ML17279A164

Cycle 3 Reload Summary Rept.

ML17279A164
Person / Time
Site:	Columbia
Issue date:	03/31/1987
From:	Humphreys M, Talbert R, Wolkenhauer W WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
Shared Package
ML17279A161	List: ML17279A160 ML17279A162 ML17279A164 ML17279A167 ML17279A168
References
WPPS-EANF-109, WPPSS-EANF-109, NUDOCS 8704030073
Download: ML17279A164 (35)

Text

8704030073 870327 PDR ADOCK 05000397 P PDR MNP-2 CYCLE 3 RELOAD

SUMMARY

REPORT S

Prepared By:

M. C. Molkenhauer, Principal Engineer, Nuclear Fuel Reviewed By:

R. J. albert, Plant Engineer Reviewed By:

M. C. Hu hreys, Plant Engineer l:oncur With:

M. R uestefeld,. Supervisor, WNP-2 Reactor Engineer Concur With:

K. D. Cowan, Manager, WNP-2 Technical Approved By:

D. L. Larkin, Manager, Engineering Analysis and Nuclear Fuel

NOTICE This report is derived in part through information provided to Washington Public Power Supply System (Supply System) by Advanced Nuclear Fuels Corpora-tion. It is being submitted by the Supply System to the U.S. Nuclear Regula-tory Commission in partial support of the WNP-2 Cycle 3 reloading licensing submittal. The information contained herein is true and correct to the best of the Supply System's knowledge, information, and belief.

WNP-2 CYCLE 3 RELOAD

SUMMARY

REPORT TABLE OF CONTENTS

~Pa e

1.0 INTRODUCTION

. 1

'.0 GENERAL DESCRIPTION OF RELOAD SCOPE 1 3.0 WNP-2 CYCLE 3 OPERATING HISTORY 3 4.0 RELOAD CORE DESCRIPTION ~ ~ ~ ~ 6 5.0 FUEL MECHANICAL DESIGN . 7 6.0 THERMAL HYDRAULIC DESIGN . 8 6.1 Hydraulic Compatability . . . . . . . 9 6.2 Fuel Cladding Integrity Safety Limit 9 6.3 Fuel Centerline Temperature . . . . . 9 6.4 Bypass Flow Characteristics . 9 6.5 Thermal Hydraulic Stability . 10

'0 7.0 NUCLEAR DESIGN .

7.1 Fuel Bundle Nuclear Design ~ ~ ~ ~ ~ ~ ~ ~ ~ 10 7.2 Core Nuclear Design . . . . . ~ ~ ~ ~ ~ ~ ~ ~ ~ 11 7.3 Comparison of Major Core Parameters . 11 8.0 ANTICIPATED OPERATIONAL OCCURRENCES 12 8.1 Core Wide Transients 13 8.2 Local Transients 14 8.3 Reduced Flow Operations . . . ~ ~ ~ ~ ~ ~ ~ ~ ~ 14 8.4 ASME Overpressurization Analysis 14 8.5 Increased Flow Operation ~ ~ ~ ~ ~ ~ ~ ~ ~ 14 8.6 Single Loop Operation . . . . . . . . 15 9.0 POSTULATED ACCIDENTS . 16 9.1 Loss of Coolant Accident 16 9.2 Rod Drop Accident . . . . . . . . . . ~ ~ ~ ~ ~ ~ ~ ~ ~ 16 9.3 Single Loop Operation . 16"

/ S TABLE OF CONTENTS (CONTD.)

Page 10.0 STARTUP PHYSICS TEST PROGRAM . ~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ ~ 16 10.1 Cor e Load Verification Test ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 17 10.2 Contro1 Rod Functiona1 Test ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 17 10.3 Subcritica1 Margin Test . . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 17 10.4 Tip Asymmetry Test ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 17

11.0 REFERENCES

. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 18

Ig 5

WNP-2 CYCLE 3 RELOAD

SUMMARY

REPORT

1.0 INTRODUCTION

The second reload of the Washington Public Power Supply System Plant No.

2 (WNP-2) will utilize Advanced Nuclear Fuels Corporation (ANF), (for-merly Exxon Nuclear Company (ENC)), 8x8 current fuel. The fuel design of this reload batch is virtually identical to the fuel design of the pre-vious reload batch. This report summarizes the reload analyses performed by ANF in support of WNP-2 operation for Cycle>3. In addition, a de-scription of the ANF reload is given along with a comparison of the char-acteristics of the Cycle 3 and Cycle' cores. A discussion of the pro-posed physics startup program is also included. The proposed license amendment (technical specification changes) are listed by title in this report for completeness.

The reload licensing submittal is composed on the WNP-2 Cycle 3 Reload Analysis Report (XN-NF-87-25) (Reference 1.0), the WNP-2 Cycle 3 Plant Transient Analysis Report (XN-NF-87-24) (Reference 2.0), the proposed changes to the WNP-2 Technical Specifications and this report. Where appropriate, this report summarizes analyses and makes reference to the above reports and other documents for detailed support. The WNP-2 Cycle 3 Reload Analysis Report (Reference 1.0) is intended to be used in con-Junction with ENC Topical Report XN-NF-80-19(P)(A), Volume 4, Revision Application of the ENC Methodology to BWR Reloads (Reference 3.0), which gives a detailed description of the methods and analyses utilized.

2.0 GENERAL DESCRIPTION OF RELOAD SCOPE For the second refueling outage for WNP-2, the Supply System will replace 148 of the General Electric (GE) initial core fuel assemblies with ANF 8x8C fuel. The 148 ANF Bx8C fuel bundles to be loaded for Cycle 3 (Ref-erence.4.0) are similar in desigri to the initial core fuel. However, the change in WNP-2 core loading requires a re-analysis by ANF. Much of this analysis, particularly the Loss of Coolant Accident (LOCA) and the Maxi-mum Average Planar Linear Heat Generation Rate (MAPLHGR) are given in Reference 4.0 as these analyses were performed for all ANF fueled cores as a part of the Cycle 2 analysis. Relevant transient analyses and Mini-mum Critical Power Ratio (MCPR) analyses for the Cycle 3 loading are reported here. Analyses of normal reactor operation consisted of evalua-tion of the mechanical, thermal hydraulic, and nuclear design character-istics. Operation at extended core flow's also addressed. Q A number of proposed changes to the WNP-2 Technical Specifications have resulted from the ANF design and safety analyses for the Cycle 3 core. A list of, these Technical Specification changes is given in Table 2.1.

TABLE 2.1 PROPOSED TECHNICAL SPECIFICATION CHANGES 2.0 Safety Limits and Limiting Safety System Setting (Introduction) 2.1.2 Thermal Power, High pressure, and High Flow 3/4.1.3.4 Four Control Rod Group Scram Insertion Times 3/4.2.1 Average Planar Linear Heat Generation Rate 3/4.2.3 Minimum Critical Power Ratio 3/4.2.4 Linear Heat Generation Rate 3/4.3.10 ,Neutron Flux Monitoring Instrumentation 8 3/4.1.3 Control Rods B 3/4.2.1 Average Planar Linear Heat Generation Rate B 3/4.2.3 Minimum Critical Power Ratio 8 3/4.7.9 Main Turbine Bypass Systems 3.0 WNP-2 CYCLE 3 OPERATING HISTORY WNP-2, a 3323 mwt BWR 5, began Cycle 2 operation on June 10, 1986. The end of Cycle 2 operation is expected to be April 13, 1987.

During Cycle 2, the plant was base loaded at or near 100 percent power for the first five months of the cycle. At this point, excessive vibra-tion was observed in Recirculation Pump A. This pump was shut down at that point and operation continued in single loop. In single loop opera-tion, WNP-2 is limited to approximately 72 percent power. WNP-2 will continue to operate in single loop for the remainder of Cycle 2.

Figure 3.1 gives a power history of Cycle 2 through March 12, 1987,'or WNP-2. The Cycle 2 operating highlights and control rod sequence ex-change schedule are found in Table 3. 1.

l WNP-2 POWER HISTORY (CYCLE 2) 199. 8 FPD OF RATEO THERMAL POWER 118.

i POWER 68, 58

'8.

28.

18, JUN JUL AUG OCT NOV OEC JAN FEB MAR APR JUNE 18 86 TO MARCH 12 87 Figure 3, 1 Power History For WNP-2 For Cycie 2

S J TABLE 3.1 NNP-2 CYCLE OPERATING HIGHLIGHTS Began Fuel Loading April 18, 1986 Began Commercial Operation June 10, 1986 Projected End of Cycle Date April 13, 1987 u

End of Cycle Core Average Exposure (Design)(mwd/mtm) 12,153 Number of Fresh Assemblies 128 Gross Generation (FPD) (through March 18, 1987) 204. 1 Control Rod Se uence Exchan e Schedule

~Se 'uence Date From To August 14, 1986 A2 B2 September 26, 1986 B2 Al November 10, 1986 Al B1 January 10, 1987 Bl A2 February 26, 1987 A2 A1

4.0 RELOAD CORE DESCRIPTION The WNP-2 core consists of 764 fuel assemblies. For the Cycle 3 reload, the core will consist of 148 ANF 8xSC fresh assemblies, 128 ENC Sx8C XN-1 fuel assemblies loaded for Cycle 2 and 488 GE 8x8RP assemblies remaining from the initial core. The 148 ANF SxSC fresh assemblies consist of 36 reload assemblies originally manufactured for loading in Cycle 2 and 112 reload assemblies manufactured for loading in Cycle 3. The two assem-blies are 'dentical in uranium dioxide (U02) enrichment, gadolinium oxide (GD203) loading, and in all other major physical characteris-tics. Minor differences, primarily in end plug design, exist between the two assembly designs. However, the two assembly designs are interchange-able with regard to all of the analyses reported here. Table 4.1 lists the assembly >pe, quantity, and initial enrichment for the assemblies which will make up the Cycle 3 core.

TABLE 4.1 WNP-2 CYCLE 3 CORE Number of Assemblies T e Enrichment 148* ANF Sx8C 2.72 w/o U-235

] 28~ ENC SxSC 2.72 w/o U-235 432 GE SxSRP 2.19 w/o U-235 56 GE Sx8RP 1.76 w/o U-235 The 148 exposed GE Sx8RP assemblies discharged are all medium enriched (1.76 w/o U-235) assemblies.

• Thirty six (36) of these assemblies were originally fabricated for reload in Cycle 2 and 112 of these were fabricated for reload in Cycle

3. They are effectively identical.

~Two of these assemblies are Lead Test Assemblies (LTA) described in Reference 4.0.

5.0 FUEL MECHANICAL DESIGN I

The mechanical design of the 8x8C Cycle 3 ANF reload fuel for WNP-2 is described specifically in Reference 5.0 and more generically in Reference 6.0 and 7.0. This fuel is essentially identical to the 8xBC Cycle 2 ENC fuel described in Reference 4.0. The fuel assembly design uses 62 fuel rods and two centrally located water rods, one of which functions as a spacer captur e rod. Seven spacers maintain fuel rod pitch. The design uses a quick-removable upper tie plate design to facilitate fuel inspec-tion and bundle reconstitution of irradiated assemblies. The fuel rods utilize Zircaloy-2 cladding, 35 mils thick. The fuel rods are pressur-ized, and contain either U02 - GD20q or UOy with a nominal den-sity of 94.5 percent TD, and an 8.5 mi"f nomina diametrical pellet to clad gap for the enriched pellets. Natural uranium is loaded in the top and bottom six inches of each fuel rod for greater neutron economy. The enriched pellets have a slightly larger diameter than the natural pellets.

The fuel mechanical design analysis performed on the ANF 8x8C Cycle 3 reload fuel evaluated the following items in Reference 8.0:

o Cladding steady state strain and stress.

o Transient strain and stress.

o Cladding fatigue damage.

o Creep collapse.

o Corrosion.

o Hydrogen absorption.

o Fuel rod internal pressure.

o Differential fuel rod growth.

o Creep bow.

o Grid. space design.

The analyses presented in Reference 8.0 justify irradiation to a 35,000 NWD/MT peak assembly burnup in WNP-2.

Some major results of these analyses are:

o The maximum end-of-life (EOL) steady state cladding strain is well below the 1 percent design limit.

o Cladding steady state stresses are calculated below the material strength limits.

o The transient strain does not exceed 1.0 percent.

S l

o The cladding fatigue usage factor is within the 0.67 percent design 1 imi t.

o The cladding diameter reduction due to uniform creepdown, plus creep ovality at maximum dens'ification, is less than the minimum initial gap. Compliance with this, criteria prevents the formation of fuel column gaps and the possibility of creep collapse.

o The maximum level of the corrosion layer was calculated to be well within the design limit.

o The maximum concentration of hydrogen was calculated to be well within the design limit.

o Evaluations of the fuel assembly growth and differential fuel rod work show that the fuel assembly design provides adequate clearance.

o The plenum spring complies with design limits.

o The spacer spring meets all design requirements.

o The maximum fuel rod internal rod pressure remains below ANF's cri-teria limit.

o The fuel centerline temperature remains below the melting point.

The structural response of the 8xSC Cycle 3 ANF reload fuel is the same as the structural response of the 8xSC Cycle 2 ENC fuel and the 8xSRP GE fuel which also reside in the MNP-2 core. As a part of Cycle 3 opera-tion, some of the 8xSC Cycle 3 ANF reload fuel assemblies will be chan-neled with new 100 mil channels fabricated by ASEA Atom. These channels are equivalent to the initial core channels. Therefore, the seismic LOCA structural response evaluation performed in support of the initial core remains applicable and continues to provide assurance that control blade insertions will not be inhibited following occurence of the design basis seismic LOCA event.

A LHGR limit will be placed on ANF 8xSC Cycle 3 reload fuel assemblies for monitoring for the reasons given previously in Reference 4.0, Page 10, for ENC SxSC Cycle 2 fuel.

6.0 THERMAL HYDRAULIC DESIGN The goal of the thermal hydraulic design analysis is to demonstrate that the ANF reload fuel meets and/or exceeds the primary thermal hydraulic design criteria. Principal design cri teria considered in the thermal hydraulic analysi s ar e found in XN-NF-80-19(A), Yolume 4, Revi sion 1 (Reference 3.0).

Analyses performed to demonstrate that these criteria are met include:

o Hydraulic compatabil ity.

o Fuel cladding integrity safety limit.

Fuel centerline temperature.

o Bypass flow characteristics.

o Thermal hydraulic stability.

These analyses are discussed in this section.

6.1 Hydraulic Com atabilit The hydraulic flow resistances for the ANF reload fuel and the GE SxS fuel have been determined in single phase flow tests of full scale assemblies. XN-NF-80-19(A), Volume 4, Revision 1 (Reference 3.0), reports the resistances measured and evaluates the effects on thermal margin of mixed ANF and GE SxS cores. The close geometrical similarity between the two fuel designs and their measured perform-ance characteristics demonstrate that the two fuel designs are sufficiently compatible for co-residence in MNP-2.

6.2 Fuel Cladding Inte rity Safety Limit The MCPR fuel cladding integrity safety limit for MNP-2 is 1.06 which is equal to the Cycle 1 and Cycle 2 MCPR safety limit. The methodology used in the MCPR safety limit calculations is found in XN-NF-80-19(A), Volume 4, Revision 1 (Reference 3.0). The MNP-2 Cycle 3 MCPR safety limit analysis methodology and input parameters are described in XN-NF-87-24, Cycle 3 Plant Transient Report (Refer-ence 2.0).

6.3 Fuel Centerline Tem erature The LHGR curve in Figure 3.4 of Reference 8.0 shows that the ANF 8xSC fuel centerline temperature is protected for 120 percent over power. The LHGR curve in Reference 8.0 is everywhere greater than 120 percent, of the LHGR limit curve in Reference 6.0. Therefore, fuel centerline melt is protected for all ANF Sx8 exposures within the bounds of the referenced LHGR curve.

6.4 By ass Flow Characteristics Core bypass flow was computed using the methodology of XN-NF-524(A)

(Reference 9.0). The bypass flow for the MNP-2 Cycle 3 is 11.6 per-cent of the total core flow which is similar to the Cycle 1 value of 11.8 percent and identical to the Cycle 2 value of 11.6 percent.

The computed bypass flow will have no adverse impact on reactor operation..

I U'

F'

6.5 Thermal Hydraulic Stabil ity The WNP-2 Technical Specifications included surveillance require-ments for detecting and suppressing power oscillations. In addi-tion, the ANF COTRAN code (Reference 10.0) was used to specifically determine that the worst case value of decay ratio is less than 0.60 in the area of the power flow map bounded by the APRM rod block line at 45 percent rated flow. The worst case decay ratio is no greater than 0.9 in the area of allowable low flow operation (detect and suppress region). The bounding power flow points in the detect and suppress region are the APRM rod block line at 27.6 percent core flow (48 percent power - minimum allowable two pump flow) and the APRM rod block line at 23.8 percent core flow (42 percent power-natural circulation) (Reference 11.0).

7.0 NUCLEAR DESIGN The neutronic methods for the design and analysis of the WNP-2 Cycle 3 reload are described in Reference 10.0. These methods have been reviewed and approved by the U.S. Nuclear Regulatory Commission for generic appli-cation to BWR reloads.

7.1 Fuel Bundle Nuclear Desi n The Cycle 3 ANF reload bundles are identical to the Cycle 2 ANF reload bundles in nuclear design. Major nuclear design characteris-tics for the ANF 8x8C reload fuel assembly are:

o The fuel assembly contains 62 fuel rods and two water rods.

One of the water rods also acts as a spacer capture rod.

o The fuel assembly average enrichment is 2.72 w/o U-235. The top and bottom six inches of the fuel rods contain natural uranium. The central 138 inch portion of the fuel rods has an average enrichment of 2.89 w/o U-235.

o Five enrichment levels are utilized in the fuel assembly to produce a local power distribution which results in a balanced design for Minimum Critical Power Ratio (MCPR) and Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) 1imits.

o Each fuel assembly contains five fuel rods with 2.0 w/o GD203 blended with 2.57 w/o U-235 enriched U02 to reduce initial assembly reactivity.

The enrichment distribution of the ANF reloa'd design was selected on

,the basis of maintaining a balance between the local power peaking factors, assembly reactivity, MAPLHGR, and MCPR. For the central enriched region of the assembly, three rods are enriched to 1.5 w/o U-235, seven rods to 2.0 w/o U-235, nine rods to 2.57 w/o U-235, 16 rods to 2.94 w/o U-235, 22 rods to 3.54 w/o U-235, and five rods to 2.57 w/o U-235 plus 2.00 w/o GD203.

7.2 Core Nuclear Desi n The core exposure for the end of Cycle 2 (EOC2), the core exposure for the beginning of Cycle 3 (BOC3), and the core exposure for the end of Cycle 3 (EOC3) were calculated with the XTGBNR Code (Reference 10.0). In addition, BOC core reactivity characteristics for the cold core were calculated along with the standby liquid control system reactivity. Some of the results of these analyses are shown in Table 7.1.

Table 7.1 CORE NUCLEAR DESIGN Core Exposures at EOC2 (mwd/mtm) 12,153 Core Exposures at BOC3 (mwd/mtm) 9,639 Core Exposures at EOC3 (mwd/mtm) 15,103 BOC Cold Keff, all rods out 1.1257 BOC Cold Keff, strongest rod out 0.9882 Reactivity Defect/R-Value, percent a K/K 0.0 Standby Liquid Control System (SBLC)

Reactivity, 660 PPM Boron, Keff 0.9722 7.3 Com arison of Major Core Parameters Some of the major core parameters for WP-2 Cycle 2 and Cycle 3 are listed in Table 7.2.

Table 7.1 COMPARISON OF MAJOR CORE PARAMETERS Parameter Cycle 2 Cycle 3 MCPR Limit (0 mwd/mtm) 1.28 1.29 Doppler Defect 9.5 X 10-6 - 9.5 X 10-6 (hK~ /K~ hT)

Cycle Length (flesign; FPD) 255 227 Core Average Exposure 7,424 9,639 (BOC; mwd/mtm)

Core Average Exposure 12,153 15,103 (EOC; mwd/mtm)

The differences between the Cycle 2 core and the Cycle 3 core are found in the core loading pattern. The Cycle 2 core consisted of a scatter load of 204 GE 8x8 medium (1.76 w/o U-235) and 432 high en-riched (2.19 w/o U-235) bundles and 128 ANF 8x8 reload (2.72 w/o U-235) bundles. The Cycle 3 core will consist of a scatter load of 52 GE 8xBR medium enriched bundles, 436 GE 8x8R high enriched bun-dles, 128 ANF 8x8C reload bundles with one cycle of exposure and 148 ANF 8x8C fresh reload bundles.

8. 0 ANTICIPATED OPERATIONAL OCCURRENCES ANF considers eight categories of potential system core wide transient occurrences for jet pump BMRs in Reference 12.0. ANF has provided an-alysis results for the three most limiting transients for WNP-2 Cycle 3 to determine the Cycle 3 thermal margins. The three transients deter-mined to be most limiting for Cycle 3 are:

o Load Rejection Mithout Bypass (LRMB).

o Feedwater Controller Failure (FMCF).

o Loss of Feedwater Heating (LOFH).

ANF's methodology for developing thermal limits is found in Reference 13.0. Reference 12.0 demonstrates that the other plant transient events are inherently nonlimiting or clearly bounded by the above events.

Two local events, Control Rod Withdrawal Error (CRME) and Fuel Loading Error (FLE) were analyzed with the methodology described in Reference 10.0. The CRWE was demonstrated to be bounding for certain parts of the fuel cycle.

The results of the core-wide and local transient analyses are provided in the WNP-2 Cycle 3 Reload Analysis Report (Reference 1.0) and in the MNP-2 Cycle 3 Transient Analysis Report (Reference 2.0). The CRME was evalu-ated and found to be most limiting up to EOC-2000 mwd/mtm at 106 percent of rated core flow, resulting in a d CPR of 0.20 for the ANF fuel and 0.23 for the GE fuel at the 106 percent rod block monitor (RBM) trip set-point. Mhen combined with the 1.06 safety limit, this transient (CRME) requires a MCPR operating limit of 1.26 for the ANF fuel and 1.29 for the GE fuel in Cycle 3 in the range from BOC to EOC-2000 mwd/mtm. The ANF reload safety analyses were performed using control rod insertion times based on plant data. For operation in the range of EOC-2000 mwd/mtm to EOC up to 106 percent core flow with these normal scram times, the FMCF transient was determined to be the limiting transient and the NCPR limit for ANF fuel is 1.30 and for GE fuel is 1.32 for this portion of the fuel cycle. In the event that plant surveillance demonstrates that these scram insertion times are exceeded, the plant thermal margins default to values which correspond to the Technical Specification insertion times (3. 1.3.4, P 3/4.1.7) for this portion of the fuel cycle (EOC-2000 mwd/mtm to EOC). For operation at EOC-2000 with core flow up to 106 percent and these technical specification scram times, the limiting transient is the

( I LRWB transient and the MCPR operating limit within EOC-2000 mwd/mtm to EOC is 1.35 for ANF fuel and 1.39 for GE fuel for Cycle 3 operation.

the Recirculation Pump Trip (RPT) should become inoperable for any reason If and assuming normal scram speeds, and operation up to 106 percent core flow, the limiting transient is then the FMCF transient and the MCPR operating limit is 1.35 for ANF fuel and 1.37 for GE fuel. Finally, the RPT becomes inoperable within EOC-2000 mwd/mtm to EOC and the plant if defaults to technical specification scram times, the LRMB transient at 106 percent flow is bounding and the MCPR operating limit is 1.39 for ANF fuel and 1.43 for GE fuel.

Additional analyses were performed to determine the MCPR operating limit with a 107 percent and 108 percent RBh1 setpoint for the CRME event. The resulting a CPRs are 0.20 for ANF fuel and, 0.23 for GE fuel at 107 per-cent, 0.22 for ANF fuel, and 0.25 for GE fuel at a 108 percent rod block setting. 'herefore, operation with a 108 percent PBM setting would require a MCPR limit of 1.28 for ANF and 1.31 for the GE fuel.

8.1 Core'Wide Transients The plant transient model used to evaluate the pressurization tran-sients, the LRWB and FWCF events, consists of the ANF COTRANSA (Ref-erence 12.0) and XCOBRA-T (Reference 14.0) codes. This axial one-dimensional model predicted reactor power shifts toward the core middle and top as pressurization occurred. This phenomenom was accounted for explicitly in determining thermal margin changes in the transient. All pressurization transients were analyzed on a bounding basis using COTRANSA in conjunction with the XCOBRA-T hot channel model. The FWCF event was found to be the most limiting core wide event at 106 percent core flow at EOC utilizing normal scram times. For technical specifications scram times, the LRWB event was found to be the most limiting core wide event at 106 per-cent core flow and EOC. With RPT inoperable and normal scram times, the FWCF event was found to be the most limiting core wide event at 106 percent core flow and EOC. With RPT inoperable and technical specification scram times, the LRMB was found to be the most limit-ing transient at 106 percent core flow and EOC. All core wide tran-sients were analyzed using bounding values as input.

The, Loss of Feedwater Heating (LOFH) transient was analyzed on a generic basis for a wide cross section of BMR configurations. This generic analysis is documented in Reference 15.0. This analysis provides a statistical evaluation of the consequences of the LOFH transient for BWR/4, BWR/5, and BMR/6 plant configurations under conditions which cover the operating power flow map including in-creased core flow conditions. At this time, the generic analysis is under review. A conservative bounding value of a h CPR of 0.09 is supported by the analysis'esults for plants with MCPR safety limit of 1.06. The MNP-2 MCPR safety limit for Cycle 3a continues to be 1.06 (Reference 2.0). Ther'efore, the LOFH transient requires a t'tCPR operating limit of 1.15 for WNP-2.

l I If f'

8.2 Local Transients Analysis given in Reference 1.0 show that the FLE transient is bounded by the CRWE transient and is therefore nonlimiting. Based on the CRWE results, the MCPR operating limit is a function of the RBM setpoint. Analyses were performed to support a RBM setpoi nt of 106 percent, 107 percent, and 108 percent. The a CPR for the CRWE with a 106 percent RBM setpoint is 0.20 for ANF fuel and 0.23 for GE fuel, for a 107 percent RBM setpoint 0.20 for ANF fuel and 0.23 for GE fuel, and for a 108 percent RBM setpoint 0.22 for ANF fuel, and 0.25 for GE fuel.

8.3 Reduced Flow 0 eration The recirculation flow run-up analysis performed for WNP-2 Cycle 2 was reviewed and the assumptions and conditions used for Cycle 2 are applicable to Cycle 3. Thus, the reduced flow MCPR operati ng limit for WNP-2 Cycle 2 is applicable to Cycle 3.

8.4 ASME Over ressurization Anal sis In order to demonstrate compliance with the ASME Code over pressuri-zation criteria of 110 percent of vessel 'design pressure, the Main Steam Isolation Valve (MSIV) closure event with failure of the MSIV position switch scram was analyzed with ANF's COTRANSA code (Refer-ence 12.0). The WNP-2 Cycle 3 analysis assumed six safety relief valves out of service. The maximum pressure observed in the an-alysis is 1313 psig in the vessel lower plenum. This is 105 percent of the reactor vessel design pressure which is well below the 110 percent design criterion.

The calculated steam dome pressure corresponding to the 1313 psig peak vessel pressure is 1285 psig, for a vessel differential pres-sure of 28 psig. The RPT is assumed to initiate at a pressur e set-point of 1170 psig. The current Technical Specification Safety limit of 1325 psig is based on dome pressure and therefore conserva-tively assumes a 50 psi vessel dp (1375-1325). Since the calculated vessel differential pressure is 28 psi, the steam dome safety limit of 1325 psig assures compliance with the ASME criterion of 1375 psig peak vessel pressure.

8.5 Increased Flow 0 eration The plant system transient events reported earlier in this document, which are potentially limiting for MCPR, were all analyzed at in-creased core flow of 106 percent. The Cycle 2 transient events analyzed at the design basis power condition with increased core flow were found to bound the same transients analyzed at the design basis power and rated flow condition for WNP-2 Cycle 2 (Reference 16.0).

I ANF has also performed analyses which demonstrate that the XN-1 8x8C fuel bundle can operate satisfactorily from a mechanical standpoint at,.this increased core flow (Reference 17.0). In addition, GE has performed analyses for the reactor internals and for the GE fuel assembly which. considered the loads created by operation at this flow level and the impacts of'hese loads on the WNP-2 core inter-nals and the GE fuel assembly. Also, flow induced vibration of the core internal s as a resul t of increased core fl ow was analyzed.

Finally, analyses were performed for feedwater nozzle and feedwater sparger fatigue at increased core flow. The results of all these analyses when considered along with the similarity between the two fuel types utilized in Cycle 3, confirm the capability of WNP-2 to operate at 100 percent power and 106 percent core flow during Cycle 3 operation (Reference 18.0).

A containment analysis was performed ,.to determine the impact of operation at increased core flow on the WNP-2 containment LOCA re-sponse. The results show that the containment LOCA response for increased core flow operation is bounded by the corresponding FSAR results (Reference 19.0).

In summary, all relevant neutronic, thermal hydraulic, mechanical, and safety analyses have been- performed to demonstrate that WNP-2 can operate safely with extended core flow up to 106 percent, of rated core flow during Cycle 3.

8.6 Sin le Loo 0 eration The NSSS Supplier, GE, has provided analyses which demonstrate the safety of WNP-2 operation with a single recirculation loop in -ser-vice for an extended period of time (Reference 20.0). Because the ANF fuel is designed'o be compatible'ith the GE fuel and because the ANF methodology gives results consistent with GE for two loop operation, the GE single loop analysis is also applicable to ANF 8x8 fuel. 'With a single loop in operation, the GE analysis supported operation with an increase of 0.01 in the MCPR safety limit. Due to compati bility, this increase is also appropriate for ANF Bx8 fuel.

For Cycle 3 operation of WNP-2 with a single loop in service, the MCPR safety limit is 1.07 which is the same value as the previous cycle.

The consequences of core wide transients for single loop operation are bounded by the consequences of these events at rated conditions.

The additional conservatism imposed by the reduced flow MCPR operat-ing limits assure that the MCPR safety limit will not be violated during single loop operation. Because the reduced flow MCPR limit curves are based on equipment performance which cannot physically happen during single loop operation, the added conservatism present in the curves compensates for the penalties associated with in-,

creased uncertainties in the MCPR limit and control rod drive per-formance. The reduced flow MCPR limit curves are applicable without modification during single loop operation.

l 9.0 POSTULATED ACCIDENTS For Cycle 2, ANF has analyzed the LOCA to determine MAPLHGR limits for ANF 8x8 fuel. The results of this analysis is presented Reference 21.0.

These results are equally applicable to Cycle 3. ANF's methodology for the LOCA analysis is given in References 22.0, 23.0, and 24.0. In addi-tion, the Rod Drop Accident (RDA) was analyzed to demonstrate compliance with the 280 cal/gm design limit. ANF's methodology for the RDA an-alysis can be found in Reference 10.0.

9.1 Loss of Coolant Accident Reference 25.0 describes ANF's WNP-2 LOCA break spectrum analysis which defined the limiting break for WNP-2. The analysis of this event for WNP-2 is described in Reference 26.0. The LOCA analysis described in Reference 26.0 was performed for an entire core of ANF 8x8C fuel and therefore provides MAPLHGR limits for ANF fuel only.

These results are applicable to operation in WNP-2 Cycle 3.

ANF 8x8C fuel is hydraulically and neutronically compatible with the GE initial core fuel. Therefore, the existing GE LOCA analysis and MAPLHGR limits are applicable to GE initial core fuel during Cycle 3 and future cycles with mixed GE/ANF cores.

9.2 Rod Dro Accident ANF's methodology for analyzing the RDA is given in Reference 10.0.

For WNP-2 Cycle 3, the analysis shows a value of 170 cal/gm for the max'imum deposited fuel rod enthalpy during the worst case postu-lated RDA (Reference 1.0). This is well below the design limit value of 280 cal/gm.

9.3 Single Loo Operation To support operation of WNP-2 with a core composed of GE Cycle 1 fuel and ANF 8x8 fuel with a single recirculation pump operating, ANF recommends the conservative use of GE MAPLHGR limits for the GE fuel design with a multiplier of 0.84 applied for single loop oper-ation. The analytical limits used by GE have yielded conservative MAPLHGR limits relative to the MAPLHGR limits obtained using the approved ANF analytical methods. The phenomena which requires the reduction in MAPLHGR limits for single loop operation are common to both fuels. Therefore, applying the more conservative GE MAPLHGR limits to ANF fuel assures conformance with the criteria of 10CFR50.46.

10.0 STARTUP PHYSICS TEST PROGRAM The Supply System has developed a restart physics test program to be carried out prior to initiation of Cycle 3. This program includes a core loading verification test, a control rod functional test, a in se-quence shutdown margin test, and a TIP asymmetry test. The proposed test goals and a brief description of each test is given below.

Core Load Verification Test Goal - To assure that the WNP-2 Cycle 3 Core is loaded according to the design analyzed by ANF.

Test Descri tion - This test will be performed with the aid of a te evision camera mounted on the fuel mast. A series of initial passes will be made with the television camera/mast set at a pre-determined hei ght to assure that all fuel assemblies are ful ly seated in the core. Then, with the aid of the camera and a visual readout on the refuel floor, the assembly serial numbers, their orientation and location will be visually checked and recorded on video tape. Subsequently, a review of the tapes will be made to check the initial verification.

Control Rod Functional Test Goal - To determine and verify control rod mobility and functionality.

~TtD each i i -Fi1 ighpiti cell of four fuel assemblies, fl'11 dig, f the control blade for that cell will be fully withdrawn and inserted. This will demonstrate the mobility of that blade, the absence of overtravel for that blade and the fact that the lattice is subcritical with that blade with-drawn. This in turn will verify that there are no gross reactiv-ity discrepancies between the actual core and the analyzed design.

After the core is fully loaded, verify that the control rod drive insertion and withdrawal times are within design specifications and technical specification limits. This action will also verify that the core is subcritical with any single rod fully withdrawn.

Subcri ti cal thar in Test Goal - To assure that the Technical Specification shutdown margin requirement is satisfied.

Test Description - The data is taken during a normal insequence startup criticality. Critical control rod positions are obtained and corrected for reactor period and moderator temperature coeffi-cient effects. The results are compared to predicted control rod positions and from this information, the shutdown margin with the analytically determined strongest control rod withdrawn is confirmed.

TIP Asymmetr Test Goal - To assure proper TIP systems operation and to verify that ttte TIP system uncertainty is within the limits assumed for tran-sient analysis.

Test Descri tion - This test is performed in the power range pre-era y a ove percent power. An octant symmetric control rod pattern is utilized. Data is gathered from all available TIP locations, and the total average uncertainty is determined for all symmetric TIP pairs.

11.0 REFERENCES

1.0 XN-NF-87-25, "WNP-2 Cycle 3 Reload Analysis Report", Advanced Nu-clear Fuels Corporation, March 1987.

2.0 XH-NF-87-24, "WNP-2 Cycle 3 Plant Transient Analysis Report", Ad-

.vanced Nuclear Fuels Corporation, March 1987.

3.0 XN-HF-80-19(A), Volume 4, Revision 1, "Exxon Nuclear Methodology For Boiling Water Reactor: Application of the EHC Methodology to BWR Reloads", Exxon Nuclear Company, September 1983.

4.0 WPPSS-C-ANF-101, "WHP-2 Cycle 2 Reload Summary Report", February 1986.

5.0 XN-NF-86-159(P), Revision 0, "Washington Public Power Supply System, WNP-2 Reload XN-2 (WPB2), Cycle 3 Design Report, Exxon Nuclear Company, November 1986.

6.0 XN-NF-81-21(A)', Revision 1, "Generic Mechanical Design For Exxon Nuclear Jet Pump BWR Reload Fuel", Exxon Nuclear Company, Septem-ber 1982.

7.0 XN-NF-81-21(A), Revi sion 1, Suppl ement 1, "Generi c Mechanical Design For Exxon Nuclear Jet Pump BWR Reload Fuel", Exxon Nuclear Company, March 1985.

8.0 XN-HF-85-67(A), Revision 1, "Generic Mechanical Design For Exxon Nuclear Jet Pump BWR Reload Fuel", Exxon Nuclear Company, July 1985.

9.0 XN-NF-524(A), Revision 1, "'Exxon Nuclear Critical Power Methodol-ogy For Boiling Water Reactors", Exxon Nuclear Company, November 1983.

10.0 XN-HF-80-19(A), Volume 1 and Volume 1 Supplements 1 and 2, "Exxon Nuclear Methodology For Boiling Water Reactors: Neutronic Methods For Design and Analysis", Exxon Nuclear Company, November 1981.

11.0 ANF Letter No. ANFWP-87-0046, J. B. Edgar to Manager, Central Con-tracts, dated March 25, 1987.

12.0 XN-NF-79-71(P), Revision 2 (as supplemented), "Exxon Nuclear Power Plant Transient Methodology", Exxon Nuclear Company, November 1981.

13.0 XN-NF-80-19(A), Volume 3, Revision 2, "Exxon Nuclear Methodology For Boiling Water Reactors: THERMEX Thermal Limits Methodology Summary Descriptions", Exxon Nuclear Company, January 1987.

l O

14. 0 XN-NF-84-105(A), Volume 1, Volume 1 Supplement 1, Volume 1 Supple-ment 2, "XCOBRA-T: A Computer Code For BWR Transient Thermal

-Hydraulic Core Analysis", Advanced Nuclear Fuels Corporation, February 1987.

15.0 XN-NF-900(P), "A Generic Analysis of the Loss of Feedwater Heating Transient For Boiling Mater Reactors", Exxon Nuclear Company, February 1986.

16.0 B. Edgar, Letter to MPPSS, Supplemental Analysis Results, EHNP-86-0067, Exxon Nuclear Company, April 15, 1986.

~

17.0 B. Edgar, Letter to'.

WPPSS, ENHP-86-0033, Exxon Nuclear Company, February 13, 1986.

18.0 HEDC-31107, "Safety Review of WPPSS Nuclear Project No. 2 at Core Flow Conditions Above Rated Flow Throughout Cycle 1 and Final Feedwater Temperature Reduction", General Electric Company, Febr uary 1986.

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

20.0 G. C. Sorensen (Supply System) to A. Schwencer (NRC), Letter No.

G02-83-814, September 8, 1983.

21.0 XN-HF-86-01, "WNP-2 Cycle 2 Reload Analysis Report", Exxon Nuclear Company, January 1986.

22.0 XN-NF-80-19(A), Volumes 2, 2A, 2B, and 2C, "Exxon Nuclear Method--

ology For Boiling Mater Reactors: EXEM ECCS Evaluation t1odel",

Exxon Nuclear Company, September 1982.

23.0 XN-NF-CC-33(A), Revision 1, "HUXY: A Generalized Multirod Heatup Code With 10CFR50, Appendix K, Heatup Option", Exxon Nuclear Com-pany, November 1975.

24.0 XN-NF-82-07(A), Revision 1, "Exxon Nuclear Company ECCS Cladding-Swelling and Rupture Model", Exxon Nuclear Company, November 1982.

25.0 XN-NF-85-138(P), "LOCA Break Spectrum Analysis for a BMR 5", Exxon Nuclear Company, December 1985.

26.0 XN-NF-85-139, "MNP-2 LOCA-ECCS Analysis NAPLHGR Results", Exxon Nuclear Company, December 1985.