is not an

ASME Boiler and Pressure Yessel Class 1 Code component, a fatigue analysis was

not originally required. WNP-2 has a welded single thermal, sleeve design whichdoes not allow leakage of feedwater flow to occur at the safe end as do otherthermal sleeve designs. This leakage flow is the primary contributor to spargerfatigue usage. Therefore, sparger fatigue usage is much less affected bychanges in feedater flow and temperarture for the welded single sleeve design.The sparger is made from stainless steel material which is less susceptible tohigh cycle fatigue than the low alloy steel of the nozzle as evidenced by thedifferences in their respective fatigue curves. Small changes in flat the (highcycle) portion of the fatigue curve can cause very significant changes infatigue usage (i.e., a r'elatively small change in stress can cause a verysignificant change in the allowable number of cycles). Thus, it becomes evidentthat the sparger fatigue damage is much less severe than nozzle fatigue damage

during feedwater condition changes like FFWTR for the welded single sleevedesign. Since the nozzle fatigue damage is so low (0.0019 per cycle), thesparger damage will be insignificant and, therefore, can be neglected.

5-3

~ > ~

NEDC-31107

Table'-1

FEEDWATER NOZZLE FATIGUE USAGE

Condition

Fatigue Usage

Due to FFWTR

(Over Normal Operation)*Per Cycle

40-Year Fatigue

Usage Factor*

Normal Operation 0.8571

FFWTR 0.0019 0.9320

*The total fatigue usage factor includes a system cycling usage factor of0.6524 due to emergency and faulted conditions as given in the originalstress analysis of the nozzle (Reference 8).

5-4

NEDC-31107

6. CONTAINMENT ANALYSIS

The impact of feedwater temperature reduction and increased core flowoperation on the containment LOCA response was evaluated.

The results show that the containment LOCA response for ICF operation alone

is bounded by the corresponding FSAR results (Reference I). Operation withFFWTR causes a slight increase in the initial drywell pressurization rate over

the rate reported in the FSAR. The calculated peak values for drywell pressure

and wetwell pressure under ICF and/or FFWTR are bounded by the correspondingvalues for the FSAR (Chapter 6) conditions. The peak value for drywell floordifferential presure (download) is bounded by the appropriate design limit of 25

psid. All other containment parameters are bounded by the results reported inthe FSAR.

The LOCA-related pool swell, condensation oscillation and chugging loads

were evaluated at the worst power/flow conditions during ICF/FFWTR operation.Pool boundary pressure load during pool swell under ICF/FFWTR conditions exceeds

the load calculated based on FSAR conditions by less than 2.2X.. However, thisload and all other pool swell loads are bounded by the appropriate design loads.The condensation oscillation and chugging loads with ICF/FFWTR conditions are

also bounded by the appropriate design loads.

6-1

NEO C-31107

7. OPERATING LIMITATION

Restrictions/limitations which are unique to ICF/FFWTR operation are

identified below.

7.1 FEEDWATER HEATERS

The FFWTR analyses have assumed that the last-stage feedwater heater isvalved out-of-service in each string of feedwater heaters (Final Feedwater

Temperature Reduction < 65'F at rated power} for exposures beyond EOC1. This

may be done at any time after EOC1 whether or not ICF is used. This is done tohelp increase or maintaine rated power after all control rods have been with-drawn at EOCl and was accounted for in the safety analyses in Sections 2.

7.2 OPERATING MAP

The allowable operating domain of the normal power-flow map has been'I

increased to allow operation at 100K power up to 106K core flow. The minimum

allowable power in this increased core flow region is bounded by the jet pump

cavitation protection interlock as shown in Figure l-l. The increased core flowreactor internal pressure differences and fuel bundle lift calculations were

analyzed and are applicable only for reactor operation within the ICF regionshown on the power flow map in Figu're 1-1.

7.3 MCPR OPERATING LIMITS

Required MCPR operating limits applicable to ICF/FFWTR have been determined

for WNP-2 as given in Table 2-2.

7.4 Kf FACTOR

For core flows greater than or equal to rated core flow, the Kf factor isequal to 1.0.

7-1

NEDC-31107

7.5 CONTROL RODS

The safety evaluation for ICF with FFMTR operation was performed with theassumption of an all-rods-out condition. This is defined as the condition ofoperation in which all control rods are fully withdrawn from the core orinserted no deeper than rod position 24.

7-2

NEO C-31107

8. REFERENCES

1. "Final Safety Analysis Report, WPPSS Nuclear Project No. 2,"as revised through Amendment 35, November 1984.

2. "General Electric Standard Application for Reactor Fuel (Supplement forUnited States)," August 1985 (NEDE-24011-P-A-7-US, as amended).

3. "Compliance of the General Electric Boiling Water Reactor Fuel Designs toStability Licensing Criteria," October 1984 (NEDE-22277-P-l).

4. Letter, C. 0. Thomas (NRC) to H. C. Pfefferlen (GE), "Acceptance forReferencing of Licensing Topical Report NEDE-24011, Revision 6, Amendme'nt

8, Thermal Hydraulic Stability Amendment to GESTAR II," April 24, 1985.

5. "BWR Core Thermal Hydraulic Stability," SIL No. 380 Revision 1, February10, 1984.

6. "BWR Fuel Assembly Evaluation of Combined SSE and LOCA Loadings," LicensingTopical Report, Amendment No. 3, October 1984 (NEDE-21175-3-P-A andNEDO-21175-3-A) .

7. "BWR Fuel Channel Mechanical Design and Deflection," General ElectricCompany, September 1976 (NEDE-21354-P).

8. "Hanford 2 - 251 BWR-5 Stress Report for Feedwater Nozzle," Section E4,Contract 72-2647, Chicago Bridge and Iron Nuclear Company; 1973.

8-1

NEDC-31107

DISTRIBUTION

Mail Code

R. J. Brandon

C. S. Chen

G. A. Deaver

S. S.

T. D.

E. C.

W. G.

J. K.

D. A.

Dua

Dunlap

Eckert

Edmonds(6)

GarrettHamon

E. C. Hansen

G. V. Kumar (3)L, K. LiuW. Marquino

J. R. PalletteA. E. Rogers

R. Seetharaman

G. L. Stevens

J. T. Teng

M. W. Thompson

J. Wallach

S. Wolf (2)C. T. Young

NEBO Library (3)

779

147

743

769

155

763

WPPS

755

769

156

770

743

763

763

763

769

747

769

156

775

763

269

528

0'a

P

W

~ N