ML17279A881

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
Download: ML17279A881 (140)

Text

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

PREVENTIVE MAINTENANCE PROGRAM UPGRADE

• PHASE I REVIEW GOALS AND OBJECTIVES o COORDINATE PM ACTIVITIES MXTH ONGOING CORRECTIVE MAXNTENANCE o MINIMIZE XNEFFXCIENCIES XN THE EXISTING PM PROGRAM o ELIMINATE MULTIPLE VISITS TO COMPONENTS o ELIMINATE TIME DEPENDENT ACTI'QTIES THROUGH CONDITION MONITORING o SUPPORT THE PHASE II EFFORT IN IMPLEMENTATION o DEVELOP SUPPLY SYSTEM READINESS TO CONTINUE RCM APPLICATION AT CONTRACT END o PROVIDE ENGINEERING SUPPORT OF THE PHASE II EFFORT VIA PERFORMING THE PRA ANALYSES OF PLANT SYSTEMS o ESTABLISH A SUPPLY SYSTEM REVIEW TEAM

PREVENTIVE MAINTENANCE PROGRAM UPGRADE

• PHASE I REVIEW STAFFING

'0 FULL STAFFIHG LEVEL:

1 SUPERVISOR AND 4 REVIEWERS CURRENT STAFFING 1 SUPERVISOR AHD 3 REVIEWERS COMPRISED OF MAINTENANCE/CONTRACT ENGINEERS AHD SELECTED CRAFT PERSONNEL FULL STAFFING BY 3/1/90 o STAFF ASSIGNMENTS WILL BE FULL TIME EFFORT WILL CONTINUE FOR A MINIMUM OF 2 YEARS

PREVENTIVE MAINTENANCE PROGRAM UPGRADE

• PHASE II REVIEW GOALS AND OBJECTIVES

.IMPROVE THE EFFICIENCY AND EFFECTIVENESS OF APPLIED MAINTENANCE EFFORTS REDUCE PROGRAM SCOPE THROUGH DIRECTED EFFORTS AT CRITICAL COMPONENTS WHERE PERFORMANCE CAN BE INFLUENCED BY PM OR WHERE FAILURE MEASURABLY IMPACTS PLANT SAFETY OR AVAILABILITY

~ . 0 PREVENTIVE MAINTENANCE PROGRAM UPGRADE

• PHASE II REVIEM RELIABILITY CENTERED MAINTENANCE APPROACH o EVALUATION OF ALL MNP-2 SYSTEMS o APPLY RCM TO SELECTED SYSTEMS UTILIZE SYSTEM PRA ANALYSES,

'QUIPMENT HISTORY, INDUSTRY HISTORY, PLANT ENVIRONMENTAL AND SERVICE CONDITIONS, SAFETY SIGNIFICANCE AND VENDOR RECOMMENDATIONS TO DEVELOP COMPONENT RECOMMENDATIONS DEVELOP REVISED PROGRAM FOR PLANNED MAINTENANCE, CONDITION MONITORING, COMPONENT REPLACEMENT, AND IDENTIFY RECOMMENDED DESIGN CHANGES DEVELOP PROCEDURES TO SUPPORT RECOMMENDED ACTIVITIES DEVELOP A LIVING RCM PROGRAM TO BE CONDUCTED BY THE SUPPLY SYSTEM

PREVENTIVE MAINTENANCE PROGRAM UPGRADE

• PHASE II REVIEM SCHEDULE PRE-SELECTION OF 5 POTENTIAL CONTRACTORS FROM A FIELD OF 17 COMPLETED

. REQUEST FOR PROPOSALS TO BE ISSUED 1/19/90 NEGOTIATIONS AND RECOMMENDATIONS FOR AWARD IN MARCH, 1990 o CONTRACT AMARD SCHEDULED FOR APRIL MOBILIZATION ON SITE AS EARLY AS MAY RCM REVIEM PERIOD APPROXIMATELY 2 YEARS

PREVENTIVE MAINTENANCE PROGRAM UPGRADE

• PHASE II REVIEW STAFFING o 7 MEMBER SELECTION PANEL APPOINTED TO GUIDE PROCUREMENT o -

PHASE I STAFF IN PLACE TO SUPPORT CONTRACT EFFORTS

.0 SUPPLY SYSTEM ENGINEERING CURRENTLY WORKING ON WNP-2 PRA ANALYSES CONTRACTOR STAFFING TO INCLUDE A MINIMUM OF 15 PEOPLE

WORK PROCESS IMPROVEMENTS

• APPROACH ASSIGNMENT OF A MAINTENANCE SUPERVISOR, FULL TIME FOR 3+ MONTHS REVIEW OF THE PROCESS FOR 5 OTHER UTILITIES REVIEW CONCERNS OF NRC/INPO/INTERNAL AUDITS CONSIDERED KNOWN INEFFICIENCIES COMMON TO WNP-2 USERS

8 WORK PROCESS IMPROVEMENTS

• GOALS REDUCE DEPENDENCY ON "SKILL OF THE CRAFT" IMPROVE PACKAGE CLARITY - AVOID MISUNDERSTANDINGS ADDRESS HUMAN FACTORS PRINCIPLES IN PACKAGING ACHIEVE INCREASED CRAFT ACCOUNTABILITY DEVELOP COMMONALITY OF CONTENT AND FORMAT IMPROVE WORK DOCUMENTATION AND FEEDBACK FROM .

CRAFT PERSONNEL INCREASE EFFICIENCY IN WORK IMPLEMENTATION THROUGH MORE ACCURATE DETAILED INSTRUCTIONS TO THE CRAFT

WORK PROCESS IMPROVEMENTS

• WORK PACKAGING AND TRAINING DEVELOP STANDARDS FOR PACKAGES (EG. TOOLING, PARTS, SETPOINTS, TOLERANCES, SETTINGS)

REQUIRE PARTS STAGXNG AND DOCUMENT INCORPORATION INTO EACH PACKAGE.

UTILIZE A COMMON FORMAT TO ASSIST IN THE REVIEW AND IMPLEMENTATION OF THE PACKAGE DEVELOP AND IMPL'EMENT A TRAINING PROGRAM PRIOR TO XMPLEMENTATION

WORK PROCESS IMPROVEMENTS

• TRANSITION TO COMPUTER DEVELOPED PACKAGES

'I "

ONE SHOP CURRENTLY CONVERTING TO PC DEVELOPED PACKAGES NEW PROCESS BEING DEVEl OPED TO COMPLEMENT PC DEVELOPED PACKAGES REMAINING MAINTENANCE SHOPS WILL CONVERT TO PC DEVELOPED PACKAGES MITHIN THE YEAR

WORK PROCESS IMPROVEMENTS

• SCHEDULE PROCEDURE DRAFT BY THE END OF JANUARY 1990 PROCEDURE POC APPROVAL BY MID FEBRUARY 1990 TRAINING COMPLETE AND IMPLEMENT BY MARCH 1990

e WORK CONTROL PROGRAM

• GOALS IMPLEMENT A "DEMAND SCHEDULE" SUPPORTING GOALS, NEEDS AND PRIORITIES OF THE PLANT IMPLEMENT AN EFFECTIVE "WORK COORDINATION .FUNCTION" TO SUPPORT DEMAND SCHEDULE

S WORK CONTROL PROGRAM

• IMPROVE THE "READY TO WORK" PROCESS INCREASE COMMITMENT AND ACCOUNTABILITY FOR WORK PACKAGE PREPARATION INCREASE EMPHASIS WITHIN SUPPORT ORGANIZATIONS FOR ACHIEVING "READY TO WORK" STATUS

WORK CONTROL PROGRAM

• COORDINATION IMPROVEMENT IMPROVE METERING OF WORK TO THE CONTROL ROOM-DECREASE CHALLENGES TO PLANT DEVELOP MANAGEMENT FEEDBACK INCREASE ACCOUNTABILITY ACROSS DISCIPLINES INCREASE COORDINATION BETWEEN MAINTENANCE AND SUPPORT ORGANIZATIONS HELP ELIMINATE BARRIERS WHICH SLOW OR STOP PLANNED WORK

WORK CONTROL PROGRAM

• ORGANIZATION STRUCTURE AND STAFFING CHARTERED IN DECEMBER, 1989 IMPLEMENT IN FEBRUARY, 1990 NEW GROUP WITH A FULL TIME SUPERVISOR IN THE PLANNING 5 SCHEDULING DEPARTMENT I

TOTAL OF 9 MEMBERS,. AT LEAST 6 IN A FULL TIME STATUS INITIALLYHEADED BY THE ASSISTANT OPERATIONS MANAGER STAFFED BY HAND-PICKED INDIVIDUALS FROM EACH DEPARTMENT, INCLUDING THOSE WHO CURRENTLY HOLD SUPERVISORY POSITIONS

WORK CONTROL PROGRAM

• 'EQUESTED INPO ASSISTANCE SS REQUESTED AN INPO ASSIST VISIT DIRECTED AT WORK CONTROL TEAM WILL BE ON-SITE THE FIRST OF FEBRUARY GOAL TO OBTAIN CRITICAL AND TIMELY COUNSELING DURING THE START-UP PHASE OF THIS EFFORT

DEFERRAL OF LONG TERM CORRECTIVE MAINTENANCEI PLANT MATERIEL CONDITION/BACKLOG REDUCTION

• CORRECTION OF LONG STANDING ISSUES MSIV GALLING REPAIRS COMPLETE IN SPRING 1990 OUTAGE SRV VACUUM BREAKER LEAKAGE REPAIRED, CONTAINMENT UNIDENTIFIED LEAKAGE BELOW

.5 GPM SRV REBUILD PROGRAM ONGOXNG WITH COMPLETION PLANNED XN 1991 REPLACEMENT OF THE RILEY LEAK DETECTION MODULES IN 1989 REPLACEMENT OF ALL MAIN TURBINE LOW PRESSURE ROTORS IN 1991

)

DEFERRAL OF LONG TERM CORRECTIVE MAINTENANCE/

'PLANT MATERIEL CONDITION/BACKLOG REDUCTION

• ATTENTION TO GENERIC CONCERNS INITIATING A LIVE-LOAD VALVE PACKING PROGRAM IN CONTAINMENT IN 1990 MOV UPGRADE PROGRAM ONGOING FOR ALL PLANT MOVs MOV DESIGN BASIS TESTING PROGRAM UNDERMAY-GENERIC LETTER 89-10 CRD HCU VALVE REFURBISHMENT EFFORT BEGINNING IN 1990

DEFERRAL OF LONG TERM CORRECTIVE MAINTENANCE/

PLANT MATERIEL CONDITION/BACKLOG REDUCTION

• ONGOING ISSUES SEMI-ANNUAL PLANT CLEANUP EFFORT ESTABLISHED BACKLOG REDUCTION PROGRAM INSTITUTED FOR MWRs/PMs INCREASED VISIBILITYOF LONG STANDING PROBLEMS THROUGH THE PLANT WEEKLY REPORT ONGOING PAINTING PROGRAM

WNP-2 EQUIPMENT TRENDING PROGRAMS

• PERFORMANCE MONITORING ON-GOING PROGRAM SHARED BETWEEN PLANT TECHNICAL/PLANT MAINTENANCE BASED ON VIBRATION MONITORING,.

OIL ANALYSIS, TRENDING OPERATIONAL PARAMETERS, THERMOGRAPHY, MOVATS PROVIDES HISTORY FOR TECH SPEC/ASME COMPONENT TRENDS PROVIDES A HISTORY OF SELECTED PLANT COMPONENT OPERATIONAL PARAMETERS AT GIVEN FREQUENCIES HELPS IDENTIFY PROBLEMS BEFORE THE FAILURE STAGE IS REACHED

WNP-2 EQUIPMENT TRENDING PROGRAMS EQUIPMENT FAILURE TRENDING NEWLY INSTITUTED IN MAINTENANCE PERFORMED ON A 6 MONTH FREQUENCY-REVIEW OF PLANT FAILURE HISTORY REQUIRES DETAILED REVIEW OF COMPONENTS WHICH EXCEED 20 FAILURES IN PLANT LIFE, 3 IN THE PAST 12 MONTHS OR SHOW AN INCREASING TREND RECOMMENDATIONS FOR INCREASED CONDITION MONITORING, REVISED WORK PRACTICES, EQUIPMENT CHANGEOUT RESULT FROM THE REVIEW PROVIDES A HARD COPY REPORT FOR EQUIPMENT HISTORY ON RESULTS OF THE EVOLUTION

WNP-2 EQUIPMENT TRENDING PROGRAMS

• FUTURE IMPROVEMENTS RCM RECOMMENDATIONS FOR CONDITION MONITORING WILL'STABLISH THE BASIS OF THE PERFORMANCE MONITORING PROGRAM XMPROVEMENTS IN WORK PROCESS (MWR) PROGRAM WILL RESULT IN MORE DETAILED AND ACCURATE FAXLURE DATA

MAINTENANCE/TRAINING INITIATIVES

• GOALS AND OBJECTIVES RE-EVALUATION IN PROCESS TO ESTABLISH PERFORMA CE BASED OBJECTIVES AND MEASURES OF MAINTENANCE ACTIVITIES RESTRUCTURING OF THE EXISTING TRAINING PROGRAM WILL FALL OUT OF THIS RE-EVALUATION

MAINTENANCE/TRAINING INITIATIVES

• JOB AND TASK ANALYSIS (JTA)

SUPPLY SYSTEM CONTRACTED (AUGUST, 1989) JOB AND TASKS ANALYSIS OF THE 3 MAINTENANCE DISCIPLINES PRODUCTS MILL INCLUDE: TRAINING OBJECTIVES, JOB PERFORMANCE MEASURES, AND INSTRUCTIONAL SEQUENCING RESULT MILt BE TO DEVELOP THE BASIS 'FOR FUTURE TRAINING PROGRAMS BASED ON ACTUAL TASKS REQUIRED SCHEDULED TO COMPLETE IN JUNE, 1990

P MAINTENANCE/TRAINING INITIATIVES

• ON THE JOB TRAINING (OJT)

TASK IS TO: 1) EVALUATE TRAINING PERFORMANCE

2) ASSIST IN OJT 3) SERVE AS PLANT POINT OF CONTACT WITH TRAINING THREE FULL TIME TRAINERS HIRED AND ASSIGNED REMOVES THIS BURDEN FROM THE MAINTENANCE SUPERVISOR

MAINTENANCE/TRAINING INITIATIVES

• QUALITY ASSESSMENT TEAM (OAT)

CHARTERED BY THE SUPPLY SYSTEM QUALITY COUNCIL TASKED WITH IDENTIFYING ISSUES CRITICAL TO THE SUCCESSFUL IMPLEMENTATION OF THE MANY CHANGES UNDERWAY IN MAINTENANCE/TRAINING CHAIRED BY THE MANAGER OF MAINTENANCE TRAINING 8 MEMBERS FROM YARIOUS LEVELS WITHIN MAINTENANCE AND TRAINING

MAINTENANCE/TRAINING INITIATIVES

• MAINTENANCE ASSIGNMENT OF PERSONNEL GOAL TO CLARIFY THE BASIS OF CRAFT WORK ASSIGNMENT CONVERTING TO COMPUTER BASED SYSTEM FOR IDEHTIFYIH CRAFT TRAINING ESTABLISH FORMAL DOCUMENTATION OF CRAFT QUALIFICAT BASIS ASSURE PERSONNEL QUALIFICATIONS BY DEMONSTRATED PERFORMANCE INCORPORATE THIS REVISED PROCESS INTO THE WORK PLANNING EFFORT FULLY IMPLEMENT BY 3/1/90

MAINTENANCE MANAGEMENT/SUPERVISORY REALIGNMENT

• GOAL IMPROVE WORK CONTROL, PACKAGE PREPARATION AND TASK ACCOUNTABILITY IN EACH DISCIPLINE SHOP

MAINTENANCE MANAGEMENT/SUPERVISORY REALIGNMENT

• ACTION TAKEN SHOP RESTRUCTURING - ADDITION OF WORK CONTROL AND ENGINEERING SUPERVISORS ASSIGNMENT OF HAND-PICKED, EXEMPT CRAFT SUPERVISORS IN EACH SHOP ASSIGNMENT OF PARTS/MATERIALS HANDLERS AND SHOP PLANNERS IN EACH SHOP PROVIDES THE ORGANIZATIONAL STRUCTURE NECESSARY IN TAKING THE NEXT STEPS IN MAINTENANCE IMPROVEMENT

0 CONTROL ROOM DEFICIENCY REDUCTION PROGRAM GOAL REDUCE THE NUMBER OF DEFICIENCIES TO 50 BY 1/1/90 REDUCE THE NUMBER OF DEFICIENCIES TO 25 AT THE END OF THE SPRING 1990 OUTAGE

CONTROL ROOM DEFICIENCY REDUCTION PROGRAM

• ACTIONS TAKEN MANAGEMENT "WHITE PAPER" DEVELOPED OUTLINING PROGRAM ELEMENTS o ASSIGNMENT OF A NEW PRIORITY WORK CLASS BETWEEN .1 AND 2 o REQUIRE WORK INSTRUCTION

.COMPLETION WITHIN 3 DAYS OF PROBLEM,INDENTIFICATION o SCHEDULE TO WORK JOBS WITHIN 3 DAYS OF ACHIEVING RTW STATUS o EXPEDITE PARTS PROCUREMENT o UTILIZE SHOP OVERTIME AS NECESSARY o INCREASE DESIGN CHANGE PRIORITY WITHIN ENGINEERING

CONTROL ROOM DEFICIENCY REDUCTION PROGRAM

)

• RESULTS TO DATE/PLANS NUMBER OF DEFICIENCIES LOWERED TO 74 FROM 94 IN OCTOBER 53 TASKS ARE CURRENTLY OUTAGE-RELATED ENGINEERING WILL COMPLETE THE DESIGN FOR OVER 20 PACKAGES PRIOR TO R-5 THE POST R-5 GOAL IS STILL ACHIEVABLE

TECHNICAL SPECIFICATION IMPROVEMENT PROGRAM

• COMPLETE REWRITE OF LCOs; EXPANSION OF BASES

• PROCEDURE TO REQUIREMENT CROSS-CHECK

• SSFI REVIEM OF APPLICABLE PROCEDURES

• MODE CHANGE SURVEILLANCE REVIEW

• VALIDATION AND-VERIFICATION ON REVISED PROCEDURES

TECHNICAL SUPPORT

• . ADDED RESOURCES FOR BACKLOG REDUCTION EFFORTS AND SYSTEM ENGINEERING SUPPORT

• COMPLIANCE SUPPORT FOR REPORTABLITY DETERMINATIONS FOR OPERATIONS

• TECHNICAL STAFF TRAINING COURSES ON:

10 CFR 50.59 ROOT CAUSE ASSESSMEHTS PROJECT MANAGEMENT PARs MWR WORK PACKAGE PREPARATION PMT

• IMPROVED LONG RANGE PLANNING

• DEVELOPED A JOINT TECH STAFF/

GENERATION ENGINEERING SYSTEM WALKDOWN PROCESS

IMPROVEMENTS IN RADIOLOGICAL WORK PRACTICES

• IMPROVED ROR IDENTIFICATION, ROOT CAUSE ASSESSMENTS

. 5, TRENDING OF RESULTS

• ESTABLISHED HP SPONSOR PROGRAM

• HP SUPERVISION ASSISTS IN RAD. REFRESHER TRAINING COURSES

• DEVELOPING MORE MOCK-UP TRAXNING FOR CRAFTS

• CONDUCTING STRENGTHENED ALARA PRE-JOB BRIEFINGS

• IMPROVED RAD. WORK PRACTICES EMPHASIZED BY CRAFT SUPERVISION, FIELD WALKDOWNS

• HP SPONSORED NUMEROUS DOSE REDUCTION INITIATIVES CRD ROOM MODS SYSTEM FLUSHES MODXFIED SHUTDOWN SEQUENCES

• PERFORMANCE INDICATORS PROVIDE POSITIVE FEEDBACK

0 ADHERENCE TO PROCEDURES

• TQA AND OAT - TRAINING ON QUALITY IMPROVEMENTS

• PER PROCESS HPES PEER ROOT CAUSE

• DISCIPLINE

• SURVEILLANCE WORK DOCUMENT/TRAINING

0 NEER I NG IMPROVEMENT PROGRAM 1/1/90 6/88 6/89 6/90 6/91 'ASKS EIP CATEGORY FY-89 FY-90 FY-90 I I I I I I PLANNING & SCHEDULING 5 DONE 9 TOTAL FEEDBACK SYSTEMS & COMMUNICATIONS 5 DONE 6 TOTAL TECHNICAL LEADERSHIP 6 DONE 7 TOTAL INTERORGANIZATIONALINTERFACES 6 DONE 12 TOTAL PROCESS IMPROYEMENTS 11 DONE I I I I 33 TOTAL I I TRAINING 8 DONE 25 TOTAL.

TOOLS 2 DONE 10 TOTAL MORALE 3 DONE 5 TOTAL TOTALS:

46 DONE 107 TOTAL LEGEND: ~ SCHEDULE EKBRI PERCENT. ACHIEVED TODAY 902004A

w ENGINEERING IMPROVEMENT PROGRAM EFFECTIVENESS FGRs ERRORS 600 300 500 250 400 200 FCRs 300 150 200 100 ERRORS SINGLE DATA POINT (ORB) 100 50 0 0 1987 1988 1989 1990

ENGINEERING IMPROVEME ROGRAM EFFECTIVENESS QUALITY CIRCLE RESULTS TECHNICAL CONSTRUCT- MOD TEST & WALKDOWN DESCRIPTION MERIT ABILITY OPERABILITY EFFECTIVENESS BDC 86-0617 IN-1 INVERTER 5.0 5.0 4.0 BDC 86-0273 FW HTR DUAL LEVEL CONTROL 5.0 4.0 4.8 4.0 BDC 84-0542-1 RRC PUMP SEAL FLOW NOT INSTRUMENTATION EVALUATED 2.5 3.0 5.0 LEGEND 1.0 - UNSATISFACTORY 2.5 - AVERAGE 5.0 - EXCELLENT

DESIGN REQUIREMENTS PROGRAM 1/1/90 6/88 6/89 6/90 6/91 6/92 6/93 6/94 TASKS FY-89 FY-90 FY-91 FY-92 FY-93 FY-94 FY-95 I

'I I I I PILOT PROGRAM (LPCS & AC)

I I I I I REACTOR FEEDWATER STANDBY SERVICE WATER I I. I I I STANDBY ELECTRICAL I I I I I SEISMIC I I I I I ELECTRICAL SEPARATION I I I I I I I RESIDUAL HEAT REMOVAL I I I I I I I MAIN STEAM/MSLC HIGH PRESSURE CORE SPRAY I I I -

I I I I REACTOR CORE ISOLATION I I I I I I I STANDBY LIQUID CONTROL EQUIPMENT QUALIFICATION I I I I I I CONTROL SYSTEM FAILURE I I I I I I HUMAN FACTORS PIPE BREAK/MISSILE COMMITTMENTS DATABASE I I I I I,I I I I I I I LICENSING COMMITMENTS I I I I I I 4 SPECIAL TOPICS, I I I I- I I I 5 NSSS SYSTEMS I I I I I I I I I I I I I I 5 SPECIAL TOPICS, 5 BOP SYSTEMS & ALL I I I I I I STRUCTURES I I I I I I 4 SPECIAL TOPICS, I I I I I I 12 BOP SYSTEMS I I I I I I

~

2 SPECIAL TOPICS, I I I 25 BOP SYSTE MS LEGEND: SCHEDULED gRgg COMPLETED 902004.B

ELECTRICAL MIRING DIAGRAMS

• END'S COMPLETE J

MOV'S 463 SYSTEM LEVEL 240 TOTAL 703

• EWD'S PLANNED NEXT 6 MOS 330

• END'S PLANNED FY' 1991 330

CONFIGURATION MANAGEMENT IMPROVEMENT PROGRAM (CMIP)

PURPOSE ESTABLISH THE REQUIREMENTS FOR CONFIGURATION MANAGEMENT, I.E., THOSE REQUIREMENTS. WHICH WOULD ENSURE PLANT HARDWARE COMPLIES 14ITH AND IS ACCURAT REFLECTED IN PLANT DOCUMENTS DEVELOP AND IMPLEMENT A PLAN TO o REVIEW CURRENT ORGANIZATIONAL WORK PROCESSES TO THE ESTABLISHED REQUIREMENTS o PROVIDE RECOMMENDATIONS FOR IMPROVEMENTS

1/1/90 6/88 6/89 6/90 6/91 6/92 CMIP TASKS FY-89 FY-90 FY-91 FY-92 ESTABLISH COMMITTEE ESTABLISH RfQUIREMENTS/ISSUE NOS 32 DEVELOP CMIP PLAN IMPLEMENT CMIP PLAN REVIEW CURRENT PROCESSES/

DEVELOP RECOMMENDATIONS ISSUE RECOMMENDATIONS MANAGEMENTREVIEW /APPROVAL OF RECOMMENDATIONS ESTABLISH MILESTONES & SCHEDULES IMPLEMENT APPROVED SCHEDULES LEGEND: ~ SCHEDULE EKKKIPERCENT ACHIEVED 902004

7 Wjg~,P SETPOINT PROGRAM TASK PLAN 'STATUS 1.0 IDENTIFY HARSH ENVIRONMENT 10/31/89 COMPLETE EQUIPMENT 2.0 PERFORM SAFE SHUTDOWN ANALYSIS 12/31/89. COMPLETE 3.A REVISE METHODOLOGY USING ISA RP67.04 12/31/89 COMPLETE .

3.B TABULATE TESTED SETPOINT ACCURACY 12/31/89 COMPLETE FROM EQ DATA 3.C RECALCULATE SETPOINTS 3/31/90 BEHIND SCHEDULE 4.0 RESOLVE SETPOINT OPERATIONAL 6/30/90 NOT STARTED PROBLEMS 5.0 REVISE PROCEDURES/RECALIBRATE 8/1/90 . NOT STARTED EQUIPMENT .

SAFETY ASSESSMENT/QUALITY VERIFICATION

• SALP ISSUES "AGGRESSIVENESS" OF OVERSIGHT GROUPS/RESPONSIVENESS IN RESOLVING PROBLEMS ORGANIZATION/STAFFING QUALIFICATIONS ROOT CAUSE PROGRAM EFFECTIVENESS QUALITY OF LICENSING SUBMITTALS

SAFETY ASSESSMENT/QUALITY VERIFICATION

• OVERALL PROGRAM STATUS/DIRECTION QUALITY IMPROVEMENT NUCLEAR SAFETY PROGRAMS PROCUREMENT QA QA/QC

EFFECTIVENESS ASSESSMENT

• PLANS APPROVED BY QUALITY COUNCIL (DIRECTOR)

STATUS REVIEWED MONTHLY BY QUALITY COUNCIL

• RESULTS REPORTED TO QUALITY COUNCIL EXPERT CONSULTANTS IN SOME CASES

EFFECTIVENESS ASSESSMENT CATEGORY STATUS/SCHEDULE

1) CHEMISTRY DRAFT COMPLETE Q COUNCIL BRIEFING 2/90

2) EMERGENCY PREPAREDNESS 2/90

3) ENGINEERING/TECH SUPPORT FOUR PHASE PLAN (8/90)

4) INDUSTRIAL SAFETY MULTIPHASE PLAN (12/90)

5) MAINTENANCE/SURVEILLANCE TWO PHASE PLAN (4/90 AND 1/91)

6) OPERATING EXPERIENCE THREE PHASES REVIEWS (2 COMPLETE) LAST PHASE SCHEDULED (4/90)

7) OPERATIONS SCHEDULED (2/90)

8) ORG AND ADMIN. MULTIPHASE PLAN (6/90)

9) OUTAGE MANAGEMENT SCHEDULED (9/90)

10) RADIOLOGICAL PROTECTION COMPLETED (ll/89)

11) SAFETY ASSURANCE/QUALITY PLAN UNDER DEVELOPMENT

12) SECURITY SCHEDULED 2/90

13) TRAINING AND QUALIFICATION PLAN UNDER DEVELOPMENT

1989 ASSESSMENT STATISTICS

• NUCLEAR SAFETY NUMBER INDUSTRY OPERATING EXPERIENCE REVIEWS: 490 NUCLEAR SAFETY/ENGINEERING ASSESSMENTS: 28 MAQOR TEAM INSPECTIONS (SSFI, SSOMI): 2

• QUALITY ASSURANCE CORPORATE QA AUDITS'NP-2 QUALITY SURVEILLANCES: 96 OFF-SITE VENDOR AUDITS/REVIEWS:

• QUALITY CONTROL INSPECTIONS MMRs REVIEWED: 5,994 MMRs ASSIGNED INSPECTION ."HOLD 'POINTS": 1, 294 OTHER 161 INSPECTIONS'ECEIVING INSPECTIONS (LINE ITEMS): 8,405

• TOTAL NUMBER OF QUALITY FINDINGS QFRs ISSUED: 289*

• INCLUDES 33 FINDINGS ASSOCIATED WITH,SSFI ISSUED AS PERS

~ ~

ORGANIZATION/STAFF QUALIFICATION

1) ORGANIZATION IMPROVEMENTS COMPLETE

2) STAFF INCREASED IN FY 90

3) RECRUITING SUCCESS INTERNAL 8 EXTERNAL 14 TOTAL 22

4) OPERATIONAL EXPERIENCE (NRC LICENSES 5 CERTIFICATES)

• 19 INDIVIDUALS

• SROs WNP-2 COMMERCIAL BMRs OTHER

• ROs 12 TOTAL 34

5) EDUCATIONAL QUALIFI CATIONS

• ADVANCED DEGREES 7 INDIVIDUALS (10 DEGREES)

• ENGR OR SCIENCE 33 INDIVIDUALS (43 DEGREES)

• CERTIFIED PEs 17 INDIVIDUALS

6) USE OF OUTSIDE EXPERTS

• CONSULTANTS

• UTILITY TRADES

• INTERNAL ASSIGNMENTS

t ROOT CAUSE ASSESSMENT PROGRAM STATUS

• IMPROVED PROBLEM REPORTING AND RCA MORE PROBLEMS REPORTED (976 IN 1989 vs 728 IN 1988)

MANAGEMENT REVIEW OF ALL PERs (MRC)

MORE FORMAL RCA (192 IN 1989 vs 29 IN 1988)

• INCIDENT INVESTIGATION PROCESS IMPLEMENTED ACTIVATED BY MANAGEMENT BROADER VIEW THAN NORMAL RCA

• TRAINING OF ADDITIONAL RCA STAFF RCA PROCESS TRAINING - 140 ENGINEERS IN-HOUSE RCA TRAINING - 17 ENGINEERS MORT TRAINING FOR RCA STAFF

• INCREASED EMPHASIS ON IMPLEMENTATION

• REPORT QUALITY/PRECURSOR ASSESSMENT

• INDEPENDENT EFFECTIVENESS ASSESSMENT

ROOT CAUSE ASSESSMENT PROGRAM STATUS (CON'T)

• EXAMPLES OF SIGNIFICANT EVENTS ANALYZED (1989)

INSULATOR FAULT (SCRAM 89-01)

ROD DRIFT EVENT SHUTDOWN COOLING EVENTS IN R-4 TURBINE BLADE CRACKS SCRAM DURING DEH TESTING (SCRAM 89-02)

RFW PUMP TRIP (SCRAM 89-03)

RFW THRUST BEARING FAILURE I & C SURVEILLANCE SCRAM (89-04)

EXTRACTION STEAM LINE EXPANSION JOINT FAILURE LIMITORQUE BOLT TORQUING ISSUE RESINS IN HVAC SYSTEM COOLING TOWER STRUCTURAL AND MECHANICAL (CONCRETE SPALLING)

QUALITY OF LICENSING SUBMITTALS

• QUARTERLY MEETINGS MITH NRR TO SPECIFICALLY ADDRESS THIS ISSUE

• NO KNOWN PROBLEMS SINCE ISSUANCE OF THE LATEST SALP

SAFETY ASSESSMENT/QUALITY VERIFICATION PROGRAM STATUS AND INITIATIVES

• QUALITY IMPROVEMENT MANAGEMENT TRAINING (QMS)

EMPLOYEE TRAINING ("THE QUALITY ADVANTAGE")

- 'EAMS PROBLEM SOLVING MANAGEMENT. EMPHASIS - QUALITY COUNCIL

NUCLEAR SAFETY ASSESSMENT

• INDUSTRY EXPERIENCE REVIEM

• 50.59 REVIEM IMPROVEMENTS PROCEDURES

'RAINING

• SSFI - AC DISTRIBUTION SYSTEM

• TECHNICAL ASSESSMENTS

• RELIABILITY/RISKDIRECTED OVERSIGHT

OER PERFORMANCE 180 16e 160 140 /ACTUAL 3i 132 126 120 106 N 102 U 1OO PLANNED M

Te Te 80 Ti R

T0 T1 T2 72 0 ei 40 20 D J F M A M J J A S 0 N D 1989 INDUSTRY OPERATING EXPERIENCE ACTIONS AWAITING IMPLEMENTATION 120 109 100 80 N

U M 60 8 60 E

R FY 90 GOAL < 40 40 i2 39 3B 32 20 J F M A M J J A S 0 N D 1989 OER REPORTS AWAITING REVIEW

QUALITY VERIFICATION

• PROCUREMENT OA RECEIVING INSPECTION VENDOR AUDITS SPECIFICATION/DOCUMENT REVIEWS

• TRENDING AND REPORTING

• OA/QC EFFECTIVENESS

SAFETY ASSESSMENT/QUALITY VERIFICATION PROGRAM STATUS AND INITIATIVES

SUMMARY

ADDRESSING SALP ISSUES "AGGRESSIVE" STATE-OF-ART SAFETY/QUALITY PROGRAMS EXCELLENT STAFF (DEDICATED/

QUALIFIED)

CONTINUOUS IMPROVEMENT o OPERATING EXPERIENCE (RCA) o NUCLEAR SAFETY ASSESSMENT o QA/QC o LICENSING PROGRAMS

SAFETY RELATED VALVE FASTENERS

• NOV RECEIVED 1/15/90

• SUPPLY SYSTEM CONCURS WITH VALIDITY OF THE VIOLATIONS

• INDUSTRY EXPERIENCE ALERTED SUPPLY SYSTEM TO POTENTIAL PROBLEMS WITH VIBRATION LOOSENING BOLTS

• SS DEVISED PREVENTIVE MAINTENANCE ACTIVITY WHICH WAS LATER PROVEN INADEQUATE

SAFETY RELATED VALVE FASTENERS

• ACTIVITY LACKED DELINEATION OF VIBRATION-SENSITIVE VALVES, POSITIVE CLAMPING FORCE VERIFICATION TECHNIQUE AND MANAGEMENT FEEDBACK ON EFFECTIVENESS

• ABSENT FEEDBACK, MANAGEMENT INACCURATELY BELIEVED PROBLEM WAS PRECLUDED

• GIVEN NEW PER PROCESS AND FORMAL ROOT CAUSE ASSESSMENTS. APPROPRIATE PM ACTIONS IN PLACE:

TORQUE VERIFICATION MONTHLY ENGXNEERED CAPTURE MECHANISM FOR SUSPECT VALVES FAILURE REPORTING

SAFETY RELATED VALVE FASTENERS

• EXPERIENCE INDICATES TORQUE SELECTION IS INADEQUATE

• SPECIFIC ACTIONS TO RE-TORQUE RHR-V-24A/B INADEQUATE

• Q/A SURVEXLLANCE ON BOLT TORQUING CONCLUDED:

- TRAINIHG NEEDED FOR WORK PACKAGE PREPARERS, QC OVERVXEWERS AND CRAFT IMPLEMEHTERS

- TORQUE SELECTXON NEEDS CLARIFICATION AND REVIEW

- GENERXC TORQUE SELECTION GUIDANCE (PPM10.2.10)

NEEDS CLARIFXCATXOH/RESTRICTIONS

• CORRECTIVE ACTIONS ON THESE ISSUES UNDER DEVELOPMENT