Text

Washington Nuclear Plant-2 Cycle 4 Plant Transient Analysis.
	ML17279A881
Person / Time
Site:	Columbia
Issue date:	01/31/1988
From:	Hibbard M, Krajicek J, Rawlings J; SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:	;
Shared Package
ML17279A877	List: ML17279A876; ML17279A878; ML17279A880; ML17279A881; ML17279A882;
References
	ANF-88-01, ANF-88-1, NUDOCS 8803150313
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pgp gDOCK 05000397 ADVANCEDNUCL.EARFUELS CORPORATION ANF-88-02 Issue Date: 1/15/38 WNP-2 CYCLE 4 RELOAD ANALYSIS Prepared By:

. E. Krajicek/H. J. Hibbard BWR Safety Analysis

'Licensing and Safety Engineering fuel Engineering and Technical Services Prepared By:

J. C. Rawlings ENSA AN AFFIUATE OF KRAFlWERKUNION Q~KWU

NUCLEAR REGULATORY COMMISSION REPORT DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Advanced Nuclear Fuels Corporation. It is being submitted by Ad-vanced Nuclear Fuels Corporation to the U.S. Nuclear Regulatory Commission as part of a technical contribution to facilitate safety analyses by licensees of the U.S. Nuclear Regulatory Commission which utilize Advanced Nuclear Fuels Corporation-fabricated reload fuel or other technical services provided by Ad-vanced Nuclear Fuels Corporation for light water power reactors and it is true and correct to the best of Advanced Nuclear Fuels Corporation's knowledge, informa-tion, and belief. The information contained herein may be used by the U.S.

Nuclear Regulatory Commission In its review of this report, and under the terms of the respective agreements, by licensees or applicants before the U.S. Nuclear Regulatory Commission which are customers of Advanced Nuclear Fuels Cor-poration In their demonstration of compliance with the U.S. Nuclear Regulatory Commission's regulations.

Advanced Nuclear Fuels Corporation's warranties and representations concem-ing the subject matter of this document are those set forth in the agreement bet-ween Advanced Nuclear Fuels Corporation and the customer to which this docu-ment is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Advanced Nuclear Fuels Corporation nor any person acting on its behalf:

A. Makes any wananty, or representation, express or Im-plied, with respect to the accuracy, completeness, or use-fulness of the Information contained in this document, or that the use of any Information, apparatus, method, or pro-cess disclosed in this document will not infringe privately owned rights, or B Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, ap.

paratus, method. or process disclosed in this document.

XN.NF-FOO 766 (fl

ANF-88-02 TABL 0 CONTEN S

~Sectie Pacae

1.0 INTRODUCTION

2.0 FUEL MECHANICAL DESIGN ANALYSIS...............................

3.0 THERMAL HYDRAULIC DESIGN ANALYSIS............................. 3 3.1 D esign Criteria.......................................

3.1.3 Fuel Centerline Temperature............................. .....

3.2 Hydr aulic Characterization...........................

3.2.5 Bypass Flow...................................................

3.3 MCPR Fuel Cladding Integrity Safety Limit..................... ~ ~ 3 3.3.1 Coolant Thermodynamic Condition............................... ~ ~ 3 3.3.2 Design Basis Radial Power Distribution........................ 3 3.3.3 Design Basis Local Power Distribution........... .. .... .

4.0 NUCLEAR DESIGN ANALYSIS.......................................

4.1 Fuel Bundle Nuclear Design Analysis................. .........

4.2 Core Nuclear Design Analysis..................................

4.2.1 C ore Configuration............................................

J ~

4.2;2 Core Reactivity Characteristics...............................

4.2.4 Core Hydrodynamic Stability...................................

5.0 ANTICIPATED OPERATIONAL OCCURRENCES...........................

5.1 Analysis Of Plant Transients At Increased Core Flow Conditions 5.2 Analyses For Reduced Flow Operation...........................

5.4 ASME Overpressurization Analysis.

5.5 Control Rod Withdrawal,Error.....

5.6 Fuel Loading Error............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

5.7 Determination Of Thermal Margins.

6.0 POSTULATED ACCIDENTS............. 12 6.1 Loss-Of-Coolant Accident.............................. " ...... 12 6.1.1 Break Location Spectrum.......... 12 6.1.2 Break Size Spectrum.............. 12 6.1.3 MAPLHGR Analyses................. ~ ~ 12 6.2 Control Rod Drop Accident........ 12

-ii- ANF TAB E OF CONTENTS (Continued)

Section Pacae 7.0 TECHNICAL SPECIFICATIONS................... 13 7.1 Limiting Safety System Settings........... 13 7.1.1 HCPR-Fuel Cladding Integrity Safety Limit. 13 7.1.2 Steam Dome Pressure Safety Limit.......... 13 7.2 Limiting Conditions For Operation......... 13 7.2.1 Average Planar Linear Heat Generation Rate Limits For ANF 8x8 Fuel.......................... 13 7.2.2 Hinimum Critical Power Ratio............................ . .. .. 13 7.2.3 Surveillance Requirements................. 14 7.2.3.1 7.2.3.2 Scram Insertion Time Surveillance Stability Surveillance.................... '

7.2.3.3 Technical Specification LHGR Surveillance. 15 9.0 ADDITIONAL REFERENCES.................................. 27 APPENDIX A ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ A-1

- iii- ANF-88-02 LIST OF TABLES Table Pa<ac

4. 1 Neutronic Design Values........................................... 16 S OF FIGUR S

~Fi ure ~Pa e 3.1 Radial Power Histogram For I/4 Core Safety Limit Model........... 18 WNP-2 Cycle 4 Safety Limit Local Peaking Factors (ANF XN-3 Fuel). 19 WNP-2 Cycle 4 Safety Limit Local'eaking Factors (ANF XN-'1,"-2

'.1 uel)............................................................

WNP-2 Cycle 4 Enriched Zone Enrichment Distribution..............

20 21 4.2 WNP-2 Cycle 4 Reference Loading Pattern By Fuel Type (One quarter Of Symmetrical Core Loading)........................ 22 5.1 WNP-2 Cycle 4 Control Rod Withdrawal Analysis Initial C ontrol Rod Pattern.............................................. 23 5.2 Reduced Flow MCPR Operating Limit For Normal Feedwater T emperature....................................................... 24 5.3 Reduced Flow MCPR Operating Limit For FFTR Operation............. 25 7.1 Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure, ANF 8x8 Fuel.................................... 26

0 ANF-88-02

1.0 INTRODUCTION

This report summarizes the results of the analyses performed by Advanced Nuclear Fuels Corporation (ANF) in support of the Cycle 4 reload for the Supply System Nuclear Project Number 2 (WNP-2). WNP-2 is scheduled to commence Cycle 4 operation in June 1988. This report is intended to be used tt Itp E N I C p y CENCE t pt I 8 t XN-NF-8 -I fd.,

Volume 4, Rev. 1, "Application of the ENC Methodology to BWR Reloads," which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list.

Section numbers in this report are the same as corresponding section numbers

~XN-NF-8-1 A, 111 8, 8 . l. App dt A I ttt 8 t dd single loop operation.

Final feedwater temperature reduction (FFTR) analysis with thermal coastdown

~ ~ ~ ~

was performed for WNP-2. This FFTR analysis is applicable after the all rods

~ ~ ~ ~

~

out condition is reached with normal feedwater temperature. That is,

~ ~ ~ ~

~

additional MCPR limit changes are applicable when Cycle 4 reactor operation is being extended with thermal coastdown and FFTR.

The WNP-2 Cycle 4 core will comprise a total of 764 fuel assemblies, including 152 ANF 8x8 unirradiated assemblies, 148 once irradiated ANF 8x8 assemblies, 128 twice irradiated ANF 8x8 assemblies, and 336 thrice irradiated P8x8R assemblies fabricated by General Electric (GE). The reference core configuration is described in Section 4.2.

The design and safety analyses reported in this document .were based on the design and operational assumptions in effect for WNP-2 during the previous operating cycle which encompass core flow up to 106% of the design basis value.

ANF-88-02 2.0 FUEL MECHANICAL DESIGN ANALYSIS Applicable Fuel Design Report: Reference 9.8 The expected power history for the fuel to be irradiated during Cycle 4 of WNP-2 is bounded by the assumed power history in the fuel mechanical design analyses.

ANF-88-02 3.0 THERMAL HYDRAULIC D SIGN ANA YSIS 3.1 Desi n Criteria 3.1.3 Fuel Centerline Tem erature The LHGR curve in Figure 3.4 of Reference 9.8 shows that the ANF 8x8 fuel centerline temperature is protected for 120% over power. The LHGR curve in Reference 9.8 is greater. than 120% above the LHGR limit curve in Reference 9.1. Therefore, fuel centerline melt is protected for all ANF 8x8 exposures within the bounds of the referenced LHGR curves.

3.2 H draulic Characterization 3.I..S

~ ~ ~F1 Calculated Bypass Flow Fraction 3.3 MCPR Fuel Claddin Inte rit Safet Limit 3.3.1 Coolant Thermod namic Condition Core Power 3817 MWt Core Inlet Enthalpy 526.4 Btu/ibm Steam Dome Pressure 1030 psia Feedwater Temperature 420'F 3.3.2 Desi n Basis Radial Power Distribution See Figure 3.1

ANF 3.3.3 Desi Basis Local Power Dist ibution See Figures 3.2 and 3.3.

ANF-88-02 4.0 NUCLEAR DESIGN ANALYSIS 4.1 Fuel Bundle Nuclear Desi n Anal sis Assembly Average Enrichment 2.64 w/o U-235 Radial Enrichment Distribution Figure 4.1 Axial Enrichment Distribution Uniform 2.81 w/o U-235 with 6-inch top and bottom natural uranium blankets Burnable Poisons Figure 4.1 Non-Fueled Rods Figure 4.1 Neutronic Design Parameters Table 4.1 Note: The reload includes 24 ANF 8x8 assemblies of the 2.72 w/o U-235 design loaded in Cycle 3 and described in the Cycle 3 Reload Analysis Report XN-NF-87-25.

4.2 Core Nuclear Desi n Anal sis 4.2.1 Core Confi uration Figure 4.2 Core Exposure at EOC3 (HWd/HTU) 15,300 Core Exposure at BOC4 (HWd/HTU) 11,200 Core Exposure at EOC4 (HWd/HTU) 16,900 4.2.2 Core Reactivit Characteristics BOC Cold k-eff, All Rods Out 1.1194 BOC Cold k-eff, Strongest Rod Out 0.9894 Reactivity Defect (R-Value) 0.0 Standby Liquid Control System (SBLC) 0.9654 660 ppm Boron, Cold k-eff

ANF 4.2.4 Core H drod namic Stabilit

.Power %Flow State Points Deca Ratio COTRAN 65/45* 0.55 46/27.6** 0.88 42/23 8***

'.82

45 percent flow - APRH Rod Block intercept point.

- Two pump minimum flow - 46 percent power.

- - Natural circulation flow - APRM Rod Block intercept point.

ANF-88-02 5.0 NTICI AT D OPERATIONA OCCURRENCES Applicable Transient Analysis Report Reference 9.3 5.1 nal s's Of Pla t Transients At Increased Core Flow Conditions Reference 9.3 and 9.11 Limiting Transient(s): Load Rejection Without Bypass (LRWB)

Feedwater Controller Failure (FWCF)

Loss of Feedwater Heating (LOFH)

Transient analyses for WNP-2 Cycle 2 anticipated operational events showed that delta CPR values at design basis conditions are bounded by delta CPR values at design basis power (104%) and increased core flow conditions (106%). Thus Cycle 4 analyses results at increased core flow conditions are conservatively applicable to rated flow conditions.

Cycle 4 specific analyses of transient events were performed with the recirculation pump (RPT) in service and out of service, with normal scram speed (NSS) and technical specification scram speed (TSSS), and at exposures of end-of-cycle and at end-of-cycle -2000 NWd/HTU (3754 HWd/HTU) as shown in following table. On a generic, basis, analyses were performed for thermal coastdown with FFTR to extend cycle operation.

The loss of feedwater heating event was analyzed on a plant specific bounding value basis and the delta CPR results are bounding values for WNP-2.

ANF Haximum Delta CPR Transient*

X Power/

/ Flow Haximum II 1 Fl 1' Haximum ll ~i Pressure F

GE ANF LRNB, NSS 104/106 119 373 1170 0.25 0.24 RPT Operable LRNB,'NSS 104/106 125 505 1181 0.32 0.29 RPT Inoperable LRNB, TSSS 104/106 125 442 1175 0.32 0.30 RPT Operable LRNB, TSSS 104/106 131 574 1189 0.38 0.35 RPT Inoperable LRNB, TSSS 104/106 110 284 1168 0.05 0.05 RPT Inoperable end-of-cycle minus 20QO HWd/HTU FWCF, NSS 47/106 50 187 1010 0.12 0.

RPT Operable FWCF, NSS 47/106 52 129 1020 0.15 0.14 RPT Inoperable FWCF, TSSS 47/106 51 110 1013 0.14 0.12 RPT Operable LOFH N/A N/A N/A N/A 0.09 0.09 5.2 Anal ses For Reduced Flow 0 eration Reference 9.3 and 9.11 Limiting Transient: Recirculation Flow Increase 5.4 ASHE Over ressurization Anal sis Reference 9.3 and 9.11 Limiting Event HSIV Closure

Normal scram speed (NSS) is based on measured plant scram inserti data, see Section 7.2.3.1.

ANF-88-02 Worst Single Failure HSIV Position Scram Trip Haximum Pressure 1315 psig Maximum Steam Dome Pressure 1286 psig 5.5 Control Rod Withdrawal Error Initial Control Rod Pattern for CRWE Analysis Figure 5.1 Rod Block ANF Fuel GE Fuel onitor Settin Distance Withdrawn Delta-CPR Delta-CPR (ft) 106%" 5.0 0.17 0.21 107% 5.5 0.18 0.22 108% 6.0 0.20 0.23

5. 6 Fuel Loadin Error With Correctly Loadin Error Loaded Core Maximum LHGR, kW/ft 16.2 13. 4 Minimum HCPR 1.25 1.41 5.7 Determination Of Thermal Mar ins Summary of Thermal Margin Requirements All system transient results at the more limiting incr eased flow conditions (106%). LRWB results for the more limiting power (design basis condition - 104%) for this transient.

"Rod Block Monitor Setting (RBH) of 106% for Cycle 4.

10 ANF Delta CPR MCPR Limit Equipment GE ANF GE ANF vent 0 erat'onal Status Fuel eel Fuel Fuel Model LRNB RPT Operable, NSS 0.25 0.24 1.31 1.30 COTRANSA/XCOBRA-T LRNB RPT Inoperable, 0.32 0.29 1.38 1.35 NSS LRNB RPT Operable, TSSS 0.32 0.30 1.38 1.36 LRNB RPT Inoperable, 0.38 0.35 1.44 1.41 TSSS LRNB RPT Inoperable, 0.05 0.05 1.11 1.11 TSSS, EOC -2000 MWd/HTU FWCF RPT Operable, NSS 0.12 0.11 1.18 1.17 0

FWCF ~

RPT Inoperable, 0.15 0.14 1.21 1.20 NSS FWCF RPT Operable, TSSS 0.14 0.12 1.20 1. 18 LOFH N/A 0.09 0.09 1.15 1.15 XTGBWR Note: For cycle extension with reduced feedwater temperature, add 0.02 to delta CPR/HCPR LRNB and subtract 0.01 delta CPR/HCPR from FWCF transient results in the above table.

HCPR Operating Limits At Rated Condition For Cycle Exposures Less Than EOC -2000 HWd/HTU (100'o 106% Flow)

~Fue1 T e MCPR Limit 106% RBS ANF 1.23 GE 1.27

ANF-88-02 HCPR Operating Limits At Rated Condition From EOC -2000 HWd/MTU To EOC (100 To 106% Flow) With Normal Feedwater Temperature

~Fuel T e CPR imit ANF 1.30 GE 1.31 HCPR Operating Limits At Rated Condition Beyond All Rods Out With Reduced Feedwater Temperature (100 To 106% Flow And Thermal Coastdown) Point (EOC4)

~Fuel T e MCPR Limit ANF 1.32 GE 1.33 HCPR Limits at Off-Rated Conditions Figure 5.2 and 5.3 Reduced Flow MCPR Limit Reference 9.3 and 9.11

12 ANF-88-02 6.0 OSTU ATED ACCIDENTS 6.1 Loss-Of-Coolant Accident 6.1. 1 B eak Location S ectrum Reference 9.4 6.1.2 Break Size ectru Reference 9.4 I'eference S

6.1.3 MAMMA A RII (ANM 9.5 Limiting Break: Split Break in the Recirculation Suction Piping With an Area Equal to Sixty Percent of the Double-Ended Cross-Sectional Pipe Area Bundle Average Exposure MAPLHGR Peak Clad Peak Local

~NMR MI ~kW ft Tem erature 'F MWR 0 13.0 1765 0.49 5,000 13.0 1766 0.48 10,000 13.0 1765 0.47 15,000 13.0 1772 0.47 20,000 13.0 1788 0.54 25,000 11.3 1699 0.34 30,000 9.4 1521 0.17 35,000 7.9 1397 0.10 6.2 Control Rod Dro Accident Reference 9.7 Dropped Control Rod Worth, mK 8.9 Doppler Coefficient dk/kdT, 1/'F 9.5 x 10 6 Effective Delayed Neutron Fraction 0.0050 Four-Bundle Local Peaking Factor 1.26 Haximum Deposited Fuel Rod Enthalpy (cal/gm) 149

13 ANF-88-02 7.0 TECHNICAL SPECIFICATIONS 7.1 Limitin Safet S stem Settin s 7.1.1 MCPR Fuel Claddin Inte rit Safet Limit MCPR Safety Limit 1.06 7.1.2 Steam Dome Pressure Safet Limit Pressure Safety Limit 1346 psig 7.2 Limitin Conditions For 0 eration 7.2.

~ ~ 1 Aver a e Planar Linear Heat Generation Rate Limits For ANF 8x8 Fuel Bundle Average Exposure MAPLHGR MWd MTU ~kW ft 0 13.0 5,000 13.0 10,000 13.0 15,000 13.0 20,000 13.0 25,000 11.3 30,000 . 9.4 35,000 7.9 These MAPLHGR limits are not impacted by the small enrichment change associated with ANF fuel loaded for Cycle 4. For single loop operation these limits also apply to ANF Fuel consistent with the flow dependent MCPR curve (1.35 at 50 percent of rated flow) 7.2.2 Minimum Cri'tical Power Ratio Rated Condition MCPR Operating Limit Up To EOC -2000 MWd/MTU Exposure (100 To 106% Flow)

ANF ~uel T e Limit 106/. RBS ANF 1.23 GE 1.27 Rated Conditions MCPR Operating Limits From EOC -2000 MWd/MTU To EOC (10N To 1061 Flow)

~Fuel T e Limit ANF 1.30 GE 1.31 Thermal Coastdown and FFTR Rated Condition MCPR Operating Limit Beyond All Rods Out Point With Reduced Feedwater Temperature (100%

to 106% Flow)

~Fuel ANF GE T e Limit 1.32 1.33 0

Reduced Flow MCPR Limit (all cycle exposures) Figures 5.2 and 5.3 7.2.3 Surveillance Re uirements 7.2.3. 1 Scram Insertion Time Surveillance The ANF reload safety analyses were performed using the control rod insertion times shown below which are based on plant data. In the event that plant surveillance shows these scram insertion times may be exceeded, the plant thermal margin limits are to default to the values which correspond to the technical specification (TSSS) control rod scram times (see Section 5.7).

15 ANF-88-02 Position Inserted From Average Rod Time In Seconds Full Withdrawn As Defined In Footnote*

Notch 45 0.404 Notch 39 0.660 Notch 25 1.504 Notch 5 2.624 7.2.3.2 Stabilit Surveillance Core hydrodynamic stability analyses require slight modification to the Technical Specifications which preclude operation in specified power/flow regions. The results of these analyses support operation below a line defined by the following power/flow points: 42% Power/23.8/. Flow, 46% Power/27.6%

Flow and 65% Power/45% Flow (see Section 4.2.4).

Surveillance requirements remain unchanged for Cycle 4, e.g., surveillance is

~

required when operating in a power flow region above the 80% rod line and less

~

than 45% core flow.

7.2.3.3 Technical S ecification LHGR Surveillance The Technical Specification linear heat generation rate (LHGR) limit versus average planar exposure for ANF 8x8 reload fuel is shown in Figure 7. 1. This figure was developed from information contained in Reference 9. 1, and the region of permissible operation is shown.

Slowest measured average control rod insertion time to specified notches for each group of four control rods arranged in a 2x2 array.

16 ANF TABLE 4.1 NEUTRONIC DESIGN VALUES r

Fuel Pellet Fuel Material U02 Sintered Pellets Density, g/cc 10.36

/o of T.D. 94.5 Diameter, inch Enriched Fuel 0.4055 Natural Fuel 0.4045 Fuel Rod Fuel Length, inch 150 Cladding Material Zircaloy-2 Clad, I.D., inch 0.414 Clad, O.D., inch 0.484 Fuel Assembl Number of Fuel Rods 62 Number of Inert Water Rods Fuel Rod Enrichments Figure 4.1 Fuel Rod Pitch, inch 0.641 Fuel Assembly Loading, kgU 176.0

17 ANF-88-02 TABLE 4.1 NEUTRONIC DESIGN VALUES (Continued)

Core Data Number of Fuel Assemblies 764 Rated Thermal Power, HW 3323 Rated Core Flow, Mlbm/hr 108.5 Core Inlet Subcooling, Btu/ibm 19.0 Reactor Pressure, psia 1008.0 Channel Thickness, inch 0.100 Fuel Assembly Pitch, inch 6.00 ater Gap Thickness (symmetric), inch 0.522 Control Rod Data Absorber Haterial B4C Total Blade Span, inch 9.75 Total Blade Support Span, inch 1.58 Blade Thickness, inch 0.260 Blade Face-To-Face Internal Dimension, inch 0.200 Absorber Rods Per Blade 76 Absorber Rod Outside Diameter, inch 0.188 Absorber Rod Inside Diameter, inch 0.138 Absorber Density, % of Theoretical 70.0

WNP-2 CVCLE 4 DESIGN BASIS RADIAL POHER 12.5 (A

Ll 10 C3 63 c 7.5 C3 2.5 0

0 0.25 0.50 0.75 1 1.25 1.50 1.FS 2 hO BUNDLE PONER FRCTOR Figure 3.1 Radial Powe togram For I/O Core Safety Limit Model

19 ANF-88-02'L L ML M M

=

ML L LL 0.93 0.95 1.02 1.06 1.06 1.02 0.95 0.92 L ML H ML H H M L 0.95 0.97 1.08 0.87 1.04 1.07 1.04 0.95 ML H H H H ML ML 1.02 1.08 1.00 0.98 1.00 0.90 1.02 M ML H M H H M

'1.06 0.87 1.00 0.00 0.90 0.97 1.03 1.06

'W M H H M H M M 1.06 1.04 0.98 0.90 0.00 0.99 0.93 1.05 ML H H H H H H 1.02 l.07 1.00 0.97 0.99 -1.00 1.06 L ML H M H ML ML 0.95 0.90 1.03 0.93 1.06 0.96 1.07 LL ML M M M ML L L'.95 0.92 1.02 1.06 1.05 1.08 1.07 1.03 Figure 3.2 WNP-2 Cycle 4 Safety Limit Local Peaking Factors (ANF XN-3 Fuel)

20 ANF LL L ML M M ML L LL 0.95 0.96 1.00 1.03 1.03 1.00 0.96 0.95 L ML H ML H H L 0.96 0.98 1.05 0.92 1.03 1.05 0.96 ML H H H H H ML ML 1.00 1.05 1.02 1.01 1.00 1.01 0.94 1.00 M ML H W M H H M 1.03 0.92 1.01 0.00 0.93 1.00 1.03 1.03 M H H M W H M 1.03 1.03 1.00 0.93 0.00 1.00 0.97 1.03 ML H H H H H M 1.00 1.05 1.00 1.00 1.02 1.05 1.04 L M ML. H. M H ML ML 0.96 1.02 0.94 1.03 0.97 1.05 0.97 1.03 LL L ML M M M ML 0.95 0.96 1.00 1.03 1.03 1.04 1.03 Figure 3.3 WNP-2 Cycle 4 Safety Limit Local Peaking Factors (ANF XN-1, -2 Fuel)

21 ANF-88-02 x*% x*%%%*% s kx*% x%%% %*%% xx LL ML ML H

ML H

H H

H

':ML H

M MLx .

LL L

ML ML>> H H M H H M: W

' M: M H H H H H M: ML~'< H M . H . MLx . ML LL ML ML LL RODS ( 3) 1.50 W/0 U235 L RODS ( 7) 1.94 W/0 U235 ML RODS ( 9) 2.50 W/0 U235 M RODS (16) 2.86 W/0 U235 H RODS (22) 3.43 W/0 U235 ML< RODS ( 5) 2.50 W/0 U235 + 2.00 W/0 GD203 W RODS ( 2) INERT WATER ROD Figure 4. 1 WNP-2 Cycle 4 Enriched Zone Enrichment Distribution

22 ANF '

1 2 3 4 6 7 8 9 10 11 12 13 14 15 8 "F F

D 8, 10 13 14 15 A ~

Fuel Number of 56 GE SxS Type II 1.76 w/o U-235 (Cycle 1) 280 GE SxS Type III 2. 19 w/o U-235 (Cycle 1) 128 ANF SxS 2.72 w/o U-235 (Cycle 2)

'148 ANF 8xS 2.72 w/o U-235 (Cycle 3) 24 ANF SxS 2.72 w/o U-235 (Cycle 4) 128 ANF 8x8 2.64 w/o U-235 (Cycle 4)

Figure 4.2 WNP-2 Cycle 4 Reference Loading Pattern by Fuel Type (One quarter of Symmetrical Core Loading)

23 ANF-88-02 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 55 00 -- 36 -- 00 51 51 24 -- 18 -- 00 -- 18 -- 24 47 43 43 39 -- 00 18 -- 00 24 -- 00 18 -- 00 -- 39 35 35 31 -- 36 00 -- 24 12 -- 24 00 -- 36 -- 31 27 27 23 -- 00 18 -- 00 24 -- 00* 18 -- 00 -- 23 19 19 24 -- 18 -- 00 -- 18 -- 24 15 00 -- 36 -- 00 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58

Control Rod Being Withdrawn Rod Position in Notches Withdrawn Full in = 00 Full Out =

Figure 5.1 WNP-2 Cycle 4 Control Rod Withdrawal Analysis Initial Control Rod Pattern

l.S NOTE: The NCPR operating limit shaIl be the maximum of this curve or the rated condition HCPR operating 1imit.

30 10 50 60 70 80 SO 100 TQTRL CQAF AEC I ACULRT ING FLQN (% ARTEO)

Figure 5.2 Reduced Flow MCPR Operating Limit For Normal Feedwater Temperature

1.6 NOTE:, The HCPR operating limit shall be the maximum of this curve or the rated condition HCPR operating limit.

30 40 50 60 TQTAL CQAE AECIAEULATING FLQW

?0 80 90 '00 L10

(% AATEO)

Figure 5.3 Reduced Flow MCPR Operating Limit For FFTR Operation I

CO CO I

18 JJgiR

~ ~ ~ 0 15.62 610 16.62 14- ." 0 2,680 ltd.l0 0

6,230 14.71

~ % ~

7,840 14.19 I 10,470 N.13 12 I 13,220 14.06 16,990 14.06 18,780 14.00 10- 21,690 l3.93

PERMISSIBLE 24,420 13.93 REGION OF 27.280 13.08 8- OP ERAT ION 30.160 12.24 33.0b0 ll.40 3b,860 10,47

a. 38.900 S.bb 0 10000 20000 30000 @0000 60000 4 1,830 S.66 Average Planar Exposure (MWD/MT} ¹4 760 777 Figure 7.1 Linear Heat Generation Rate (LHGR) Limit Versus Average Planar Exposure, ANF 8x8 Fuel I

CO 00

27 ANF-88-02 9.0 DDITIONAL REFERENCES 9.1 S. F. Gaines, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Ri d F I," ~XN-NF-81-21A, R 11 I, E N I C 9 I',

Richland, WA 99352, January 1982.

9.2 R. H. Kelley, "Exxon Nuclear Plant Transient Methodology for Boiling Ilt R 9,"~8-87-78-717, R I I 2, E N I 0 8 y, Richland, WA 99352, November 1981.

9.3 J. E. Krajicek, "WNP-2 Cycle 4 Plant Transient Analysis," ANF-88-01, Advanced Nuclear Fuels Corporation, Richland, WA 99352, January 1988.

9.4 l. E. 2 'I k, "EIICA 8 k Rp t f BIIR 5," ~XN-NF-85-128 P, E Nuclear Company, Inc., Richland, WA 99352, December 1985.

9.5 D. J. Braun, "WNP-2 LOCA-ECCS Analysis, MAPLHGR Results," XN-NF-85-139, Exxon Nuclear Company, Inc., Richland, WA 99352, December 1984.

9.6 M. H.

141 d Smith, "Generic Mechanical F I," ~XN-Ef- I- I, R Design 1*1 for Exxon Nuclear Jet I, 8 881 t I, E Pump N

BWR Company, f 0 ig Inc., Richland, "Exxon Nuclear Methodology dA WA 99352, March 1985.

for Boiling lyi,"ENNNF..19AA,RI Water Reactors-Neutronics I ddddi t,EMethods Nuclear Company, Inc., Richland, WA 99352, May 1980.

9.8 "Generic Mechanical Design for Exxon Nuclear Jet 'Pump BWR Reload Fuel,"

~XE-Ny- -87 A, R I I I, E N I 0 p y, I ., Ill 11 d, NA 99352, September 1986.

9.9 "Exxon Nuclear Methodology for Boiling Water Reactors Neutronics Methods f 0 Ig A lyi,'~IPNF- -I A, Ill I, Rppl t I d 2, Exxon Nuclear Company, Inc., Richland, WA 99352, March 1983.

9. 10 J. B. Edgar, Letter to WPPSS, Supplemental Licensing Analysis Results, ENWP-86-0067, Exxon Nuclear Company, Inc., Richland, WA 99352, April 15, 1986.

9. 11 J. E. Krajicek, "WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction," XN-NF-87-92, Advanced Nuclear Fuels Corporation, Richland,, WA 99352, June 1987.

A-1 ANF-88-02 APPENDIX A Single Loop Operation (SLO)

ANF recently performed analyses for WNP-2 which demonstrate the safety of plant operation with a single recirculation loop out of service 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 WNP-2 single loop operation (SLO). The results of the SLO analyses are summarized below:

o The two loop MCPR operating lsmsts (rated condstions) bound the transient requirements for SLO. The single loop transient analyses need not be performed on a cycle by cycle basis and the two loop MCPR operating limits applicable for a cycle are appropriate for single loop conditions for that cycle.

o The postulated pump seizure accident, evaluated for SLO conditions, is calculated to have a less severe radiological release than the LOCA. The radiological consequences of this postulated accident are bounded by the radiological evaluation performed by General Electric (GE) for the LOCA and are well within the 10 CFR 100 limits.

o 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 consistent with the flow dependent MCPR curve (1.35 at 50 percent of rated flow). Single loop operation 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 10 CFR 50.46 for loss-of-coolant accident breaks up to and including the double-ended severance of a reactor coolant pi'pe.

The transient and pump seizure accident'nalyses are described in ANF-87-119 and the LOCA analyses are described in ANF-87-118.

A-2 ANF With a single recirculation loop in operation, the Gf analyses supported continued operation with an increase of 0.01 in the HCPR safety limit. ANF performed a single loop HCPR safety limit calculation 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 4 operation with both recirculation loops in operation, the HCPR safety limit is 1.06, which is the same value as was used for the previous cycles. For Cycle 4 operation with a singl recirculation-loop-.in. ser vice, the HCPR safety limit is 1.07, which is a the same value used for the previous cycles.

ANF-88-02 Issue Date: ~/15/88 WNP-2 CYCL'E 4 RELOAD ANALYSIS

.00i*t 'Ob

0. C. Brown R. E. Collingham R. A. Copeland L. J. Federico M. J. Hibbard J. G. Ingham S. E. Jensen T. H. Keheley J. E. Krajicek J. L. Haryott J. N. Morgan J. C. Rawlings (ENSA)

A. Reparaz

'.G. L. Ritter H. E.

R. Tandy Williamson J. B. Edgar/WPPSS (50)

Document Control (5)

ENCLOSURE 2 Ce

.9003080150 SUPPLY SYSTEM/NRC - REGION V MANAGEMENT MEETING JANUARY 18, 1990 WALNUT CREEK, CA AGENDA INTRODUCTION D. W. MAZUR 5 MIN II. SALP STATUS

OVERVIEW OF C. M. POWERS 5 MIN OPERATIONS ACTIVITIES

MAINTENANCE R. L. WEBRING 60 MIN

- OPERATIONS C. M. POWERS '0 MIN

ENGINEERING TECHNICAL J. P. BURN 30 MIN SUPPORT

SAFETY ASSESSMENT/ G. D. BOUCHEY 30 MIN QUALITY VERIFICATION III. FASTENER ISSUES C. M. POWERS 15 MIN IV.

SUMMARY

A. L. OXSEN 10 MIN

OPERATIONS

MAINTENANCE ENHANCEMENT PROGRAM

TECH SPEC IMPROVEMENT PROGRAM

TECHNICAL SUPPORT RADIOLOGICAL WORK PRACTICES/EFFLUENT MONITORING ISSUE

ADHEREhKE TO PROCEDURES t

EQUIPMENT OPERABILITY DETERMINATIONS

0 MNP-2 MAINTENANCE INITIATIVES

DRIYEN BY:

f NRC CONCERNS (SALP/SSOMI REPORTS) o PROCEDURAL INADEQUACIES AND MEAKNESS RESULTING IN OVER-RELIANCE ON '"SKILL OF THE CRAFT" o WORK CONTROL PROCESS INADEQUACIES-DETAIL, CONTENT, RIGOR AND COMPLIANCE o, PLANT MATERIEL CONDITION INCLUDING WORK BACKLOG/DEFERRAL OF LONG-TERM CORRECTIYE MAINTENANCE

WNP-2 MAINTENANCE INITIATIVES SUPPLY SYSTEM MANAGEMENT PERSPECTIVE (FEEDBACK FROM INTERNAL AUDITS AND CONTRACTED AUDITORS) INCLUDING:

o SUPPl Y SYSTEM QA MAINTENANCE ASSESSMENT o INPO EVALUATION REPORT o SUPPLY SYSTEM CONTRACTED AUDITS (IMPELL AND HARE) o MAINTENANCE SELF-ASSESSMENT, o INCREASED EXPECTATIONS FOR MAINTENANCE BASED ON INDUSTRY TRENDS PERFORMANCE DURING AND FOLLOWING THE SPRING, 1989 OUTAGE o SHUTDOWN COOLING ISOLATIONS o REACTOR SCRAM RESULTING FROM PERSONNEL ERROR - 8/17/89 o OTHER PERSONNEL/PROCEDURE REt ATED ERRORS

WNP-2 MAINTENANCE INITIATIVES

CONCLUSIONS:

GENERAL CONSENSUS OF EVALUATION FINDINGS NEED FOR IMPROVEMENT IN THOSE AREAS IDENTIFIED BY SALP/SSOMX ADDITIONAL XSSUES FOR IMPROVEMENT INCLUDE:

o EXCESSIVE CONTROL ROOM DEFICIENCIES o PREVENTIVE MAINTENANCE PROGRAM o TRENDING OF EQUIPMENT FAILURES o MAINTENANCE TRAINING o CRAFT TASK ASSIGNMENT AND COORDINATION OF WORK ACTIVITIES

FUNDING HAS BEEN ALLOCATED FOR THIS FISCAL YEAR AND IS PLANNED FOR FUTURE BUDGET CYCLES

PROCEDURE UPGRADE PLAN

GOALS/OBJECTIVES IMPROVE EXISTING PROCEDURES:

CONTENT/LEVEL OF DETAIL TECHNICAL ACCURACY SETPOINTS/TOLERANCES TOOLING/TEST EQUIPMENT REQUIREMENTS MORKING CONDITIONS AND LIMITATIONS I

IDENTIFY AND DEVELOP HElre PROCEDURES INCORPORATE LESSONS LEARNED INCORPORATE IN A COMMON FORMAT--

OTHER DEPARTMENTS AND INPO GUIDELINES IMPROVE "HUMAN FACTORS" ELEMENT INSTITUTE A VALIDATION AND VERIFICATION REVIEM - ALL PROCEDURES

PROCEDURE UPGRADE PLAN I

STAFFING STATUS FULL STAFFING:

1 SUPERVISOR AND 7 + WRITERS CURRENTLY:

1 SUPERVISOR AND 4 WRITERS COMPRISED OF MAINTENANCE/

CONTRACT ENGINEERS AND TECHNICIANS FULL STAFFING BY 3/1/90 ALONG WITH COMPl ETION OF FACILITY STAFF ASSIGNMENTS WILL BE FULL TIME INCLUDING OUTAGES

PROCEDURE UPGRADE PLAN

SCHEDULE UPGRADES COMPLETE 1sv QUARTER 1992 PRIORITY ASSIGNED TO PREVIOUSLY IDENTIFIED CRITICAL AREAS (RCA/LERs/NOV)

PROCEDURES CRITICAL TO PLANNED PLANT EVOLUTIONS GIVEN PRIORITY (EG. EXCESS FLOW CHECK VALVE TESTING)

, SURVEILLANCES UPGRADED IN CONJUNCTION WITH THE TECH SPEC IMPROVEMENT PROGRAM

PROCEDURE UPGRADE PLAN

TO DATE A REVERIFICATION HAS BEEN CONDUCTED TO ENSURE EACH TECH SPEC SURVEILLANCE REQUIREMENT HAS BEEN MET-NO DISCREPANCIES MERE IDENTIFIED DETAILED REVIEWS OF SELECTED SURVEILLANCE

.PROCEDURES BY THE SUPPLY SYSTEM SSFI TEAM HAVE DETERMINED THAT TECH SPEC REQUIREMENTS ARE ADEQUATELY ADDRESSED AND DOCUMENTED.

PROCEDURE WRITER'S GUIDE/HUMAN FACTORS PLAN FOR WNP-2

SHORT TERM - UTILIZE THE MAINTENANCE PROCEDURE WRITER'S GUIDE WHICH INCORPORATES INPO GUIDELINES PROVIDE 'EACH WRITER WITH HUMAN FACTORS PRINCIPLE'S TRAINING PROVXDE EACH WRITER WITH TRAINING ON THE EXISTING WRITER'S GUIDE

PROCEDURE WRITER'S GUIDE/HUMAN FACTORS PLAN FOR WNP-2

LONG TERM - DEVELOP AND IMPLEMENT THE WNP-2 WRITER'S GUIDE APPLICABLE TO MAINTENANCE AND OPERATIONS WITH SOME DEPARTMENT"-SPECIFIC GUIDELINES COMPLETE AND AVAILABLE FOR USE BY 4/1/90 UTILIZE LESSONS LEARNED FROM INDUSTRY EOP AUDITS AND UPGRADE PROCESS ESTABLISH METHODOLOGY FOR PERFORMING

- VERIFICATION AND VALIDATION OF PLANT PROCEDURES WILL HELP TO ENSURE:

TECHNICAL ACCURACY INCORPORATION OF HUMAN FACTORS PRINCIPLES PROCEDURE USEABILITY OPERATIONAL CORRECTNESS

PROCEDURE COMPLIANCE

I & C SURVEILLANCE EFFORT, R-4 TO PRESENT CRITICAL SURVEILLANCES INITIALLY LIMITED TO SPECIFIC CRAFT FULL TIME SUPERVISION OF CRITICAL SURVEILLANCES IN-DEPTH REVIEW OF SURVEILLANCE PRACTICES INTERVIEWS OF CRAFT, SUPERVISION, ENGINEERS DEVELOPED A DETAILED SURVEILLANCE WORK PRACTICE DOCUMENT TRAINING CONDUCTED BY THE I & C SUPERVISOR WITH EACH TECHNICIAN ELIMINATED REQUIREMENT FOR FULL-TIME SUPERVISION OF SURVEILLANCE ACTIVITIES

PROCEDURE COMPLIANCE

IMPROVED PROCEDURES TECHNICIAN FEEDBACK ON PROCEDURAL INADEQUACIES HAS INCREASED DRAMATICALLY TECHNICIANS UNMILLING TO "MAKE" A PROCEDURE MORK - REQUIRING DEVIATIONS OR REVISIONS TO REMOVE ERRORS PROCEDURE IMPROVEMENTS INCLUDE HUMAN FACTOR ELEMENTS

PROCEDURE COMPLIANCE

DISCIPLINE FOR COMPLIANCE PROBLEMS DISCIPLINE INCLUDING TIME OFF WITHOUT PAY, PERSONNEL LETTERS ON FILE AND LIMITATIONS ON WORK ASSIGNMENTS HAVE BEEN ENACTED DRIVEN HOME THE MESSAGE OF PROCEDURAL COMPLIANCE