ML17278A805
| ML17278A805 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/31/1986 |
| From: | Artigas R, Rogers A, Wolf S GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML17278A803 | List: |
| References | |
| NEDC-31107, TAC-60804, NUDOCS 8605020194 | |
| Download: ML17278A805 (97) | |
Text
8605020194 860430 PDR ADOCK 05000397 NEDC-31107 P PDR DRF L12-00737 Class II March 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. Wolf Technical Project Engineer Approved: Approved:
A.E. Rogers, Manager R. Art gas, Manager Plant Performance Engineering Licensing Services
gt P
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 contained in this document shall be construed as changing the contract. The use of this information by anyone other than WPPSS or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, Gereral Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
N EDC-31107 CONTENTS
~Pa e ABSTRACT ACKNOWLEDGMENTS vi
- l. INTRODUCTION AND
SUMMARY
- 2. SAFETY ANALYSIS 2-1 2.1 Abnormal Operational Transients 2-1 2.1.1 Limiting Transients 2-1 2.1.2 Overpressurization Analysis 2-2 2.1.3 Rod Withdrawal Error 2-3 2.2 Fuel Loading Error 2-3 2.3 Rod Drop Accident 2-3 2.4 Loss-of-Coolant Accident Analysis 2-3 2.5 Thermal-Hydraulic Stability 2-4
- 3. MECHANICAL EVALUATION OF REACTOR INTERNALS AND 3-1 FUEL ASSEMBLY
- 3. 1 Loads Evaluation 3-1 3.2 Loads Impact 3-2 3.2. 1 Reactor Internals 3-2 3.2.2 Fuel Assemblies 3-2
- 4. FLOW-INDUCED VIBRATION 4-1
- 6. CONTAINMENT ANALYSIS 6-1 7 .. OPERATING LIMITATIONS 7-1
- 8. REFERENCES 8-1
NEDC-31107 TABLES TABLE Title ~Pa e 2-1 Core-Wide Transient Analysis Results at ICF and/or FFWTR 2-6 2-2 Required MCPR Operating Limits at ICF and/or FFWTR 2-7 2-3 Overpressurization Analysis Results 2-8 5-1 Feedwater Nozzle Fatigue Usage 5-5 5-2 Feedwater Sparger Fatigue Usage ILLUSTRATIONS
~Fi ure Title Paae Operating Map 1-3 2-1 Generator Load Rejection with Bypass Failure at 104.2/ 2-9 Power, 106% Flow and Normal Feedwater Temperature 2 2 Generator Load Rejection with Bypass Failure at 104.5% 2-10 Power, 106K Flow and Reduced Feedwater Temperature 2-3 Feedwater Controller Failure, Maximum Demand, at 104.2~ 2-11 Power, 106% Flow and Normal Feedwater Temperature 2-4 Feedwater Controller Failure, Maximum Demand, at 104.5/ 2-12 Power, 106'A Flow and Reduced Feedwater Temperature 2-5 MSIV Closure, Flux Scram, at 104.2~ Power, 106Ã Flow and 2-16 Normal Feedwater Temperature
~
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-power reactivity 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) for Increased 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 the combined efforts of many individual contributors, including:
C. S. Chen, G. G. Chen, D. A. Copinger, S. K. Dhar, M. L. Gensterblum, J. K. Garrett, B. Haaberg, B. H. Koepke, M. 0. Lenz, H. X. Nghiem, J. R. Pallette, R. Seetharaman, G. L. Stevens, M. W. Thompson, S. Wolf and C. T. Young
- 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 Figure 1-1. This report presents the results of a safety evaluation for operation with ICF for Cycle 1 [up to and including End-of-Cycle 1 (EOC1) exposure]. The safety evaluation also covers operation for exposure beyond standard EOC1* with ICF and/or last-stage feedwater heaters valved out, followed by a natural reactivity coastdown bounded by 106 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 65/ power) should occur only at the end-of-cycle. The extended region of operation with increased core flow followed 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 flow with and without FFWTR. The loss-of-coolant accident (LOCA), fuel loading error accident, rod drop accident, and rod withdrawal error event were also reevaluated for increased core flow operation.'hese events were also reevaluated for end-of-cycle operation with ICF and the last-stage feedwater heaters valved out.
- EOC1 is defined as the core average exposure at which there is no longer sufficient reactivity to achieve rated thermal power with rated core flow, all control rods withdrawn (beyond Rod Position 24), all feedwater heaters in service 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 that the 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 technical specifications 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 within the increased core flow region of the operating map as illustrated in Figure l-l for Cycle 1 and for exposures beyond EOC1 with the conditions assumed in the analysis. The LCPRs and the minimum critical power ratio (MCPR) operating limits 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 appropriate values as shown in Table 2-2.
1-2
0 '
~ ~
~ ~
130 INCAEASED FLOW CONTROL PUMP CORE FLOW 120 VALVE POSITION SPEED APRM STP SCRAM REGION 0'108,'I CURVE )% OF FULL STROKE) )% RATED)
APRM ROD BLOCK 110 0 NAT CIRC ROD BLOCK MONITOR 1 0 2 100 I 26 00) 3 0 4 14 6 21 8 28 ~ 100 7 38 8 46 9 68 RATED ROD LINE 10 88 TO ALLOWABLE m OPERATING tD I DOMAIN I w 80 r2 0.68 W~ s 61'% 80% C) 0.68 Ws+ 42%
40 60%
0.88' 40%
CAVITATION 30 CAVITATIONLINES: INTERLOCK JET PUMP NOZZLE JET PUMP SUCTION RECIRC PUMP 10 10 30 40 50 60 70 80 90 100 IO CORE F LOW (psrcsntl Figure l-l. Operating t)ap
- 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) with and 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 licensing submittal (Reference 1). The transient performance responses are presented in Figures 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 licensing submittal (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 RWE event'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 on whether final feedwater heaters are in service. The 104.5 power level provides a 5X steam flow margin to the 100% power operating conditions to simulate eventual stretch power operation, similar to the original FSAR analyses.
2-1
~ ~
e NEOC-31107 Oecreasing the power from the 1005 rated condition along the 106% core flow line will result in an increase in transient sCPR 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 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, 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 with FFWTR. The ICF for the LRNBP event results in a less severe overpressure transient 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 (which is flow biased) increases, giving an unacceptably high MCPR limit. Thus, the RBM 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 of operation 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 with the margin to the OLCPR. Thus, the FSAR bCPR would not be affected by this event 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 fuel enthalpy 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 without FFWTR bounds operation with ICF and FFWTR.
The effect of increased core flow on LOCA analyses is not significant because the parameters which most strongly affect the calculated peak cladding temperature (PCT), i.e., high power node boiling transition time and core ref 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 flow condition.
Therefore, it is concluded that the LOCA PCT is acceptable and that the current maximum average planar linear heat generation rates (MAPLHGRs) for WNP-2 are applicable for ICF.
2-3
0 ~~
,I
NEOC-31107
- 2. 5 TMERMAL-HYORAULI C 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 fuel design limits being exceeded. Furthermore, the onset of power oscillations'for which corrective actions are necessary is reliably and readily detected and sup-pressed by operator actions and/or automatic system functions. The stability compliance of all licensed GE BWR fuel designs including those fuels contained in the General Electric Standard Application for Reactor Fuel (GESTAR, Reference
- 2) is demonstrated on. a generic basi's 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 analysis for 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 to WNP-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 is bounded by the fuel integrity analyses in Reference 3.
Similarly, operation in the FFWTR mode is bounded by the fuel integrity analyses 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 cycle oscillations. Reference 3 also demonstrates that even if neutron flux limit cycle oscillations did occur just below the neutron flux scram setpoint, fuel design limits are not exceeded for those GE BWR fuel designs contained in General 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 large oscillations.
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 during all plant maneuvering and operating states, a generic set of operating recom-mendations has been developed as set forth in Reference 5 and communicated to all 6E BWRs. These recoranendations instruct the operator on how to reliably detect 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 Rated Haximum Hax imum Feedwater Haximum Core Ave. Haximum Haximum Steam Temperature Neutron Surface Dome Yessel L ine Transient Figure Power Flow Reduction F lux Heat Flux Press Press Press Description Number (I NBR) (I NBR) ('F) ('X NBR) ('l Initial) (psig) (psig) (psig) aCPR LRNBP Ref. 1 104.4 100 0 236.4 107.8 1173 1202 1168 0.09 LRNBP 2.1 104.2 106 0 252.4 108. 8 1172 1203 1168 0. 11 LRNBP 2.2 104.5 106 -65 243.2 108.8 1160 1191 1157 0. 11 FMCF Ref. 1 104.4 100 0 154. 3 108. 7 1148 1177 1140 0.08 FMCF 2.3 104.2 106 0 163. 7 109. I 1145 1177 1141 <0. 13 FMCF 2,4 104.5 106 65 174,7 113. 9 1138 1166 1135 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 rated conditions.
- 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 Initial Initial a Core Core Transient Power Flow Description (X NBR) (X NBR) aCPR OLCPR A
OLCPR 8
LRNBPf(FSAR) 104.4 100 0.09 1.20 1.12 LRNBP 104.2 106 0.11 1.22 1.14 FWCF (FSAR) 104.4 100 0.08 1.19 1.16 FMCFg 104.5 106 0.13 1.24 1.21 aCPR OLCPR RWE (FSAR) 104.4 100 0.18 1.24
- a. LRNBP = Load rejection with bypass failure, FWCF = feedwater controller failure at maximum demand, RME = rod withdrawal error.
- c. Includes Option A adjustment factors.
- d. Includes Option 8 adjustment factors.'.
Option A and 8 adjustment factors are specified in the NRC safety evaluation report on ODYN (NEDO-24154 and NEDE-24154P).
h.. Required OLCPR using either Option A or Option 8 adjustment factor with rod block monitor of 106Ã at rated flow 2-7
I e
EEOC-31107 Table 2-3 OVERPRESSURIZATION ANALYSIS RESULTS Maximum Initial Initial Vessel Power Flow Pressure Transient (%) (~) (psig) Figure No.
MSIV Closure - Flux Scram 104.3 100 1266 Reference 1 (FSAR)
MSIV Closure - Flux Scram 104. 2 106 1264 Figure 2-5
( ICF w/o FFWTR) 2-8
0
', )i l
I NEUTRON LUX I VESSEL P ES AISE (PS I) 2 PEAK FUEl CENTER TEMP 2 STH Llt(E PRES RISE (PSI) 3 AVL SUAF( CF. HEAT FLUX 3 TURBINE f RES RISE (PSI) 150. <I COAE It(L I SUB (BTU/LB) k FfE044IL( FLOH 5 VESSEL 5 EAH FLOH 5 RELIEF V LVE FLOH (PC'f) 6 TURB STE 4 FLOH (PCT) 100.
g (5 50. 0.
- 0. -I 00.
0 20 6. 0 2. 6. 8.
TIME (SEC) . TIME (SEC)
I LEVEL ( I H-AEF-SEP-SKIRT I VOIO AEA TIVITT 2 4 A SENS 0 LEVEL(INCHES) 2 DOPPLEA EACT I V ITT 3 N 4 SENS 0 LEVEL(INCHES) 3 SCRAH RE CTI VITT 200. L TVD~Xtf) 5 OAIVE FL 4 I (PCT)
IOO.
- 0. -I.
I
-I 00. -2 0 LJ. 6. 8. 0. 2. 3.
TIME (SEC) TIME (SEC)
I Figure 2-1. Generator Load Rejection with Bypass Failure at 104.2X Power, 106% Flow and Normal Feedwater Temperature
l
~ ~ ~
I NEUTRON LUX I VESSEL P ES RISE (PS I) 2 PEAK FUE( CENTER TEHP 2 STN LINE PRES RISE (PSI) 150. 3 AVF SURF( CE HFAT FLUX 200. 3 T()RBINE I RES f((SE (PSI) 0 FEEE)HATE( FLOW 4 Ri:LIEF V(LVCKOH (PCT) 5 VESSEL S EA(( FLOH 5 RELIEF V LVE FLOH (PCT) 6 TURB S'(E H FLOH (PCT) 3 100. 100.
I 0.
- 0. -100, 0 20 6. 8. 0. 20 6. 8.
TINE (SEC) TI((E (SEC)
C7 Dl I
I LEVEL(I H-REF-SEP-SKIRT I VOID RER TIVITT
, 2 H A SENS 0 LEVEL(INCHES) 2 OOPPLER EACT I V ITT C) 3 N R SENS 0 LEVEL(INCHES) 3 SCRAH RE CTI VI7Y fVCG~T) 5 ORIVE FL W I (PCT) 100. 0.
0.
-100. -2
- 0. 2. 6. 0. 2. 3. 4.
TINE (SEC) TINE (SEC)
Figure 2.2. Generator Load Rejection With Bypass Failure at 104.5X Power, 106% Flow and Reduced Feedwater Temperature of 65 F at Rated Power.
I NEUTRON I LUX I VESSEL P ES RISE (PS I) 0 Pffl( FUEI CENTER YEHP 2 STH LINC PACS f)ISE (PSI) 150. 3 AVE SURf'( CF. HEAT FLUX 3 YUAOfNE f AE~ fl(Sf.'PSI )
- l. C(:OMiI.l FLON i) C(iAE INA I SUA lBlllfLB)
VESSEL 5 EAH FLON 5 '( IEF V LVE FLO)l(PCT) 6 U Sf( f H FLON (PC'f) a 100. 100.
I Ki 50. 0.
W
-100.
- 0. 10. )5. 20. 0. 10. 15. 20.
TIHE (SEC) TIHE (SEC) m nI I 'LEVEL ( IN H-AEF-SEP-SK lRT I VOIO BE TIVITT 2 N A SENQO LEVEL(INCHES) 2 OOPPLER EACTI V IT C) 3 N A SEN 0 LEVEL(INCHES) 3 AAH RE CTI VITT 150,
~l5%1NL ~LAN (((T) C 5 BYPASS 5 EAH FLO)I(PCT) 100. 0.
- 0. -2
~ ~ 5. 10. 15. 0. 5. 10. 15. 20.
TINE (SEC) TINE (SEC)
Figure 2-3. Feedwater Controller Failure, Maximum Demand at 104.2X Power, 106% Flow and Normal Feedwater Temperature
f I IJF(ITA()JI I ((Ir I Vl Ssf.l. f'f f") Afsf lf "I) 2 ('I.t(K I (If'I I,f,Nfff1 (FHP c 'i(It I INI I I<t i AISI (I'Sl) p()f(I J ~c(: Nfl(I f(,UX
'( l(i(IH)lll I Jg,s .I(/sf (f".ilI
- ) (JVI 2(JO.
JI III (I(JAII I I I OH il LJ)I<1 INII I:I(IFJ ((J)(VI,A) 5 VLSSLL s L'(JH f LOH ! (0) It.l V( I Vi' LOJJ JI 0 r')
~
( IJ)ftn Sll( H I I.t)JI tl'(.I) aUJ IM.
5 h
I I so. 0.
- 0. -100.
- 0. 5. 10. 15. 20. 0. 10. 15. 20.
TIME (SECI TIME (SEC)
I LEVEL(I H-AFF-SEP-SKIAT I VOID AE TIVITT 2 H A SO(St 0 LI.VFL()NCHCS) 2 OOPPLEA I'ACIJVTlT 3 N A SENcl 0 I.FVEI (INC((FS) 3 SCAAH AF MVITT 150. tl CdhE TN(.i T Fiof( (ICI) WDI CTivTTV S 0)I'ASS S EAH f LOH(f'Cf) 0.
-I I
LJ CI UI CC
- 0. -2
- 0. 5. 10. 15 20. 0. 5. 10. 15. 20.
TIHE ISEC) TIHE (SEC)
Figure 2-4. Feedwater Controller Failure, Maximum Demand, at 104.5% Power, 106K Flow and Reduced Feedwater Temperature of 65 F at Rated Power
I NEUTRON f LUX I VESSEL P FS AISE (PSI) 2 PFAK FUEI CENTER TEMP 2 SIM LINE PRES RISE (f'SI) 3 AVF. SUnf( CE MCAT FLUX 3 S(lFETV Vf LVE FLOH (PC')
300. CORE l(H. I STD (fi(U/LO) 4 FEEOHAI):( FLOH >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 is valved 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 to help 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 flow reactor internal pressure differences and fuel bundle lift calculations were analyzed and are applicable only for reactor operation within the ICF region shown 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 is equal to 1.0.
7-1
NEOC-31107 7.5 CONTROL ROOS The safety evaluation for ICF with FFWTR operation was performed with the assumption of an all-rods-out condition. This is defined as the condition of operation in which all control rods are fully withdrawn from the core or inserted 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 for United States)," August 1985 (NEDE-24011-P-A-7-US, as amended).
- 3. "Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," October 1984 (NEDE-22277-P-1).
- 4. Letter, C. 0. Thomas (NRC) to H. C. Pfefferlen (GE), "Acceptance for Referencing 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, February 10, 1984.
- 6. "BWR Fuel Assembly Evaluation of Combined SSE and LOCA Loadings," Licensing Topical 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 Electric Company, 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
'ail Code R. J. Brandon 779 C. S. Chen 147 G. A. Deaver 743 S. S. Dua 769 T. D. Dunlap 155 E. C. Eckert 763 W. G. Edmonds(6) WPPS J. K. Garrett 755 D. A. Hamon 769 E. C. Hansen ~
156 G. V. Kumar (3) 770 L. K. Liu 743 W. Harquino 763 J. R. Pallette 763 A. E. Rogers 763 R. Seetharaman 769 G. L. Stevens 747 J. T. Teng 769 H. W. Thompson 156 J. Wallach 775 S. Wolf (2) 763 C. T. Young 269 NEBO Library (3) 528
l NE DC-31107 DR F L12-00737 CLASS II MARCH 1986 DAC 310 SAFETY REYIEW OF WPPSS NUCLEAR PROJECT NO. 2 AT CORE FLOW CONDITIONS ABOYE RATED FLOW THROUGHOUT CYCLE 1 AND FINAL FEEDWATER TEMPERATURE REDUCTION S. WOLF GENERAL ELECTRIC
NEDC-31107 DRF L12-00737 Class II March 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. Wolf Technical Project Engineer Approved: Approved:
A.E. Rogers, Manager R. Art gas, Manager Plant Performance Engineering Licensing Services NUCLEAR ENERGY BUSINESS OPERATIONS ~ GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 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 vi
- l. INTRODUCTION AND
SUMMARY
- 2. SAFETY ANALYSIS 2-1 2.1 Abnormal Operational Transients 2-1 2.1.1 Limiting Transients 2-1 2.1.2 Overpressurization Analysis 2-2 2.1.3 Rod Withdrawal Error 2~3 2.2 Fuel Loading Error 2~3 2.3 Rod Drop Accident 2~3 2.4 Loss-of-Coolant Accident Analysis 2-3 2.5 Thermal-Hydraulic Stability 2-4
- 3. MECHANICAL EVALUATION OF REACTOR INTERNALS AND 3-1 FUEL ASSEMBLY 3.1 Loads Evaluation 3-1 3.2 Loads Impact ' 2 3.2.1 Reactor Internals 3~2 3.2.2 Fuel Assemblies 3-2
- 4. FLOW-INDUCED VI BRATION 4-1
- 5. FEEDWATER NOZZLE AND FEEDWATER SPARGER FATIGUE USAGE 5-1 5.1 Method and Assumption 5-1 5.2 Feedwater Nozzle Fatigue 5-2 5.3 Feedwater Sparger Fatigue '5-3
- 6. CONTAINMENT ANALYSIS 6-1 7 .. OPERATING LIMITATIONS 7-1
- 8. REFERENCES 8-1 1 11
NEDC-31107 TABLES TABLE T it 1 e ~Pa e 2-1 Core-Wide Transient Analysis Results at ICF and/or FFWTR 2-6 2-2 Required MCPR Operating Limits at ICF and/or FFWTR 2-7 2-3 Overpressurization Analvsis Results 2-8 5-1 Feedwater Nozzle Fatigue Usage 5-5 5-2 Feedwater Sparger Fatigue Usage ILLUSTRATIONS
~Fi ere Title Paae Operating Map 1-3 2-1 Generator Load Rejection with Bypass Failure at 104.2% 2-9 Power, 106% Flow and Normal Feedwater Temperature 2-2 Generator Load Rejection with Bypass Failure at 104.5% 2-10 Power, 106% Flow and Reduced Feedwater Temperature 2-3 Feedwater Controller Failure, Maximum Demand, at 104.2% 2-11 Power, 106% Flow and Normal Feedwater Temperature 2-4 Feedwater Controller Failure, Maximum Demand, at 104.5% 2-12 Power, 106% Flow and Reduced Feedwater Temperature 2-5 MSIV Closure, Flux Scram, at 104.2% Power, 106% Flow and 2.-16 Normal Feedwater Temperature
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 P) can increase core flow to operate within the region of the operating map 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-power C
reactivity under standard feedwater conditions) with all control'ods out, can continue to operate in the region of the operating map bounded by the 106% core flow line between 100'A 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 'atio (MCPR) operating limits will be changed from the values established by the Final Safety Analysis Report licensing submit'tal, to the appropriate values (Table 2-2) for Increased Core Flow (ICF) and Final Feedwater Temperature Reduction (FFWTR) operating conditions. All other operating limits established in the Cycle 1 licensino basis have been found to be bounding for the ICF 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. Chen, G. G. Chen, 0. A. Copinger, S, K. Dhar, M. L. Gensterblum, D. K. Garrett, B. Haaberg, B. H. Koepke, M. 0. Lenz, H. X. Nghiem, J. R. Pallette, R. Seetharaman, G. L. 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 Figure 1-1. This report presents the results of a safety evaluation. for operation with ICF for Cycle 1 [up to and including End-of-Cycle 1 (EOC1) exposure]. The safety evaluation also covers operation for exposure beyond standard EOC1* with ICF and/or last-stage feedwater heaters valved out, followed by a natural reactivity 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 only at the end-of-cycle. The extended region of operation with increased core flow followed'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 flow with and without FFWTR. The loss-of-coolant accident (LOCA), fuel loading error accident, rod drop accident, and rod withdrawal error event were also reevaluated for increased core flow operation. These events 'were also reevaluated 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 longer sufficient reactivity to achieve rated thermal power with rated core flow, all control rods withdrawn (beyond Rod Position 24), all feedwater heaters in service 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-induced vibration response of the reactor internals was also evaluated to ensure that the 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 technical specifications 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 within the 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) operating limits 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 appropriate values as shown in Table 2-2.
1-2
130 FLOW CONTROL PUMP INCREASED 120 VALVEPOSITION SPEED APRM STP SCRAM CORE FLOW CURVE l% OF FULL STROKEI OC RATED) REGION 13.6 APRM ROD BLOCK 110 0 NAT CIRC ROD BLOCK MONITOR 1 0 2 100
~
3 0 l108,100) 4
~
14 6 21 6 28
~
7 38
~
100 8 46 9 66 RATED ROD LINE 10 88 70 10NC ALLOWABLE I 4 OPERATING 6o DOMAIN
/2 0.68 Wr t 611' 80%
0.68 Wr + 42%
40 0.68 Wr i 40m 30 CAVITATION CAVITATIONLINES: INTERLOCK JET PUMP NOZZLE JET PUMP SUCTION RECIRC PUMP 10 0
0 10 20 30 40 50 60 70 80 110 CORE FLOW (percent)
Figure 1-1. Operating Map
- 2. SAFETY ANALYSIS 2.1 ABNORMAL OPERATIONAL TRANSIENTS 2.1. 1 Limitin Trans i ents The limiting abnormal operational transients analyzed in the Cycle 1 FSAR licensing submittal (Reference 1) were reevaluated for increased core flow and/or FFWTR.
Nuclear transient data for 104.5% power*, 106% core flow (104.5, 106) with and 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 (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 licensing submittal (Reference 1). The transient performance responses are presented in Figures 2-1 through 2-,4. The results demonstrate that the ACPR values and the critical power ratio operating limits for the LRNBP and FMCF events increase compared with the corresponding FSAR values. However, the FSAR licensing submittal (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 on whether final feedwater heaters are in service. The 104;5 power level provides a 5X steam flow margin to the 100% power operating conditions to simulate eventual 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 flow line 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 limit MCPR'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 with FFWTR. The ICF for the LRNBP event results in a less severe overpressure transient 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 of operation 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 with the margin to the OLCPR. Thus, the FSAR hCPR would not be affected by this event 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 fuel enthalpy 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 without FFWTR bounds operation with ICF and FFWTR.
The effect of increased core flow on LOCA analyses is not significant because the parameters which most strongly affect the calculated peak cladding temperature (PCT), i.e., high power node boiling transition time and core reflooding time, have been shown to be relatively insensitive to increased core flow.
Results of the LOCA analysis performed show that the PCT for ICF increases ,
by less than 5'F throughout the break spectrum compared to the rated core flow condition.
Therefore, it is E concluded that the LOCA PCT is acceptable and that the current maximum average planar linear heat generation rates (MAPLHGRs) for WNP-2 are applicable for ICF.
2-3
- 2. 5 THERMAL-HYDRAULIC STABILITY 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 fuel design limits being exceeded. Furthermore, the onset of power oscillations for which corrective actions are necessary is reliably and readily detected and sup-pressed by operator actions and/or automatic system functions. The stability compliance of all licensed GE BWR fuel designs including those fuels contained in the General Electric Standard Application for Reactor Fuel (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 analysis for 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 to WNP-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 is bounded by the fuel integrity analyses in Reference 3.
Similarly, operation in the FFWTR mode is bounded by the fuel integrity analyses 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 cycle oscillations. Reference 3 also demonstrates that even if neutron flux limit cycle oscillations did occur just below the neutron flux scram setpoint, fuel design limits are not exceeded for those GE B'WR fuel designs contained in General Electric Standard Application for Reactor Fuel (GESTAR, Reference 2).
These evaluations demonstrate that substantial thermallmechanicaI margin is available for the GE BWR fuel designs even in the unlikely event of very large oscillations.
2-4
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 during all plant maneuvering and operating states, a generic set of operating recom-mendations has been developed as set forth in Reference 5 and communicated to all GE BWRs. These recomnendations instruct the operator on how to reliably detect 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 Rated Haximum Feedwater Maximum Core Ave. Haximum Hax1mum
'aximum Steam Temperature Neutron Surface Oome Vessel Line Transient Figure Power Flow Reduction Flux Neat Flux Press Press Press Descr1ption Number (X NBR) (X NBR) (oF) (X NBR) ('l Init ia 1 ) (ps ig) (psig) (psig) aCPR LRNBP Ref. 1 104.4 100 0 236.4 107.8 1173 1202 1168 0.09 LRNBP 2.1 104.2 106 0 252.4 108.8 1172 1203- 1168 0.11 LRNBP 2.2 104.5 106 65 243.2 108.8 1160 1191 1157 0.11 FMCF Ref. 1 104.4 100 0 154.3 108.7 1148 1177 1140 0.08 FWCF 2.3 104.2 106 0 163. 7 109. 1 1145 1177 1141 <0. 13 FMCF 2.4 104.5 106 $5 174.7 113.9 1138 1166 1135 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 rated conditions.
- 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 Ini ti al Initial a Core Core Transient Power Flow Description (X NBR) (X NBR) aCPR OLCPR OLCPR A B LRNBPf(FSAR) 104.4 100 0.09 1.20 1.12 LRNBP 104.2 106 0.11 1'. 22 1.14 FWCF (FSAR) 104.4 100 0.08 1.19 1.16 FWCFg 104.5 106 0.13 1.24 1.21 a,CPR OLCPR RWE (FSAR) 104.4 100 0;18 1.24
- a. LRNBP = Load rejection with bypass failure, FWCF = feedwater controller failure at maximum demand, RWE = rod withdrawal error.
- c. Includes Option A'djustment factors.
- d. Includes Option B adjustment factors.
- e. Option A and B adjustment factors are specified in the NRC safety evaluation report on ODYN (NEDO-24154 and NEDE-24154P).
- h. Required OLCPR using either Option A or Option B adjustment factor with rod block monitor of 106% at rated flow 2-7
NEOC-31107 Table 2-3 OVERPRESSURI ZATION ANALYS!S RESULTS Maximum Ini ti al Initial Vessel Power Flow Pressure Transient (X) (l) (psig) Figure No.
MSIV Closure - Flux Scram. 104.3 100 1266 Reference 1 (FSAR)
MSIV Closure - Flux Scram 104.2 106 1264 Figure 2-5 (ICF w/o FFWTR) 2-8
I NEUTRON LUX I VESSEL P ES AISE (PSI) 2 PEAK FUE CENTER TEHP 2 SIH LINE PRES RISE IPSI) 150. 3 AVE SURF CE HEAT FLUX 3 TU88)NC RES RISE (PSI) 4 FEEOHAIT) FLON 200.
4 CORE INL I SUI) IBTU/LB) 5 VESSEL S EAH FLOH 5 RELIEF V LVE FLOH (PCT) 6 TURB SIE H FLOH IPCT) 100. 100.
p lh 0.
5
- 0. -100.
- 0. 2. 4. 6. 8. 0. 20 4. 6. 8.
TIHE lSEC) TINE (SEC)
I LEVEL(1 H-REF-SEP-SKIRT I VOIO BE TIVITT 2 II R SENS D LEVELI INCHES) 2 ODPPLEA EACT I V ITT 200. 3 N A SENS D LEVEL(INCHES) 3 SCRAH RE CT I V I TT T) 5:DRIVE FL 4 I (PCT) 100. 0.
0.
-100. 20 0 2. 4. 6. B. 0 2. 3. 4.
TIHE )SEC) TIHE (SEC)
Figure 2-1. Generator Load Rejection with Bypass Failure at 104.2X Power, 106K Flow and Normal Feedwater Temperature
1 I NEUTRON LUX I VESSEL P ES AISE (PSI) 2 PEAK FUEl CFNTER TEHP 2 STH LINE PRES RISE (PSI) 150. 3 AVE SuflFf CE t)EAT FLUX 200. 3 T()ABIDE f A)S R/SE (P51)
W FEEOWA(f:f FLOW 'I AELfEF Vf LVE FLOW (PCTl 5 VESSEL 5 EAH FLOW 5 BELIEF V LVE FLOW (PCT) 6 1UAB STEf H FLOW (PCT) 100.
0.
- 0. -100.
- o. 2. 6. 8. 0 20 Q. 6. 8.
TIHE (SEC) TINE (SEC)
I LEVEL(I H-AEF-SEP-SKIRT I VOIO REA TIVITT 2 W R SENS D LEVE(.(INCHES) 2 OOPPLEA EACT I V ITT 200. 3 N A SENS D LEVEL(INCHES) 3 SCAAH RE CTIVITT I) 5 ORIVE FL ff I (PCT) 100. 0.
0.
-100. 20 0 2. 6. 8. 0. 2. 3.
TIHE (SEC) TIHE (SEC)
Figure 2.2. Generator Load Rejection With Bypass Failure at l04.5f. Power, 1061, Flow and Reduced Feedwater Temperature of 65 F at Rated Power.
I NEUTRON LUX I VESSEL P ES RISE (PSI) 9 PEAK FUF) CENTER TEHP 2 STH L INE PRES RISE (PSI) 150. AVE 5URFf CE f(EAT FLUX 200. 3 TURBINE f RES fl)SL (PS)i FEI.OWAIE( FLOW <I CIIRE INll I 5Uh (RIU/LB)
VESSL'L S EflH FLOW 5 'LIEF V lVE f'LOWIPCT) 6 U STEf H FLOW (PCT) 100. 100.
h 5 50. 0.
4J
- 0. -100.
- 0. 5. 10. 15. 20. 0 10. 15. 20.
TIME lSEC) TIHE (SEC)
I LEVEL (IN H-REF-SEP-SKIRT I VOIO REA TIVITY 2 W 8 SENS 0 LEVEL(INCHES) 2 DOPPLER EACT I V IT 3 N 8 SENS 0 LEVEL(INCHES) 3 RAH RE CT IVITY 150.
5 BTPASS N
5
~W EAH l((.T)
FLOW(PCT)
'100. 0.
- 0. -2
- 0. 5. 10. 15. 20. 0. 10. 15. 20.
TIME (SEC) TIHE (SEC)
Figure 2-3. Feedwater Controller Failure, Maximum Demand at 104.2% Power, 106% Flow and Normal Feedwater Temperature
I NF.UTAON LUX I VrSSCL If rs AISE (rsl) 2 PEAK FU(I CCNTFA TEHP 2 5(H I, if(I I'IiCS A I SC I('S I I 150. 3 AVf S()AI ( CF HFI)I~FUX Ff'I OHA) (.l FIOH 200. l ll)(8)INI" I A(,S A IS[ (P:il )
All)/I.B) 4 5 VLSSLL 5 CAH PLOH
~I Ci)HL IN)I I:il)[i I
'.I IN I'(CF LVf'LOH(I'Cf)
( IUAI) 51(I
~ H FL(IH (I'l I aW 100. 100.
h I
g 50. 0.
- 0. -100.
0 5. 10. 15. 20. 0. 5. 10. 15. 20.
TINE (SEC) TIHE (SEC)
I LEVEL ( IN H-AFF-SEP-SKIAT I VOIO AEA TIVITT 2 H A SENS( 0 LI.VEL(INCHCS) 2 OOPPLEA CAC 150. 3 N A SENSI 0 I.FVCL( INCIIFS) 3 SCAAH AF VITT SIE fNI. WLOH AT) 61 CffvTTV 5 OT('ASS 5 CAH F LOH(f'CT) 0.
50.
- 0. -2
- 0. 5 10 15 20. 0. 5. 10. 15 20.
TINE (SEC) TINE (SEC)
Figure 2-4. Feedwater Controller Failure, Maximum Demand, at 104.5X Power, 106K Flow and Reduced Feedwater Temperature of 65"F at Rated Power
I NEUTRON LUX I VESSEL P ES AISE IPSI) 2 PEAK FUEI CENTER TEHP 2 STH LINE PRES RISE II'51) 150. 3 AVg SURF) CE Hl;RT FLUX 300. 3 SAFETT VI LVC FLOW IPCT) u FEEOWAII'.i FLOW
~
NEDC-31107 Table'-1 FEEDWATER NOZZLE FATIGUE USAGE Fatigue Usage Due to FFWTR Condition (Over Normal Operation)* 40-Year Fatigue Per Cycle Usage Factor*
Normal Operation 0.8571 FFWTR 0.0019 0.9320
- The total fatigue usage factor includes a system cycling usage factor of 0.6524 due to emergency and faulted conditions as given in the original stress analysis of the nozzle (Reference 8).
5-4
- 6. CONTAINMENT ANALYSIS The impact of feedwater temperature reduction and increased core flow operation 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 with FFWTR 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 corresponding values for the FSAR (Chapter 6) conditions. The peak value for drywell floor differential presure (download) is bounded by the appropriate design limit of 25 psid. All other containment parameters are bounded by the results reported in the FSAR.
The LOCA-related pool swell, condensation oscillation and chugging loads were evaluated at the worst power/flow conditions 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, this load 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 is valved 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 to help 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 flow reactor internal pressure differences and fuel bundle lift calculations were analyzed and are applicable only for reactor operation within the ICF region shown 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 is equal to 1.0.
7-1
NEDC-31107 7.5 CONTROL RODS The safety evaluation for ICF with FFMTR operation was performed with the assumption of an all-rods-out condition. This is defined as the condition of operation in which all control rods are fully withdrawn from the core or inserted 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 for United States)," August 1985 (NEDE-24011-P-A-7-US, as amended).
- 3. "Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," October 1984 (NEDE-22277-P-l).
- 4. Letter, C. 0. Thomas (NRC) to H. C. Pfefferlen (GE), "Acceptance for Referencing 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, February 10, 1984.
- 6. "BWR Fuel Assembly Evaluation of Combined SSE and LOCA Loadings," Licensing Topical 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 Electric Company, 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 779 C. S. Chen 147 G. A. Deaver 743 S. S. Dua 769 T. D. Dunlap 155 E. C. Eckert 763 W. G. Edmonds(6) WPPS J. K. Garrett 755 D. A. Hamon 769 E. C. Hansen 156 G. V. Kumar (3) 770 L, K. Liu 743 W. Marquino 763 J. R. Pallette 763 A. E. Rogers 763 R. Seetharaman 769 G. L. Stevens 747 J. T. Teng 769 M. W. Thompson 156 J. Wallach 775 S. Wolf (2) 763 C. T. Young 269 NEBO Library (3) 528
0
'a P
W
~ N