ML17311B303

Forwards Trip Repts Providing Results of 950829-30 NRC on- Site Discussion of Plant risk-ranking Methodology & Expert Panel Process & 950906-07 Assessment of Graded QA Program

ML17311B303
Person / Time
Site:	Palo Verde
Issue date:	12/04/1995
From:	Thomas C NRC (Affiliation Not Assigned)
To:	Stewart W ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
Shared Package
ML17311B304	List: ML17311B303 ML17311B305 ML17311B307
References
NUDOCS 9512120368
Download: ML17311B303 (65)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSlON WASHINGTON, D.C. 2055&0001 December 4,

1995 Hr. William L. Stewart Executive Vice President, Nuclear Arizona Public Service Company Post Office Box 53999

Phoenix, AZ 85072-3999

SUBJECT:

TRIP REPORTS ASSESSMENT OF THE PALO VERDE NUCLEAR GENERATING STATION GRADED EQUALITY ASSURANCE PROGRAM

Dear Hr. Stewart:

The enclosed trip reports provide the results of the August 29-30,

1995, NRC on-site discussion of the Palo Verde Nuclear Generating Station (PVNGS) risk-ranking methodology and expert panel process and the September 6-7,

1995, assessment of the graded quality assurance (gA) program applied to commercial grade item (CGI) dedication process.

The NRC and PVNGS staffs discussed aspects of the PVNGS risk-ranking methodology and expert panel process.

These discussions helped the NRC staff better understand the structure and rigor of the PVNGS process for considering information and evaluating risk significance.

The risk-ranking process appeared to have identified many of the high-risk-significant systems.

However, the NRC staff identified several issues that you should consider in applying your risk-ranking methodology.

Many of these risk-ranking issues were discussed in June 15,

1994, and October 14, 1994, letters from James L.

Milhoan of the NRC to Rasin and Tipton, respectively, of the Nuclear Energy Institute (NEI).

Enclosed with the October 14, 1994, letter was detailed guidance concerning the functions of the expert panel.

If incorporated into the PVNGS process, this. guidance would address some of these risk-ranking issues.

The NRC staff concluded that PVNGS has made a significant effort to establish a graded gA process for procurement and commercial grade item (CGI) dedication activities.

However, based on a review of PVNGS's gA program and implementing procedures for grading the procurement process and our evaluation of several low-risk-significant CGI dedication

packages, we have concluded that you should enhance the following areas of your graded gA program:

(1) procedural guidance for performing low-risk-significant procurement and CGI dedications, (2) use of post-installation testing in the CGI dedication

process, (3) verification of seismic qualification, and (4) corrective action feedback.

During the assessment, the NRC staff discussed expectations contained in the June 15,

1995, NRC (Hilhoan) letter to NEI (Rasin) and the need to reassess gA controls and safety significance based on information obtained from operational experience.

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95i2i20368 95i204 PDR ADQCK 05000528 P

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William L. Stewart Sincerely,

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In conclusion, the NRC and PVNGS staffs agreed that further interaction would be scheduled to continue the review of PVNGS's progress in implementing the risk-ranking methodology and graded QA requirements for low-risk-significant procurement and dedication of CGIs.

Dockets Nos.

STN 50-528, STN 50-529, and STN 50-530

Enclosures:

As stated cc w/encls:

See next page DOCUMENT NAME:

P:iPVGNS.LR2 Original signed by:

C.

R.

Thomas Charles R. Thomas, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Regulation DISTRIBUTION:

BBoger RGardner, RIII

JBlake, RII LSpessard JPetrosino RCorreia PGwynn GCwalina Central Files/PDR

MModes, RI WAng, RIV EJButcher RGramm JPeralta WHaass MMalloy RWoods HQMB/DRCH R/F BReckley

WDean, OEDO THiltz LCampbell RLatta BHolian RZimmermann CSerpan DOCUMENT NAME:

PVGNS.LR2 SEE PREVIOUS CONCURRENCE To receive a copy of this document, indicate in the bord "C" ~ Copy without enclosures "E" ~ Copy with enclosures "N" ~ No copy OFFICE HQMB ORCH DD DRSS DRCH NRR NRR ADT PDIV-2 PM NAME SBlack EJButcher BBo er 'Thadani, CThomas DATE 10/31/95*

10/14/95*

10/15/95*

OFFICIAL RECORD COPY

'll/22/95*

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William L. Stewart December 4,

1995 In conclusion, the NRC and PVNGS staffs agreed that further interaction would be scheduled to continue the review of PVNGS's progress in implementing the risk-ranking methodology and graded QA requirements for low-risk-significant procurement and dedication of CGIs."

Sincerely, Dockets Nos.

STN 50-528, STN 50-529, and STN 50-530

Enclosures:

As stated cc w/encls:

See next page DOCUMENT NAME:

P:iPVGNS.LR2 Original signed by:

C.

R.

Thomas Charles R. Thomas, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Regulation DISTRIBUTION

'Boger RGardner, RIII

JBlake, RII LSpessard JPetrosino RCorreia PGwynn GCwalina

~Central-Files/PDR

HModes, RI WAng, RIV EJButcher RGramm JPeralta WHaass HHalloy RWoods HQMB/DRCH R/F BReckley

WDean, OEDO THiltz LCampbell RLatta BHolian RZimmermann CSerpan DOCUMENT NAME:

PVGNS.LR2 SEE PREVIOUS CONCURRENCE To receive a copy of this document, indicate in the box: "C" Copy without enclosures "E" ~ Copy with enclosures "N" ~ No copy OFFICE HQMB DRCH DD DRSS ORCH NRR NRR ADT PDIV-2 PH NAME SBlack EJButcher BBo er AThadani CThomas DATE 10/31/95*

10/14/95*

10/15/95*

OFFICIAL RECORD COPY ll/22/95*

12/

/95

,3 William L. Stewart December 4,

1995 cc w/encls:

Mr. Steve Olea Arizona Corporation Commission 1200 W. Washington Street

Phoenix, Arizona 85007 T.

E. Oubre, Esq.

Southern California Edison Company P. 0.

Box 800

Rosemead, California 91770 Senior Resident Inspector USNRC P. 0.

Box 40 Buckeye, Arizona 85326 Regional Administrator, Region IV U. S. Nuclear Regulatory Commission Harris Tower 8 Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064

Chairman, Board of Supervisors ATTN:

Chairman 301 W. Jefferson, 10th Floor Phoenix, Arizona 85003 Mr. Aubrey V. Godwin, Director Arizona Radiation Regulatory Agency 4814 South 40 Street

Phoenix, Arizona 85040 Mr. Curtis Hoskins Executive Vice President and Chief Operating Officer Palo Verde Services 2025 N. 3rd Street, Suite 200

Phoenix, Arizona 85004 Roy P.

Lessey, Jr.,

Esq.

Akin, Gump, Strauss, Hauer'nd Feld El Paso Electric Company 1333 New Hampshire Avenue, Suite 400 Washington, DC 20036 Ms. Angela K. Krainik, Manager Nuclear Licensing Arizona Public Service Company P.O.

Box 52034

Phoenix, Arizona 85072-2034 Mr. R. Rehkugler Nuclear Assurance Houston Lighting and Power Company P. 0.

Box 289 Wadsworth, Texas 77483 Mr. M. J. Meisner, Director Nuclear Safety and Regulatory Affairs Entergy Operations, Inc.

P. 0.

Box 756 Port Gibson, Mississippi 39150

PVNGS Plant Overview Ul C) tA QO Auaust 25. 1995 TP 1 nf7

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Plant Description Plant Description

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Design/Construction NSSS: Combustion Engineering PWR Turbine-Generator Vendor: General Electric Architectural Engineer: Bechtel Power Corporation

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Unit Specifics Three identical units, producing 1,250 MWe net per unit.

Commercial operations declared on 1/86, 9/86, and 1/88.

August 25. 1995 TP 2 of7

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Contairunent Description

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Design Large, dry PWR type Net free volume: 2.6E6 cu-ft

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Power Operations Parameters Maintained between -0.3 and 2.5 psig Maintained below 120 F

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Design Basis Conditions Maintained below design pressure of 60 psig Leak rate not to exceed 0.1% by weight of free volume per day at an internal pressure of 60 psig

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Containment Penetrations Equipment hatch Personnel air-lock Emergency escape air-lock Fuel transfer tube Piping and Electrical penetrations

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Plant Systems Primary Plant Systems

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Reactor Coolant System 7wo independent loops, including two RCPs and one SG per loop T-hot: 616'F, T-cold: 560'F (at 2,250 psia)

Four ASME Code Pressurizer Safety Valves No PORVs Engineered Safety Features (AllStand-by)

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Safety Injection Two independent 100% capacity trains ofHPSI (1900 psig shutoff head),

Two independent 100% capacity trains ofLPSI Shutdown Cooling Heat Exchangers for long-term subcooled DHR Borated water supplied to SI pumps from Refueling Water Tank, 750,000 gal.

Four Safety Injection Tanks (SITS) containing >2300 ppm borated water Long term, post-accident core cooling provided by. recirculating borated water from containment sump AuLrnst 25. 1995 TP 4 nf 7

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Containment Cooling 7wo independent 100% capacity trains of Containment Spray provide containment heat removal and long-term core heat removal for Medium and Large LQCAs

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Support Systems 7wo redundant 100% capacity Essential. Chilled Water trains 7wo redunda'nt 100% capacity Essential Cooling Water trains 7wo redundant 100% capacity Essential Spray Ponds 7wo redundant 100% capacity Emergency Diesel-Generators and electrical distribution Other Important Plant Systems

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Station Blackout Gas-Turbine Generators Aum>st 25. 1995 Tp 5 of7

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Secondary Plant Systems

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AuxiliaryFeedwater System Three 100% capacity trains Two Safety-related Trains:

Train A: Class-powered, Seismic Cat I, turbine-driven pump Train B: Class-powered, Seismic Cat I, motor-driven pump One non-Safety-related Train:

Train N: Class-powered, non-Seismic Cat I, motor-driven pump Trains A and N are powered from Train Aof Class electrical distribution Train B is powered from Train B of Class electrical distribution Water source provided by 300,000 gallon dedicated reserve in Condensate Storage Tank

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Alternate Feedwater System Backup feed capability provided via Condensate pump(s)

(Requires off-site power)

Auuust 25. 1995 TP 6 nf7

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Secondary System Depressurixation Four main steam lines (2 per SG)

Four Atmospheric Dump Valves (1 per main steam line)

Remote-manual and Local-manual, control capability Twenty Main Steam Safety Valves (5 per main steam line)

Eight Turbine Bypass Valves downstream ofMSIVs Six to Main Condenser Toro to Atmosphere Automatic and Remote-manual control capability Either TBVs or ADVs operability required to bring plant to SDC entry conditions August 25. 199S TP 7 nf7

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IPE Insights and Response to Results at Palo Verde Nuclear Generating Station Aum)st 25. 1995 TP I nf7

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IPE INSIGHTS Two scenarios were identified as being significant:

Loss of Channel A DC Power Distribution Panel due to blown fuse Loss of HVACto Train A DC Equipmeni Rooms due to damper failures Together they accounted for over 70% of the calculated Core Damage Frequency of 1.0E-3 per reactor year.

IPE showed how certain inter-system dependencies could lead to unanticipated transient effects with multiple failures of mitigating equipment.

August 25. 1995 TP 2 of7

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LOSS OF CHANNELA DC POWER gg~

TRANSIENTEFFECTS Main Steam Isolation Valves close resulting in 100% load rejection, Feedwater Isolation Valves close resulting in additional plant heat-up.

Pressurizer safety relief valves challenged; Approximately a 2% chance one safety valve willnot re-close resulting in an induced Small LOCA (need for HPSP ITI ATI AP 8 IT KF ECT Main Feedwater lost due to Main Steam and Feedwater Isolation, Start-up AFW pump and Alternate FW blocked by fail-closed isolation valves, Two of three auxiliary feedwater pumps are without control power, Train A KCCS unavailable due to loss ofcontrol power.

Aumtst 25. 1995 Tp 3 of7

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LOSS OF HVACto TRAINA fi::;: ,

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DC EQUIPMENT ROOMS TRANSIENTEFFECTS Channel A and C Rooms heat up to equipment failure temperature.

Two channels ofVitalAC lost as inverters fail resulting in Reactor Trip and all ESF actuaiions; Battery chargers fail.

Batteries last about two hours.

Subsequent effects similar to Loss ofDC Power.

MITIGATIN ABILITYEFFE T

Flow from Main Feedwater and one Aux Feedwater Pump blocked by Main Steam Isolation Signal.

MSIS can be overridden, but when batteries are depleted, towpath blocked again.

Train A Essential Chiller unavailable - requires Channel AVitalAC Impacts reliabilityof ECCS pumps and Control Room HVAC Aum>st 25. 1995 Tp 4 nf 7

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SUMMARY

OF COMPENSATORY MEA URES Developed procedure for local manual operation of steam-driven auxiliary feedwater pump without DC control power. Performed test during a unit shutdown.

Revised Functional Recovery Procedure to include steps to recover towpaths and auxiliary feedwater pumps after loss ofcontrol power.

Modified Simulator model to accommodate Loss of DC scenario.

Job Performance Measures implemented for AuxiliaryOperators.

Periodic monitoring ofDC Equipment Room temperatures.

Staged temporary ventilating equipment for DC Equipment Rooms.

Reviewed Preventive Maintenance practices; revised Safety System Unavailability targets.

Auaust 25. 199S TP 5 nf7

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DESIGN MODIFICATIONS

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~W'a'a.'Y Installed temperature detectors in all four DC Equipment Rooms with high alarm in Control Room.

Greatly reduces the likelihood ofthe initiating event.

Shift power supply for Main Steam and Feedwater Isolation Valve Logic Cabinets away from supply for ESF mitigating equipment.

Greatly reduces likelihood ofa single failure causing both a severe transient and loss ofone Train ofmitigating equipment Changed SG downcomer isolation valves to fail open on loss of control power.

Maintains a flospath for three means offeeding the steam generators.

Install back-up power supply for control power to the non-essential auxiliary feedwater pump.

Improves reliabilityofthis pump.

Aumist 25. 1995 TP 6 of 7

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CORE DAMAGE;t."",;":.",+@,4.

FREQUENCY IMPROVEMEN

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g. 4v CDP UCT PRELIMINARY 1.03E-3/Rx-yr WITHCOMPENSATORY MEASURES 3.81E-4 63%

WITHPLANTMODIFICATIONS and minor model refinements 8.96E-5 91%

Cost of modifications was approximately $330,000 per unit Resolution of identified vulnerabilities was consistent with NUIUIARC Severe Accident Closure Guidelines Auauxt 25. 1995 TP 7 nf7

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AuxiliaryFeedwater System 5/a%3 m/A1tm SGN-VII2 SGN4%1H3 SGA-UVL72 SGB-UV13I SGg.y442 i SGg.y452 SG-1 Cl'AWVII4 CKL-HVII1 AFN.VII1 V~> g'4 APN-Pll

AFN-V133 bf.

>c icr APN VI12 APN VI13 SGN44V1143 nV/A1tnV SGN-VIIS SGA UV175 SGMW135 SGg.y453 Q SGg.y453 SG-2 Condensate Stonge Tank CTE-T01 t--, i~:i+~~i,i~ "'<&-AUXShll APA-V096 AFA-V055 SGA-VI43 SGA~134 APA.HV54 APA-VII2 SGN.HV1145 SG-1 SG-2 Tvrbiae Driver APA-KIl SGA-VI44 SGA~13S Atmosphere AFA.HVI32 APC-UVI34 l APA-VI79 Cfh-VI15 APA-VIII APA-VII7 AFA-PIl AFA VI15 AFA VI14 AFC-HVI33 APAQJVI37 AFA-VPl7

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"~~""::8'- EMWT g5'RE V005 APA VPSS AFB-HVI3I AFB-UVI34 Cta-VI14 APB-VI21 AFB.VI22 P

AFB-F41 AFB.V13S: AFB-VI24 AFB-VI25 AFB-HV431 AFBWVI35 AFB-VOÃ

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e ec e usiness rocesses or niia m emenaion Risk Analysis Managed configuration changes governing maintenance, surveillance, modifications, and outages Risk Based Technical Specifications Maintenance Prioritization (incl 10CF R50.65)

Operations Experience Assessment and Feedback Risk Based Decision Making and Submittals ISI/IST Graded QA in Procurement Business Risk Based Decision Making (economic and public risk)

Others (MOV, SOV, etc.)

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Sept. 30, 1995-Plan Developed Dec. 31,

'I 996 (tentative)

Plan Implemented 1997 and Beyond - Plan Maintenance, Improvements, and Expansion

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PRA Model Documentation Task Prioritization

~ Based on risk significance developed for Maintenance Rule.

Phase 1

~ General Information-Initiating Events, Event Trees, Success Criteria, etc.

~ High-Risk Systems-Safety Actuation, Safety Injection, Auxiliary Feedwater, etc.

Phase 2

~ Low-Risk Systems-HVAC, Electrical Support Systems, Gas Turbines, Instrument Air, etc.

Anm)st 29. 1995 TP 1 nf 2

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e PRA Documentation Project Schedule

~ General information Current Status: 16% complete Scheduled Completion: 12/08/85

~ High-Risk Systems Current Status: 10% complete Scheduled Completion: 12/22/95

~ Low-Risk Systems Current Status: 0% complete Scheduled Completion: - 02/23/95 Aumist 29. l995 TP 2 nf 2

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PRA Model Documentation Task Prioritization

~ Based on risk significance developed for Maintenance Rule.

Phase 1

~ General lnformation-Initiating Events, Event Trees, Success Criteria, etc.

~ High-Risk Systems-Safety Actuation, Safety Injection, Auxiliary Feedwater, etc.

Phase 2

~ Low-Risk Systems-HVAC, Electrical Support Systems, Gas Turbines, Instrument Air, etc.

August 29. 1995 TP I nf2

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PRA Documentation Project Schedule

~ General information Current Status:

1 6% complete Scheduled Completion: 12/08/85

~ High-Risk Systems Current Status: 10% complete Scheduled Completion: 12/22/95

~ Low-Risk Systems Current Status: 0% complete Scheduled Completion: '02/23/95 Auaust 29. 1995 TP 2 nf2

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Overview of the 1994 PRA. Model Presented By: L. M. Bullington August 29. 1995 TP 1 nf 3

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Core Damage Frequency

~ Unrecovered - 1.22E-4/yr

~ Recovered - 4.74E-5/yr Aum>st 29. 1995 TP 2 nf 3

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Other Insights

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Human error impacts over 34% of the total CDF (Most Important Actions Based upon Risk Increase)

Fail to Recover Control Room HVAC Pail to AlignAlternate Feedwater (APN pump or Alternate feedwater)

Fail to AlignHot Leg Injection

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CDF Contribution from System Unavailability due to corrective maintenance is 12%

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23% oftotal CDF has applied recoveries August 29. 1995 TP 3 of 3