ML17285A302
| ML17285A302 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 02/28/1989 |
| From: | Humphreys M, Petraglia J, Wolenhauer W WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | |
| Shared Package | |
| ML17285A298 | List: |
| References | |
| TAC-72251, WPPSS-EANF-124, NUDOCS 8903090455 | |
| Download: ML17285A302 (56) | |
Text
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Technical Report s No. NPPSS-EANF-lay HNP-2 CYCLE 5 RELOAD SUHHARY REPORT Hashington Public Power Supply Rich/and, Nashington System February 1989 8903090455 890m397 pg~ ~@@CD, 050~
PN U
WNP CYCLE 5 RELOAD
SUMMARY
REPORT Prepared By:
W. C. Wolkenhauer, Principal Engineer, Nuclear Fuel Reviewed By:
. P. Petragl a, Senior Reactor Engineer Reviewed By:
M. C. Humphr ys, ead Reactor Systems Engineer Concur With:
R. 0. Vosburgh, ager, Safety Analysis & Simulator Engineering Concur With:
R. 3. a bert, Supervisor, WNP-2 Reactor Engineering
~-
Concur With: 2(3
. L. K igs, Manager, WNP-2 Technical Approved By:
D. L.'hitcomb, Manager, Nuclear Fuel Approved By:
D. L. ar sn; anager; Engineenng Analysss uclear Fuel
NOTICE
'his report is derived in part through information provided to Washington Public Power Supply System (Supply System) by Advanced Nuclear Fuels Corporation. It is being submitted by the Supply System to the U.S. Nuclear Regulatory Commission in support of the WNP-2 Application For Technical Specifications Changes Relating to WNP-2 Cycle 5 operation. The information contained herein is true and correct to the best of the Supply System's knowledge, information, and belief.
WNP-2 CYCLE 5 RELOAD
SUMMARY
REPORT
1.0 INTRODUCTION
2.0 GENERAL DESCRIPTION OF RELOAD SCOPE 2.1 Summary'f Results . . .
3.0 WNP-2 CYCLE 4 OPERATING HISTORY 4.0 RELOAD CORE DESCRIPTION ~ ~
5.0 FUEL MECHANICAL DESIGN (BxBC FUEL).....
6.0 LEAD FUEL ASSEMBLY PROGRAM . ', 13 7.0 THERMAL HYDRAULIC DESIGN 20 7.1 Hydraulic Compatabi lity . . . . 20 7.2 Fuel Cladding Integrity Safety Limit 20 7.3 Fuel Centerline Temperature . 21
, 7.4 Bypass Flow Characteristics . 21 7.5 Thermal Hydraulic Stability . 21 8.0 NUCLEAR DESIGN . 21 8.1 Fuel Bundle Nuclear Design 21 8.2 Core Nuclear Design . 22 8.3 Comparison of Major Core Parameters . 23 9.0 ANTICIPATED OPERATIONAL OCCURRENCES 24 9.1 Core Wide Transients 25 9.2 Local Transients 26 9.3 Reduced Flow Operations . 26 9.4 ASME Overpressurization Analysis 27
9.5 Increased Flow Operation 27 9.6 Single Loop Operation . 28 9.7 Final Feedwater Temperature Reduction . 29 10.0 POSTULATED ACCIDENTS . 29 10.1 Loss of Coolant Accident 29 10.2 Rod Drop Accident . ~ ~ 30 10.3 Single Loop Operation . 30
.11.0 STARTUP PHYSICS TEST PROGRAM . 30 11.1 Core Load Verification Test . 30 11.2 Control Rod Functional Test . 31.
11.3 Subcritical Margin Test . 31 11.4 Tip Asymmetry Test 31
12.0 REFERENCES
31
HNP-2 CYCLE 5 RELOAD
SUMMARY
REPORT 1.0 The fourth reload of. the Washington Public Power Supply System Plant No. 2 (HNP-2) will utilize Advanced Nuclear Fuels Corporation (ANF) 8x8 fuel plus four Lead Fuel Assembly (LFA) bundles utilizing a 9x9 pin array. The 8x8 fuel design of this reload batch is virtually identical to the fuel design of the previous reload batch. The 9x9 array LFA bundles represent an advanced design described in 'more detail in Section 6.0 of this report. This report summarizes the reload analyses performed by ANF in support of HNP-2'peration for Cycle 5. In addition, a
,description of the ANF reload is given along with a comparison of the characteristics of the Cycle 4 and Cycle 5 cores. A discussion of the proposed physics startup program is also included. The proposed license amendments (technical specification changes) are listed by title in this report for completeness.
The reload licensing submittal is composed of the HNP-2 Cycle 5 Reload Analysis Report (ANF-89-02) (Reference 1.0), the HNP-2 Cycle 5 Plant Transient Analysis Report (ANF-89-01) (Reference 2.0), the proposed changes to the HNP-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 HNP-2 Cycle 5 Reload Analysis Report (Reference 1.0) is intended to be used in conjunction with ANF Topical Report XN-NF-80-19(P)(A), Volume 4, Revision 1, Application of the ANF Methodology to BWR Reloads (Refer-ence 3;0), which gives a detailed description of the methods and analyses utilized.
During the fourth refueling outage for WNP-Z, the Supply System will replace 144 of the General Electric (GE) initial core, fuel assemblies with ANF reload fuel. One hundred thirty-six (136) of the Cycle 5 reload fuel assemblies will be 8x8 current design reload assemblies which con-tain a bundle average enrichment of 2.62 weight percent U-235, four (4) of the Cycle 5 reload fuel assemblies will be 8x8 current design reload assemblies originally fabricated for use in Cycle 4 with a bundle average enrichment of 2.64 weight percent U-235 and four (4) of the Cycle 5 reload assemblies will be LFA's as described in Section 6.0 of this report. Two of the one hundred thirty six (136) 2.62 weight percent U-235 reload fuel assemblies have been manufactured with cladding having a composition controlled within a limited part of the ASTM specification for zircaloy-2. These assemblies are identical'n all aspects to the others in this reload batch. This change in HNP-2 core loading required a partial'e-analysis by ANF. The Loss of Coolant Accident (LOCA) and the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) analyses relevant to Cycle 5 operations are given in Reference 4.0 as these analyses were performed for all ANF fueled cores as a part of the Cycle 2 (initial reload) analysis. These analyses also bound operation of the
o
~ '
9x9 LFA fuel assemblies as discussed in Section 6.0. A MAPLHGR relation-ship for the 9x9 LFA's is included in the proposed technical specifica-tion change. The 9x9 MAPLHGR is based upon the above referenced MAPLHGR analysis for the 8x8 reload fuel and is different only to address the increased number of fuel rods in the 9x9 LFA bundles relative to the 8x8 reload fuel. A linear heat generation rate (LHGR) was developed for the 9x9 LFA fuel bundles and is included in the proposed Technical Specifi-cation changes. Relevant transient analyses and Minimum Critical Power Ratio (MCPR) analyses for the Cycle 5 loading are reported herein. These analyses identified the need for a revised reduced flow MCPR relationship for WNP-2, Cycle 5. This revised'educed flow MCPR. relationship is included in the proposed Technical Specification changes. Analyses of normal reactor operation consisted of evaluation -of the mechanical, thermal hydraulic, and nuclear design characteristics. Operation at extended core flow, single loop operation and final feedwater temperature reduction are also addressed.
A number of proposed changes to the WNP-2 Technical Specifications have resulted from the ANF design and safety analyses for the Cycle 5 core. A list of these Technical Specification changes is given in Table 2.1.
TABLE 2.1 TI HAN INDEX B2.1.2 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 B3/4.2.1 Average Planar Linear Heat Generation Rate 5.3 Reactor Core
- Editorial change only to reflect change submitted and approved as Technical Specification Amendment No. 28.
- Included in this submittal for convenience. This change corrects a previous oversight in the Tech. Specs.
e The limiting transient HCPR results for the analyses described in sections 8 and 9 of this document are summarized in columns 3 and 4 of Table 2.2.
WNP-2 will be entering its fifth cycle of operation and is approach-ing an'quilibrium cycle, Analysis results between Cycles 3, 4 and 5 have shown little change. For Cycle 4 operation, WNP-2 chose to add a small CPR penalty for margin to envelope future anticipated analysis results. The resulting CPR limits, including these self-imposed penalties, are summarized in columns 5 and 6 of Table 2.2.
The CPR limits listed for ANF fuel in column 4 are applicable to both the ANF 8x8 and ANF 9x9 fuel designs included in this reload.
The values listed in columns 3 and 4 of Table 2.2 form the basis for the requested changes to Tech. Spec. Table 3.2.3-1 for WNP-2 for Cycle 5 operations.
TABLE 2.2 MCPR OPERATING LIMITS
- 2. 3. 4, 5. 6.
Cycle-5 Cyc 1 e-4 Cyc 1 e Equipment An
~E~ 1 i R 1
~NET 1 GF~Fe3
~1 SKJ~
0-3750 MHD/MT 1.23 1.23 1.40 1.28 3750 EOC Normal SCRAM 1.34 1.31 1.40 1.31 times 3750 EOC TS SCRAM 1.41 1.37 1. 50 1.38 times 3750 EOC RPT inop 1.41 1.37 1.50 1.37 Normal SCRAM times 3750 EOC RPT inop 1.47 1.41 1.55 1.43 TS SCRAM times 0 - EOC Single Loop 1.35 1.35 1.40 1.37 Operation
- In this portion of the fuel cycle, operation with the given MCPR operating limits is allowed for both normal and Tech. Spec. scram times and for both RPT operable and inoperable.
30 ~W- Y PR WNP-2, a 3323 mwt BWR 5, began Cycle 4 operation on June 23, 1988*. The end of Cycle 4 operation is expected to be April 14, 1989. Ouring Cycle 4, the plant was base loaded at or near 100 percent power for all of the cycle.
Figure 3:1 gives a power history of Cycle 4 through January 1989, for WNP-2. The Cycle 4 operating highlights and contro1 rod sequence exchange schedule are found in Table 3.1.
" Sign)ficant thermal operation began on this date. Significant electrical generation commenced on June 28, 1988.
L 0 I ~ <a,.
oY Figure 3.1 Power History for l(HP-2 for Cycle 4
" 7-
TABLE 3.1 Began Fuel Loading Hay 10, 1988 Began Electric Power Production June 28, 1988 Projected End of Cycle Date April 14, 1989 End of Cycle Core Average Exposure (Design)(mwd/mtm) 16,870 Number of Fresh Assemblies 152 Gross Generation (FPD) (projected) 237 August 19, 1988 A2 B2 October 27, 1988 B2 Al
~ . December 30, 1989 Al Bl ~
February 22, 1989* Bl A2 August 24 through September 8, 1988 October 27 through October 29, 1988 December 1 through December 8, 1988 January 6 through January 9, 1989 January 30 through February 2, 1989
'Estimated date; also, may alternately choose to go to A-1 sequence,
4.0 The WNP-2 core consists of 764 fuel assemblies. For the Cycle 5 reload, the core will consist of 144 ANF fresh assemblies, 152 ANF 8x8C assem-blies loaded for Cycle 4, 148 ANF 8x8C assemblies loaded for Cycle 3, 128 ENC 8x8C fuel assemblies loaded for Cycle 2 and 192 GE SxSRP assemblies remaining from the initial core. The 144 ANF= fresh assemblies consist of four reload 8x8C assemblies originally manufactured for loading in Cycle 4 (ANF-3), 136 reload Sx8C assemblies manufactured for loading in Cycle 5 (ANF-4) and four 9x9 LFA reload assemblies. The four ANF-3 assemblies have a bundle average enrichment of 2.64 weight percent U-235 and the 136 reload ANF-4 fuel assemblies have a bundle average enrichment of 2.62 weight percent U-235. The four ANF-3 8x8C assemblies have five fuel rods containing gadolinium oxide (Gd203) at a loading of 2.0 weight per-cent as a neutron poison. The 136 ANF-4 8xSC assemblies have six fuel rods containing gadolinium oxide (Gd203) at a loading of 2.0 weight percent. Minor differences, primarily in end plug design, exist between the two assembly Sx8C designs. The four 9x9 LFA reload assemblies are described in detail in Section 6.0. The assembly designs are inter-changeable with regard to all of the analyses reported here. Table 4.1 lists the assembly type, quantity, and initial enrichment for the assem-blies which will make up the Cycle 5 core.
4 TABLE 4.1 Number of hm n 1'40* ANF 8x8C 2.62/2.64 w/o U-235 ANF 9x9 2.53/2.59 w/o U-235 1 52*% ANF Sx8C 2.64/2.72 w/o U-235 1
48*** ANF Sx8C 2.72 w/o U-235 1
28**** ENC SxSC 2.72 w/o U-235 136 GE PSx8R 2.19 w/o U-235 56 GE PSx8R 1.76 w/o U-235 The 144 exposed GE PSxSR assemblies to be discharged are all high enriched (2.19 w/o U-235) assemblies.
- Four (4) of these assemblies were fabricated for reload in Cycle 4 (ANF-3) and have an enrichment of 2.64 weight percent U-235 and one hundred thirty-six (136) of these assemblies were fabricated for reload in Cycle 5 (ANF-4)'nd have an enrichment of 2.62.weight percent. Two (2) of these ANF-4 assemblies contain characterized fuel pins.
- Twenty-four (24) of these assemblies were originally fabricated for reload in Cycle 3 (XN-2) and have an enrichment of 2.72 weight percent U-235 and one hundred twenty-eight (128) of these assemblies were fabricated for reload in Cycle 4 (ANF-3) and have an enrichment of 2,64 weight percent U-235. Two of these ANF-3 assemblies are Lead Test Assemblies (LTA) of standard ANF SxSC design.
- Thirty-six (36) of these assemblies were originally fabricated for reload in Cycle 2 (XN-1) and one hundred twelve (112) of these were fabricated for reload in Cycle 3 (XN-2). They are effectively identical.
5.0 The mechanical design of the 8x8C Cycle 5 ANF reload fuel for WNP-2 is described briefly in Reference 5.0 and more completely in References 6.0, 7.0 and 8.0. This fuel is essentially identical to the 8x8C 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 capture 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 mi ls thick. The fuel rods are pres-surized, and contain either U02 Gd203 or U02 with a nominal density of 94.5 percent of theoretical density (TD), and an 8.5 mil nominal 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 reload fuel evaluated the following items (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 designs The analyses presented in Reference 8.0 justify irradiation to a 35,000 MWD/NT 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.
o The cladding fatigue usage factor is within the 0.67 percent design limit.
o The cladding diameter reduction due to uniform creepdown, plus creep ovality at maximum densification, 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.
Evaluations of the fuel assembly growth and differential fuel rod growth show that the fuel assembly design provides adequate clearance.
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 criteria limit.
o The fuel centerline temperature remains below the melting point.
The structural response of the 8x8C ANF-4 reload fuel is the same as the structural response of the SxSC ANF-3 fuel, the SxSC XN-2 ANF fuel, the 8x8C XN-1 fuel and the P8xSR GE fuel which also reside in the WNP-2 Cycle 5 core. As a part of Cycle 5 operation, some of the Sx8C Cycle 4 ANF reload fuel assemblies may be channeled with new 100 mi channels fabri-1 cated by ASEA Brown Boveri (ABB) and/or Cartech as has been the practice in the past. These channels are equivalent to the initial core chan-nels. The remainder of the reload fuel bundles will be channeled with channels which have been previously discharged from WNP-2. Prior to reuse, these channels are measured with the WNP-2 channel measuring machine and qualified for reuse based upon a predetermined criteria.
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 occurrence of the design basis seismic LOCA event.
A LHGR limit is placed on ANF 8x8C Cycle 5 reload fuel assemblies for monitoring for the reasons given previously in Reference 4.0, Page 10, for ENC SxSC Cycle 2 fuel; 12
6.0 The lead fuel assembly program for Cycle 5 of WNP-2 consists of four fuel assemblies all of which utilize a 9x9 fuel pin array with an interior water channel which displaces the central 3x3 fuel pin configuration of the assembly and which functions as a spacer/capture device. Two of the LFA assemblies are designated 9x9-IX and two of the assemblies are desig-nated 9x9-9X. The 9x9-IX bundles incorporate the use of a zirconium liner within the cladding on all fuel pins except for those pins con-taining Gd203. These fuel pins are clad with beta quenched zircaloy cladding as is used in the rest of the Gd203 fuel pins in this reload. The 9x9-9X bundles do not utilize a pure zi'rconium liner and thus are made with standard ,zircaloy cladding except for the Gd203 fuel pins which use beta quenched clad.
Table 6.1 gives a comparison of key parameters for the ANF-4 8xSC bundle, the 9x9-IX bundle and the 9x9-9X bundle. The lead fuel assemblies are placed in non-limiting core locations within the reload core.
TABLE 6.1
~AR Qfg Number of Rods, Total 64 72 72 Fuel Rod Pitch, Inches .641 .569 .569 Fuel Assembly Loading, kg U02 199.7 201.0 190.6 Fuel Assembly Loading, KgU 176.0 176.8 167 '
Fuel Pellet Haterial UO UO UO Oensity g/cc 1).36 1).55 " 1).36 Percent of T.D. 94.5 96.26 94.5 Oiameter Enriched .4055 .374* .3665 Natural .4045 .374* .3665 Bish Volume (Percent of pellet Volume)
Enriched U02 1.50 1.00 1.00 Enriched U02 - GO203 1.00 1.00 1.00 Natural 0.00 1.00 1.00 Fuel Rod (6" Natural Enriched Haterial on each end)
Fuel Length 150" 1 50 II 150n Cladding Haterial Zr-2"* Zr-2/Zr Liner"* Zr-2""
Clad I.O., Inches .414 .3807 .373 Clad 0.0., Inches .484 .433 .433 Enriched Length 138" 1 38II 1384 Fuel Rod Inventory Low-Low Enrichment 1 (1.5 w/o U-235)
Low Enrichment 5 (2.0 w/o U-235) 4 ( 1.92 w/o U-235) 4 (1.92 w/o U-235)
Hedium Low Enrichment 9 (2.5 w/o U-235)
Hedium Low Enrichment 6 (2.5 w/o U-235 with Gd203 +2.0 w/o Gd203) 5 (2.51 w/o U-235 5 (2.51 w/o 6.235
+1.8 w/o GdyO~) + 1.8/ Gd 0 )
1 (2.51 w/o U-235 1 (2.51 w/o ).335
+4.5 w/o Gd20q) + 4.5 w/o GdyOq)
Hedium Enrichment 21 (2.64 w/o U-235 12 (2.51 w/o U-235) 12 (2.51 2/o U-235)
High Enrichment 20 (3.43 w/o U-235) 50 (2.82 w/o U-235) 50 (2.90 w/o &-235 Inert Water Rod 2 Large Central Can Large Central Can Unvoid Water Area .368 2.447 2. 447 (Inches)
Haximum Width 5.251 5.252 5.252 (Inches)
- Gd203 Rods use 9x9-9X design parameters
""Gd203 Rods use Zr-2 (Beta Heat Treated)
Analyses have been performed consistent with ANF methodology (references 3.0, 10.0 and 12.0) to establish a licensing basis for the two ANF 9x9-IX and the two ANF 9x9-9X LFA's in the WNP-2 Cycle 5 core. The analyses demonstrate the applicability of the WNP-2 Cycle 5 operating limits to these four LFA's unless stated otherwise.
The insertion of four ANF 9x9 LFA's in the Cycle 5 core will have negli-gible effects upon core wide transient performance. However, some 9x9 LFA specific analyses have been performed to assure that the Cycle 5 operating limits are also applicable to the LFA's. Fuel specific LHGR and MAPLHGR limits have been developed for these LFA's.
The dynamic response of the LFA is expected to be almost identical to that of the Bx8 already in .the core. This is due to the fact that the fuel assembly stiffness is provided by the assembly channel, which is the same in both designs.
The 9x9 LFA's are hydraulically compatible with the co-resident ANF 8x8 fuel assemblies based on a comparison of fuel component hydraulic resistances.
Steady state thermal hydraulic analysis has shown that even though the ANF 9x9 LFA design has a somewhat smaller flow area than the ANF 8x8 design, no reduction in thermal margin is experienced in the Cycle 5 core. This is due to the increased critical power performance of the ANF 9x9 design relative to the ANF 8x8 design at WNP-2 Cycle 5 conditions.
The average enrichment and enrichment distribution for the 9x9-IX and 9x9-9X fuel assemblies have been selected to match, as closely as pos-sible, the neutronic performance of the four 8x8 ANF-3 2.64 w/o U-235 reload assemblies included in the Cycle 5 reload. Based on this selec-tion, the neutronic characteristics of the 9x9 LFA's are expected to match to the characteristics of the ANF 8x8 fuel included in the Cycle 5 reload.
Analyses of the WNP-2 Cycle 5 limiting transients have been performed for ANF 8x8, ANF 9x9 LFA and GE P8x8R fuel. It has been shown that using the XN-3 ANF Critical Heat Flux (CHF) correlation, the bundle power required to produce transition boiling in an ANF 9x9 LFA is higher than that for an ANF 8x8 bundle. The Cycle 5 Safety Limit Analysis considered the LFA's such that the MCPR safety limit of 1.06 is also applicable to the LFA's. Therefore, the ANF 9x9 LFA's can be monitored to the ANF 8x8 MCPR fuel limits.
Since heatup is primarily a planar and not an axial phenomena, the appro-priate bundle power limit derived from a LOCA analysis is the peak bundle planar power. The ANF 9x9 LFA's have better cooling during LOCA condi-tions relative to an ANF Bx8 fuel assembly due to the lower stored energy in the fuel rods, a greater surface area provided by the larger number of fuel rods, and more inert surface from the central water channel. Thus, a 15
~ o LOCA analysis for the ANF 9x9 LFA's would yield lower Peak Cladding Temperatures (PCT's) and metal-water reaction rates than an ANF 8x8" assembly at the same bundle peak planar power. The HAPLHGR limits for the ANF 9x9 LFA's restrict the peak bundle planar power to that analyzed for the ANF 8x8 fuel and assure that the USNRC 10CFR50.46 criteria are met for the ANF 9x9 LFA's in Cycle 5.
The fuel loading error was analyzed for the ANF 9x9 LFA's. Results show that if the loading error went undetected, the offsite consequences would remain well within the guidelines specified in 10CFR Part 100.
All operational limits used for ANF Bx8 fuel are applicable to the ANF 9x9 LFA's except for fuel type specific HAPLHGR limits and the 9x9-IX and 9x9-9X LHGR limits. The LHGR limits for the 9x9-IX and 9x9-9X are shown in Figures 6.1 and 6.2 respectively, and the MAPLHGR limits for the LFA's are shown in Figure 6.3.
14 Bundle ~
Average Exposure LHGR (MWD/MT) kw/ft 12 0 13.7 5,000 13.7 10>000 13.7 15,000 13.7 20,000 13.0 30>000 11.5 40>000 10.0 10 50,000 8.5 60>000 7.0 70,000 5.5 0 5,000 10,000 20,000 30,000 40,000 50,000 60,000 70,000 80>000 Average Planar Exposure {MWD/MT)
ANF 9 X 9 - lX Reload Fuel Linear Heat Generation Rate {LHGR) Limit Versus Average Planar Exposure Figure 6.1 890040.3
14 Bundle Average 13 Exposure LHGR (MWD/MT) kw/tt 12 0 13.1 5,000 13.1
~ ~ 10>000 13.1 E 11 15>500 13.1 20>000 12.5 Q 30,000 11.2 40,000 9.9 6$ 10 K 50,000 60,000 8.6 7.3 C
~
0~ 9 70,000 6.1 Q
C 8 Q
Q C 7 Q
= 6$
Q 6 C
0 5,000 10,000 20,000 30,000 40,000 50,000 60,000 70,000 80,000 Average Planar Exposure (MV/D/MT)
ANF 9 X 9- 9X Reload Fuel Linear Heat Ge'neration Rate (LHGR) Limit Versus Average Planar Exposure Figure 6.2 890040.4
12.0 11.5 ~ ~ > ~
11.0 10.5 h ~
>D 10.0 g$
CL 0 9.5
- 8) ~
6$ 9.0 Bundle
>D Average I PD Exposure MAPLHGR E C9 8.5
~ (MwD/MT) kwlft
~
E m-0>
8.0
>D 5>000 11.2 c 10,000 11.2 7.5 15,000 11.2 20>000 11.2 7.0 25,000 9.7 30,000 8.1 6.5 35,000 6.8 6.0 5,000 10,000 15>000 20,000 25,000 30,000 35,000 Bundle Average Exposure (MWD/MT)
ANF 9 X 9 - IX AND 9 X 9 - 9X Reload Fuel Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Bundle Average Exposure Figure 6.3 890040M
7.0 TH MA HYD I D 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 criteria considered in the thermal hydraulic analysis are found in XN-NF-80-19(A), Volume 4, Revision 1 (Reference 3.0).
Analyses performed to demonstrate that these criteria are met include:
o Hydraulic compatabili ty.
o Fuel cladding integrity safety .limit.
o Fuel centerline temperature.
o Bypass flow characteristics.
o Thermal hydraulic stability.
These analyses are discussed in this section for Bx8 fuel. Specific thermal hydraulic design considerations for the 9x9 fuel are discussed in Section 6.0.
71 r li m The hydraulic flow resistances for the ANF reload fuel and the GE 8x8 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 8x8 cores. The close geometrical similarity between the fuel designs and their measured performance characteristics demonstrate that the fuel designs are sufficiently compatible for co-residence in NNP-2. Hydraulic compatability of the 9x9 reload fuel is discussed in Section 6.0.
7.2 F The MCPR fuel cladding integrity safety limit for Cycle 5 is 1.06 which is equal to the Cycle 1, Cycle 2, Cycle 3 and Cycle 4 MCPR safety limit. The methodology used in the MCPR safety limit calcu-lations is found in XN-NF-80-19(A), Volume 4, Revision 1 (Reference 3.0). The NNP-2 Cycle 5 MCPR safety limit analysis methodology and input parameters are described in ANF-89-01, Cycle 5 Plant Tran-sient Report (Reference 2.0).
7.3 rl n The LHGR curve in Figure 3.4 of Reference 8.0 for 8x8 fuel and the LHGR curves in Figures 6.1 and 6.2 for 9x9 fuel are everywhere greater than 120 percent of 'the LHGR limit curve in Reference 8.0.
Therefore, the fuel centerline temperature is protected for 120 percent over power and fuel centerline melt is protected fior all fuel exposures within the bounds of the referenced LHGR curve.
r 7.4 w Core bypass flow was computed using the methodology of XN-NF-524(A)
(Reference 9.0). The bypass flow for the WNP-2 Cycle 5 is 10,7 per-cent of the total core flow which is similar to the Cycle 1 value of 11.8 percent and to the Cycle 4 value'f 11.5 percent. The computed bypass flow will have no adverse impact on reactor operation.
7.5 r The WNP-2 Technical Specifications included survei'llance require-ments for detecting and suppressing power osci llations. 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 (47 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 1.0).
The 9x9 LFA's were included in the core-wide stab) lity analysis.
Local instability tests were performed on 9x9 loads in a BWR-3. No detectable difference was noted in stability performance relative to the co-resident Bx8 fuel (Reference 1.0).
8.0 The neutronic methods for the design and analysis of the WNP-2 Cycle 5 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.
8.1 The Cycle 5 8x8 ANF reload bundles (labeled ANF-4) are similar to the ANF-3 ANF reload bundles (used in the WNP-2 Cycle 4 reload) in nuclear design in all major parameters except for fuel enrichment.
Major nuclear design characteristics for the ANF reload fuel assemblies (ANF-4) are:
21
o The 8x8C fuel assembly contains 62 fuel rods and two water rods. One of the water rods also acts as a spacer capture rod. The. 9x9 LFA fuel assembly contains 72 fuel rods and one central water channel.
o The ANF-4 8x8C fuel assembly average enrichment is 2.62 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 2.79 w/o U-235. The 9x9 LFA fuel assembly average enrichments are 2.53 and 2.59 w/o U-235 with six inches of natural uranium at the top and bottom.
o Five enrichment levels. are utilized in the ANF-4 8x8C fuel assembly to produce a local power distribution which results in a balanced design for Minimum Critical Power Ratio (HCPR) and Maximum Average Planar Linear Heat Generation Rate (HAPLHGR) limits. The 9x9 LFA fuel designs utilize three enrichment levels to accomplish the same goal.
o Each ANF-4 8x8C fuel assembly contains six fuel rods with 2.0 w/o GD203 blended with 2.50 w/o U-235 enriched U02 to reduce initial assembly reactivity. The 9x9 LFA fuel assem-blies contains five 1.8'/. w/o Gd203 pins and one 4.5 w/o Gd203 pin all enriched to 2.51 w/o U-235 for the same purpose.
The enrichment distributioh of the ANF reload designs was selected on the basis of maintaining a balance between the local power peak-ing factors, assembly reactivity, MAPLHGR, and HCPR. For the cen-tral enriched region of the AN-4 SxS assembly, one rod .is enriched to 1.5 w/o U-235, five rods to 2.0 w/o U-235, nine rods to 2.50 w/o U-235, 21 rods to 2.64 w/o U-235, 20 rods to 3.43 w/o U-235, and six rods to 2.50 w/o U-235 plus 2.00 w/o GDp03. The fuel bundle design features of the 9x9 LFA designs can be found in Table 6.1.
In this table, a comparison is also made to the design characteris-tics of the SxS ANF-4 design.
8.2 The core exposure for the end of Cycle 4 (EOC4), the core exposure for the beginning of Cycle 5 (BOC5), and the core exposure for the end of Cycle 5 (EOC5) were'alculated with the XTGBNR Code (Ref-erence 10.0). In addition, BOC core reactivity characteristics for the cold core were calculated along with the standby liquid control system reactivity. The results of these analyses are summarized in Table 8.1.
Table 8.1 CORE NUCLEAR DESIGN Core Exposures at EOC4 (mwd/mtm) 16,700 Core Exposures at BOC5 (mwd/mtm) 12,300 Core Exposures at EOC5 (mwd/mtm) 18,100 BOC Cold Keff, all rods out 1.1133 BOC Cold Keff, strongest rod out 0.9868 Reactivity Defect/R-Value, percent '1. hK/K 0.0 Standby Liquid Control System (SBLC)
Reactivity, 660 PPM Boron, Keff 0.9633 8.3 Some of the major core parameters for NNP-2 Cycle 4 and Cycle 5 are listed in Tabl,e 8.2.
Table 8.2 COMPARISON OF MAJOR CORE PARAMETERS
~Pr g~ 6~1~ GVS.~~
MCPR Limit* (0 mwd/mtm) 1.23 1.23 Doppler Defect 9.5 X 10-6 10.0 X 10-6 (1. dK/K/T)
Cycle Length** (Design; FPD) 227 227 Core Average Exposure 11,200 12,300 (BOC; mwd/mtm)
Core Average Exposure 16,700 18,100 (EOC; mwd/mtm)
- Based on CRHE, 106/. RBM setpoint
- All rods out; full power, 106K flow The differences between the Cycle 4 core and the Cycle 5 core are found in the core loading pattern. The Cycle 4 core consisted of a scatter load of 152 ANF 8x8C unirradiated reload assemblies, 148 once irradiated ANF SxSC reload assemblies, 128 twice irradiated ANF Sx8C reload assemblies and 336 GE PSx8R initial core assemblies.
The Cycle 5 core will consist of a scatter load of 144 ANF 8x8C unirradiated reload assemblies, 152 once irradiated ANF 8xSC reload assemblies, 148 twice irradiated SxSC reload assemblies, 128 thrice irradiated 8xSC reload assemblies and 192 P8xSR assemblies fabri-cated by General Electric (GE) left over from the initial core load.
r 9.0 D E ANF considers eight categories of potential system core wide transient occurrences for jet pump BWRs (Reference 11.0) and has provided analysis results for the three most limiting transients for WNP-2 Cycle 5 to determine the Cycle 5 thermal margins. The three transients determined to be most limiting for Cycle 5 are:
o Load Rejection No Bypass (LRNB).
o Feedwater Controller Failure (FHCF).
o Loss of Feedwater Heating (LOFH).
ANF's methodology for developing thermal limits is found in Reference 12.0. The discussion in Refe'rence 11.0 demonstrates that the other plant transient events are inherently nonlimiting or clearly bounded by the above events.
Two local events, Control Rod Withdrawal Error (CRWE) and Fuel Loading Error (FLE) were analyzed with the methodology described in Reference 10.0. The CRHE was demonstrated to be bounding for certain parts of the fuel cycle ~
The analysis reported here is applicable to both the Sx8C and 9x9 LFA fuel included in the Cycle 5 reload (References 1.0 and 2.0). The results of the core-wide and local transient analyses are provided in the WNP-2 Cycle 5 Reload Analysis Report (Reference 1.0) and in the WNP-2 Cycle 5 Transient Analysis Report (Reference 2.0). The CRHE was evalu-ated and found to be most limiting up to EOC-2000 mwd/mtm at 106 percent of rated core flow, resulting'n a change in CPR of 0.17 for the ANF fuel and 0.17 for the GE fuel at the 106 percent rod block monitor (RBH) trip setpoint. When combined with the 1.06 safety limit, this transient (CRHE) requires a MCPR operating limit of 1.23 for the ANF fuel and 1.23 for the GE fuel in Cycle 5 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 LRNB transient was determined to be the limiting transient and the HCPR limit for ANF fuel is 1.31 and for GE fuel is 1.34 for this portion 24
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 inser-tion times (3.1.3.4, P 3/4.1.8) for this portion of the fuel cycle (EOC-2000 mwd/mtm to EOC). 'or operation at and beyond EOC-2000 with core flow up to 106 percent and these technical specification scram times, the limiting transient is the LRNB transient and the MCPR operat-ing limit within EOC-2000 mwd/mtm to EOC is 1.37 for ANF fuel and 1.41 for GE fuel for Cycle 5 operation. If the Recirculation Pump Trip (RPT) should become inoperable. for any reason and assuming normal scram speeds, and operation up to 106 percent core flow in this exposure range, the limiting transient is then the LRNB transient and the HCPR operating limit is 1.37 for ANF fuel and 1.41 for GE fuel. Finally, if the RPT becomes inoperable within EOC-2000 mwd/mtm to EOC and the plant defaults to the technical specification scram times, the LRNB transient at 106 percent flow is bounding and the HCPR operating limit is 1.41 for ANF fuel and 1.47 for GE fuel.
Additional analyses were performed,to determine the MCPR operating limit with a 107 percent and 108 percent RBH setpoint for'the CRWE event. The resulting changes in CPR are 0.19 for ANF fuel and, 0.19 for- GE fuel at 107 percent, and 0.21 for ANF fuel, and 0.21 for GE fuel at a 108 percent rod block setting. Therefore, operation with a 108 percent RBM setting would require a MCPR limit of 1.27 for ANF and 1.27 for the GE fuel.
The plant transient model used to evaluate the pressurization tran-
'ients, the LRNB and FWCF events, consists of the ANF COTRANSA (Reference 11.0) and XCOBRA-T (Reference 13.0) 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 LRNB 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 LRNB 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 LRNB 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 LRNB was found to be the most limit-ing transient at 106 percent core flow and EOC. All core wide transients were analyzed using bounding values as input.
The Loss of Feedwater Heating (LOFH) events were evaluated with the ANF core simulator model XTGBWR (Reference 10.0) by representing the reactor i.n equilibrium before and after the event. Actual and pro-jected operating statepoints were used as initial conditions. Final conditions were determined by reducing the feedwater temperature by 100'F and increasing core power such that the calculated eigenvalue remained unchanged.
Based on a bounding value analysis, a MCPR operating limit of 1.15 for WNP-2 with a MCPR safety limit of 1.06 is supported (i.e., a change in CPR of 0.09). The WNP-2 HCPR safety limit for Cycle 5 continues to be 1.06; hence the LOFH transient requires a HCPR operating limit of 1.15, for WNP-2 (Reference 2.0).
9.2 Analyses 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 HCPR operating limit is a function of the RBM setpoint. Analyses were performed to support a RBH setpoint of 106 percent, 107 percent, and 108 percent. The change in CPR for the CRWE with a 106 percent RBH setpoint is 0.17 for ANF fuel and 0.17 for GE fuel, for a 107 percent RBM setpoint 0.19 for ANF fuel and 0.19 for GE fuel, and for a 108 percent RBH setpoint 0.21 for ANF fuel, and 0.21 for GE fuel:
9.3 The recirculation flow run-up analysis performed for WNP-2 Cycle 2 was reviewed and was found to change due to a change of model input data influenced by operational data. The reduced flow HCPR operat-ing limit for WNP-2 Cycle 5 is found to be more conservative than the previous reduced flow HCPR operating limit. For final feedwater temperature reduction (FFTR) conditions, the previously reported reduced flow HCPR operating limit remains applicable. The appli-cable limit for Cycle 5 is given in Table 9.1 (Reference 2.0).
Table 9.1 Reduced Flow HCPR Operating Limit For Cycle 5, WNP-2 Core Flow Reduced Flow MCPR r i 100 1.07 90 1.13 80 1.19 70 1.26 60 1.34 50 1.44 40 1.59 0-9.4 ME vr In order to demonstrate compliance with the ASME Code over pres-surization criteria of 110 percent of vessel design pressure (1375 psig), the Main Steam Isolation Valve (MSIV) closure event with failure of the MSIV position switch scram was analyzed with ANF's COTRANSA code (Reference 11.0). The WNP-2 Cycle 5 analysis assumed six safety relief valves out of service. The maximum pressure observed in the analysis is 1315 psig in the vessel lower plenum.
This is 105. percent of the reactor vessel design pressure which is below the 110 percent design criterion of 1375 psig.
The calculated steam dome pressure corresponding to the 1315 psig peak vessel pressure is 1286 psig, for a vessel differential pres-sure of 29 psig. The RPT is assumed to initiate at a pressure 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 pressure differential (1375-1325).
Since the calculated vessel differential pressure is 29 psi, the steam dome safety limit of 1325 psig assures compliance with the ASME criterion of 1375 psig peak vessel pressure.
9.5 The plant system transient events reported earlier in this document, which are potentially limiting for MCPR, were all analyzed at increased 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 fo'r WNP-2 Cycle 2 (Reference 14.0).
ANF has also perfo'rmed analyses which demonstrate that the XN-1 8x8C fuel bundle can operate satisfactorily from a mechanical standpoint at this increased core flow (Reference 15.0). In addition, GE has performed analyses for the reactor internals and for the GE fuel assembly which cons,idered the loads created by operation at this ilarityy flow level and the impacts of these loads on the WNP-2 core inter-nals and the GE fuel assembly. Also, flow induced vibration of the core internals as a result of increased core flow was analyzed.
Finally, analyses were performed for feedwater nozzle and feedwater sparger fatigue at increased core flow (Reference 16.0). The results of all these analyses when considered along with the sim-with the fuel types utilized in Cycle 5, confirm the capability of WNP-2 to operate at 100 percent power and 106 percent core flow during Cycle 5 operation. A review of the 9x9 LFA fuel, discussed in Section 6.0, confirms its capability for increased flow operation.
A containment analysis was performed to determine the impact of operation at increased core flow on the WNP-2 containment LOCA respon'se. The results show that the containment LOCA response for increased core flow operation is bounded by the corresponding FSAR results (Reference 17.0).
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 5.
9.6 Atua ANF performed analyses for WNP-2 which demonstrate the safety of plant operation with a single recirculation loop out of service at 757! of rated power for an extended period of time. These analyses were performed for the most limiting transient events, the pump seizure accident and the loss-of-coolant-accident (LOCA) for the maximum extended power state during 'HNP-2 single loop operation (SLO). The results of the SLO analyses are summarized below:
o The two loop MCPR operating limits (rated conditions) bound the transient requirements for SLO. The single loop transient analyses need not be performed on a cycle by cycle basis and a MCPR 1,35 is appropriate for single loop conditions, o The postulated pump seizure accident, evaluated for SLO condi-tions, is calculated to have a less severe radiological release than the LOCA. The radiological consequences of this postu-late'd accident are bounded by the radiological evaluation performed by GE for the LOCA and are well within the 10CFR100 limits, o The single loop ECCS analysis supports the use of the HIP-2 two loop MAPLHGR limits for ANF fuel when the reactor is operating in the SLO mode consistent with the single loop MCPR Operation limit (1.35 at 50 percent of rated flow). Single loop opera-tion of WNP-2 with the two loop ANF fuel MAPLHGR limits assures that the emergency core cooling systems for the WNP-2 plant will meet the U.S. NRC acceptance criteria of 10CFR50.46 for loss-of-coolant accident breaks up to and including the double-ended severence of a reactor coolant pipe, The transient and pump seizure accident analyses are described as Reference 18.0 and the LOCA analyses are described in Reference 19.0.
With a single recirculation loop in operation, the GE analyses sup-ported continued operation with an increase of 0.01 in the MCPR safety 1'imit. ANF performed a single loop MCPR safety limit calcu-lation and found that less than one tenth of one percent of the rods to be in boiling transition which supports a MCPR safety limit of 1.07. Because of the similarity between the ANF and GE fuel types making up the core, and because of the similarity in the magnitude of the uncertainties which determine the MCPR safety limit, this small increase in the safety limit value can be used for operation with ANF .fuel .and single loop analyses. For Cycle 5 operation with both recirculation loops in operation, the MCPR safety limit is 1.06, which is the same value as was used for the previous cycles.
For Cycle 5 operation with a single recirculation loop in service, the MCPR safety limit is 1.07, which is also the same value used for the previous cycles. The LFA single loop operation limits are bounded by the two-loop operation limits (Reference 1.0).
9.7 i w Reference 20.0 presents a final feedwater temperature reduction (FFTR) analysis with thermal coastdown for HNP-2. This analysis is the subject of a currently proposed Tech. Spec. change and is appli-cable to future WNP-2 .fuel cycles. The FFTR analysis was performed for a 65'F temperature reduction. This FFTR analysis is applicable after the all rods out condition is reached with normal feedwater temperature. The FFTR analysis results show that CPR changes for the LRNB and FHCF transients of + 0.02 and 0.01, respectively, are applicable to these respective anticipated operational occurrence (AOO) events. That is, these LRNB and FHCF limit changes are appli-cable when Cycl'e 5 reactor operation is being extended with thermal coastdown at FFTR conditions and are applicable to both the SxS and 9x9 reload fuel designs (Reference 2 ').
10.0 P T For Cycle 2, ANF had analyzed the LOCA to determine MAPLHGR limits for ANF SxS fuel; The results of this analysis are presented in Reference 21.0. For the 9x9 LFA fuels, these same MAPLHGR limits as modified to account for the difference in activated fuel pin length between the two fuel types, are also applicable to the 9x9 LFA fuel (Reference 1.0).
These Cycle 2 results are equally applicable to Cycle 5. ANF's method-ology for the LOCA analysis is given in References 22.0, 23.0, and 24.0.
In addition, the Rod Drop Accident (RDA) was analyzed to demonstrate compliance with the 280 cal/gm design limit. ANF's methodology for the RDA analysis can be found in Reference 10.0.
10.1 Reference 25.0 describes ANF's WNP-2 LOCA break spectrum analysis which defined the limiting break for HNP-2. The analysis of this event for HNP-2 is described in Reference 26.0. The LOCA analysis described in Reference 26.0 was performed for an entire core of ANF SxSC fuel and therefore provides MAPLHGR limits for ANF fuel only.
These results are applicable to operation in WNP-2 Cycle 5 including the 9x9 LFA fuel (Reference 1.0).
ANF reload 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 5 operation.
10.2 Dl ANF's methodology for analyzing the RDA is given in Reference 10.0.
For WNP-2 Cycle 5, the analysis shows a value of 121 cal/gm for the maximum deposited fuel rod enthalpy during the worst case postulated RDA (Reference 1.0). This is well below the design limi-t value of 280 cal/gm.
10.3 1 i n To support operation of WNP-2 with a core composed of GE Cycle 1 fuel and ANF reload 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 opera-tion. The single loop ECCS analysis supports the use of the WNP-2 two loop MAPLHGR limits for ANF fuel when the reactor is operating in the SLO mode, Single loop operation of WNP-2 with the two loop ANF fuel MAPHGR limits assures that the emergency core cooling systems for the WNP-2 plant will meet the U.S. NRC acceptance criteria of 10CFR50.46 for loss-of-coolant accident breaks up to and including the double-ended severance of a reactor coolant pipe.
11.0 The Supply System has developed a restart physics test program to be carried out prior to or during Cycle 5 startup. This program includes a core loading verification test, a control rod functional test, an in sequence shutdown margin test, and a TIP asymmetry test. The proposed test goals and a brief description of each test is given below.
Qga1 To assure that the WNP-2 Cycle 5 Core is loaded according to the design analyzed by ANF.
t t i1 b p d4tht 1d television camera. A series of initial passes will be made with the fuel mast set at a predetermined height to assure that all fuel assemblies are fully 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'.
ggaJ. To determine and verify control. rod mobility and functionality.
Following the completion of fuel loading, for each cell of four fuel assemblies, 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 wi 11 verify that there are no gross reac-tivity 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.
11.3 ggyl To assure that the Technical Specification shutdown margin
~tJ requirement is satisfied.
startup h d criticality. Critical tk d q 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 makgin with the analytically determined strongest . control rod withdrawn is confirmed.
11.4 QgJ. To assure proper TIP systems operation and to verify that the TIP system uncertainty is within the limits assumed for tran-sient analysi.s.
IRR~MGLM~ This test is performed in the power range pre-ferably above 75 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.
12.0 REEHKKCKS 1.0 ANF-89-02, "WNP-2 Cycle 5 Reload Analysis Report", Advanced Nuclear Fuels Corporation, January 1989.
2.0 ANF-89-01, "HNP-2 Cycle 5 Plant Transient Analysis Report",
Advanced Nuclear Fuels Corporation, January 1989.
3.0 XN-NF-80-19(A), Volume 4, Revision 1, "Exxon Nuclear Methodology For Boiling Hater Reactor: Application of the ENC Methodology to BHR Reloads", Exxon Nuclear Company, September 1983.
4.0 WPPSS-C-ANF-101, "WNP-2 Cycle 2 Reload Summary Report", February 1986.
5.0 ANF-88-178, "Washington Public Power Supply System, WNP-2 Reload ANF-4, Cycle 5 Neutronics Design Report", Advanced Nuclear Fuels Corporation, November 1988.
6.0 XN-NF-81-21(A), Revision 1, "Generic Mechanical Design For Exxon Nuclear Jet Pump BHR Reload Fuel", Exxon Nuclear Company, January 1982.
7.0 XN-NF-81-21(A), Revision 1, Supplement 1, "Generic Mechanical Design For Exxon Nuclear Jet Pump BWR Reload Fuel", Exxon Nuclear Company, March 1985.
8.0 XN-NF-85-67(A), Revision 1, "Generic Mechanical Design For Exxon Nuclear Jet Pump BHR Reload Fuel", Exxon Nuclear Company, September 1986.
9.0 XN-NF-524(A), Revision 1, "Exxon Nuclear Critical Power Method-ology For Boiling Water Reactors", Exxon Nuclear Company, November 1983.
10.0 XN-NF-80-19(A), Volume 1 and Volume 1 Supplements 1 and 2, "Exxon Nuclear Methodology For Boiling Hater Reactors: Neutronic Methods For Design and Analysis", Exxon Nuclear Company, March, 1983.
11.0 XN-NF-79-71(P), Revision 2 (as supplemented), "Exxon Nuclear Power Plant Transient Methodology", Exxon Nuclear Company, November 1981.
12.0 XN-NF-80-19(A), Volume 3, Revision 2, "Exxon Nuclear Methodology For Boiling Hater Reactors: TKERMEX Thermal Limits Methodology Summary Descriptions", Exxon Nuclear Company, January 1987.
13.0 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 Kydraulic Core Analysis", Advanced Nuclear Fuels Corporation, February 1987.
14.0 J. B. Edgar, Letter to WPPSS, Supplemental Analysis Results, ENWP-86-0067, Exxon Nuclear Company, April 15, 1986.
32-
15.0 J. B. Edgar, Letter to WPPSS, ENWP-86-0033, Exxon Nuclear Company, February 13, 1986.
16.0 NEDC-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, February 1986.
17.0 "Final Safety Analysi s Report, WPPSS Nuclear Project No. 2", as reviewed through Amendment 35, November 1984.
18.0 ANF-87-119, "WNP-2 Single Loop Operation Analysis", Advanced Nuclear Fuels Corporation, September 1987.
19.0 ANF-87-118, "WNP-2 LOCA Analysis for Single Loop Operation",
Advanced Nuclear Fuels Corporation, September 1987.
20.0 XN-NF-87-92, "WNP-2 Plant Transient With Final Feedwater Tempera-ture Reduction", Advanced Nuclear Fuels Corporation, June 1987.
21.0 XN-NF-86-01, Rev. 1, "WNP-2 Cycle 2 Reload Analysis Report", Exxon Nuclear Company, February 1986.
22.0 XN-NF-80-19(A), Volumes 2, 2A, 2B, and 2C, "Exxon Nuclear Meth-odology For Boiling Water Reactors: EXEM ECCS Evaluation Model",
Exxon Nuclear Company, September 1982.
23.0 XN-NF-CC-33(A), Revision 1, "HUXY: A Generalized Multirod Heatup Code With 10CFR50, Appendix K, Hhatup Option", Exxon Nuclear Company, 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 BWR 5", Exxon Nuclear Company, December 1985.
26.0 XN-NF-85-139, "WNP-2 LOCA-ECCS Analysis MAPLHGR Results", Exxon Nuclear Company, December 1985,
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