ML17299B141

Forwards Response to EA Licitra 860109 Request for Addl Info Re Generic Ltr 83-28,Items 2.1,2.2.1.1,2.2.2 & 4.5.3

ML17299B141
Person / Time
Site:	Palo Verde
Issue date:	04/03/1986
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Knighton G Office of Nuclear Reactor Regulation
References
ANPP-35938-EEVB, GL-83-28, TAC-59163, TAC-59164, TAC-59173, TAC-61838, TAC-61839, TAC-61840, NUDOCS 8604080057
Download: ML17299B141 (26)

Text

REGULATOR NFORMATION DISTRIBUTION EM (RIDS)

ACCESSION NBR: 8604080057 DOC. DATE: 86/04/03 NOTARIZED:

NO FACIL:STN-50-528 Palo Verde Nuclear Stationi Unit ii Arizona Publi STN-50-529 Palo Verde Nuclear Stationi Unit 2i Arizona Publi STN-50-530 Palo Verde Nuclear Station>

Unit 3i'rizona Publi AUTH. NAME AUTHOR AFFILIATION VAN BRUNT> E'. E.

Arizona Nuclear Power Prospect (formerly Arizona Public Serv REC IP. NAME RECIPIENT AFFILI ATION KNIGHTONiG. W.

PWR ProJect Directorate 7

SUBJECT:

Forwards response to EA Licitra 860109 request for addi info re Generic Ltr 83-28'tems

2. 1i 2. 2. I. 1.2. 2. 2 8c 4. 5. 3.

DISTRIBUTION CODE:

A055D COPIES RECEIVED'TR ENCL 'IZE:

TITLE: OR/Licensing Submittal:

Salem ATWS Events GL-83-28 NOTES: Standardized plant.

Standardi zed plant.

Standardized plant.

05000528 05000529 05000530 RECIPIENT ID CODE/NAME PWR-B ADTS PWR-B PEICSB PWR-B PD7 PD 01 PWR-B PE ICSB INTERNAL: ACRS ELD/HDS3 IE/DGAVT NRR LASHER' NRR PWR-B ADTS NRR/TAMB RGN5 COPIES LTTR ENCL 1

1 2

2 3

3 1

1 6

6 1

0 1

1 1

1 1

1 1

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1 RECIPIENT ID CODE/NAME PWR-B EB PWR'-B FOB LICITRAe E PWR-B RSB

• DM/LFMB IE/DI NRR BWR ADTS NRR PWR-A ADTS NRRJJ) 0/RSIB EG ILE 04 COPIES LTTR ENCL 1.

1 1

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1 1

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1 1

1 1

1 1

1 1

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EXTERNAL: 24X NRC PDR 1

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TOTAL NUMBER OF COPIES REQUIRED:

LTTR 33 ENCL 31

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Arizona Nuclear Power Project P.o. BOX 52034 e

PHOENIX, ARIZONA85072-2034 April 3, 1986 ANPP-35938-EEVB/KLM/98.05 Director of Nuclear Reactor Regulation Attention:

Mr. George W. Knighton, Project Director PWR Project Directorate II7 Division of Pressurized Water Reactor Licensing B

U.ST Nuclear Regulatory Commission Washington, D.C.

20555

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2 and 3

Docket Nos.

STN 50-528 (License NPF-41)

STN 50-529 (License NPF-46)

STN 50-530 Additional Information on Generic Letter 83-28 Items 2. 1. 2.2. 1. 1, 2.2.2, and 4 ' '

File: 86-056-026;Gal.01.

10

References:

1) 2)

Letter from E.

A. Licitra, NRC, to E.

E.

Van Brunt, Jr.,

ANPP, dated January 9,

1986;

Subject:

Request for Additional Information Palo Verde Responses to Items 2. 1,

2. 2. l. 1, 2.2.2 and 4'.3, in Generic Letter 83-28.

Letter to G ~

W. Knighton,

NRC, from E.

E.

Van Brunt, Jr.,

ANPP, dated February 28, 1986 (ANPP-35348);

Subject:

Request for Additional Information on Generic Letter 83-28 Submittal Schedule.

Dear Mr. Knighton:

Attached are the responses requested by Reference 1).

If you should have any questions concerning this matter, please contact Mr.

W.

F. (}uinn of my staff.

Very truly yours, EEVB/KLM/rw Attachments E.

E.

Van Brunt, Jr.

Executive Vice President Project Director cc:

E.

A. Licitra (all w/a)

R.

P.

Zimmerman A.

C. Gehr 8goyOSO057 FOR ADQCK 05 P

PVNGS RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION GL 83-28, ITEMS 2.1, 2.2.1.1, 2.2.2 and 4.5.3 I

Information Re uest for Item 2.1 ( art 2)

Submit detailed information describing your vendor interface program for reactor trip system components.

Information supplied should 'tate how the program assures that vendor technical information is kept

complete, current and controlled throughout the life of the plant and should also indicate how the program will be implemented at PALO VERDE 1, 2

& 3.

~Res onse "PVNGS-ANPP Technical/Instruction Manual Control and Distribution" has been changed to ensure that revisions and/or addenda to technical manuals are reviewed by the responsible engineering group for impact on

existing, effective station procedures

and, where necessary, changes to those procedures implemented as required.

This is accomplished as follows:

Changes to technical manuals are distributed to the responsible engineering group for review.

The responsible engineer reviews the revision and makes comments on the accompanying "Document Review Control Form".

The engineer has the responsibility to ensure that any procedures impacted by the revision are updated to reflect that revision.

In

parallel, the revision and the "Document Review Control Form" are returned to PVNGS Drawing and Document Control for inclusion in station documents and permanent retention, as applicable (the exact process is contained in the station procedure).

It should be noted that this process is applicable to all station technical manuals, not gust for Reactor Trip System Components.

This procedure contains all the requirements for controlling vendor technical information throughout the life of the plant, as well as the requirements for permanent plant retention.

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Information Re uest for Item 2.2.1.1 Discuss how your revised classification procedure complies with the position presented in Generic Letter 83-28.

~Res esse The determination as to whether a

component is safety-related is based on one or more of the following criteria:

(1)

Function or operational mode of the component Required to open and/or close or remain as is.

(2)

Design Code Examples are:

ASME Section III, ANSI B31.1, and IEEE.

(3)

Pressure Retaining Requirements a)

Appendix A (attached)

Provides guidance for determining quality group A (safety related) classification of pressure retaining components, and is in accordance with Regulatory Guide 1.26 and 10CFR50.

b)

Appendix B (attached)

Provides guidance for determining quality group B

(safety related) classification of pressure retaining components, and in accordance with Regulatory Guide 1.26.

c)

Appendix C (attached)

Provides guidance for determining quality group C

(safety related) classification of pressure retaining components, and is in accordance with Regulatory Guide 1.26.

(4)

Environmental Design Conditions Designed for conditions such as Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB).

(5)

Seismic or Structural Requirements Seismic Category 1, and is in accordance with Regulatory Guide 1.29.

(6)

References a)

To determine if a specific component is safety related, use:

1)

Mechanical P&ID's

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Information Re uest for Item 2.2.1.1 (Continued) 2)

Instrument Index 3)

Valve Designation List 4)

Equipment Index b)

Appendix D (attached)

Provides guidance with respect to determining Class IE equipment classification, and is in accordance with IEEE standards, Regulatory Guides 1.30 and 1.97.

c)

Appendix E (attached)

Lists safety related functions for various types of components.

1 (7)

Definition for Safety Related:.

Any structure,

system, component, or material that is necessary to ensure:

1)

The integrity of the reactor coolant boundary.

2)

The capability to shut down the reactor and maintain it in a safe condition.

3)

The capability to prevent or mitigate the consequences of an accident which could result in potential off-site exposures comparable to the guideline exposures of 10CFR Part 100.

Im lementation of the Criteria The determination of which components are quality related or non-quality

related, the determination of the quality classification for new equipment, and the evaluation to determine quality class changes for components are all engineering responsibilities.

They must be approved by the Nuclear Engineering Manager and concurred with by the Corporate QA Director.

The evaluations are done via Engineering Evaluation Requests (EER) to determine the quality class and via Equipment Change Evaluations (ECE) to evaluate classification changes.

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APPENDIX A QUALITY GROUP A This provides guidance with respect to determining Quality Group classification in accordance with Regulatory Guide 1.26 and lOCPR50.

The following system(s) are Quality Group A.

1.0 The system or component is part of the reactor coolant pressure boundary as defined in 10CFR 50.2(v) unless excepted per 10CPR 50.55a(c)(2).

2.0 10CPR 50.2(v)

"Reactor'oolant pressure boundary means all those pressure-containing components of pressurized water-cooled nuclear power

reactors, such as pressure

vessels, piping, pumps and valves, which are:

I 2.1 Part of the reactor coolant system, or 2.2 Connected to the reactor coolant

system, up to and including any and all of the following:

2.2.1 The outermost containment isolation valve in system piping which penetrates primary reactor containment.

2.2.2 The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary reactor containment.

2.2.3 The reactor coolant system safety and relief valves."

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APPENDIX B QUALITY GROUP B

This provides guidance with respect to determining Quality Group B

classifications in accordance with Regulatory Guide 1.26.

The following systems are classified as Quality Group B.

1.0 Water and steam containing pressure

vessels, heat exchangers, storage

tanks, piping,

pumps, and valves that are part of the Reactor Coolant system but are excluded from Quality Group A by footnote 2 of 10CFR50.55a.

2.0 The system*,

portions of systems or component is designed for emergency core cooling.

3.0 The system*,

portions of systems or component is designed for post-accident containment heat removal.

4.0 The system*,

portions of systems or component is designed for residual heat removal from the reactor.

5.0 The system*,

portions of systems or component is designed to provide reactor shutdown.

6.0 The system*,

portions of systems or component is designed for post-accident containment atmosphere cleanup.

7.0 The system*, portions of systems or component is designed for containment isolation.

8.0 Those portions of the steam and feedwater systems of pressurized water reactors extending from and including the secondary side of steam generators up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure during all modes or normal reactor operation.

NOTE:

The class break for R 6 S class instruments and associated tubing which moni,tor Quality Group B system shall be at a normally closed root valve or an excess flow check valve.

• The piping system boundary includes those portions of the system required to accomplish the specified safety function and connected piping up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure when the safety function is required.

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APPENDIX C QUALITY GROUP C

This provides guidance with respect to determining Quality Group C

classifications in accordance with Regulatory Guide 1.26.

The following systems are classified as Quality Group C.

1 ~ 0 Cooling water, component

cooling, and auxiliary feedwater systems*

or portions of systems that are required for (1) emergency core cooling, (2) post-accident containment heat

removal, (3) post-accident containment atmosphere

cleanup, (4) residual heat removal from the reactor or (5) cooling the spent fuel pool.

2 ~ 0 Cooling water or portions of systems that axe required for functioning of reactor coolant system components which are safety

related, such as reactor coolant pumps.

3.0

Systems, other than radioactive waste management

systems, that contain or may contain radioactive

material, and whose postulated failure would result in conservatively calculated potential offsite doses that are more than 0.5 rem to the whole body or its equivalent to any part of the body.

NOTE:

The class break for R&S classed instruments and associated tubing which monitor Quality Group C systems shall be at a normally closed root valve or an excess flow check valve.

• The piping system includes those portions of the piping system requixed to accomplish the specified function and connected piping up to and including the first valve (including a safety or relief valve) that is either normally closed or capable of automatic closure when the safety function is required.

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APPENDIX D CLASS IE EQUIPMENT This provides guidance with respect to determining Class IE equipment classification in accordance with IEEE Standards and Regulatory Guides 1.30 and 1.97.

Design Criteria Class IE power systems shall be designed to assure that no design basis event will cause:

A.

A loss of electric power to a

number of engineered safety

features, surveillance

devices, or protective systems devices such that a protective function cannot be performed.

B.

A loss of electric power to equipment that could result in a power transient capable of causing significant damage to the fuel or to the reactor coolant system.

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Protective Systems Class IE systems include the electrical and mechanical devices (from measured process variable to protective action system input terminals) involved in generating those systems associated with the protective functions.

These signals include those that initiate reactor trip, engineered safety features (for example, containment isolation and safety injection), and auxiliary supporting features.

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Items Included in Class IE Systems

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APPENDIX E SAFETY RELATED FUNCTIONS This attachment lists various possible safety related functions for typical components.

Com nent Functions Accumulator Air Dryer Annunciator Battery Blower Circuit Breaker Control Rod CEDM Demineralizer Electrical Conductor Engine Filter Pass Fluid Alarm Delivery Energy Deliver Head Open Control Withdraw Pass Fluid Pass Current Deliver Torque Pass Fluid Close

Insert, Scram Demineralize Block Particles Fuel Burn Generator Heater Heat Exchanger Ins trument Motor Pipe Deliver Energy Pass Fluid Pass Fluid Convert Signal Deliver Torque Pass Fluid Heat Separate Flow Pump Deliver Head

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APPENDIX E (Continued)

Com nent Functions Recombiner Relay Transformer Turbine Valve Valve Operator Ve ssel Recombine H2 Open Deliver Energy Deliver".Energy Pass Fluid Open Pass Fluid Close Insulate Block Fluid Flow Close

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Information Re uest for Item 2.2.2 The staff found the NUTAC program fails to address the concerns about establishing and maintaining an interface between all vendors of safety-related equipment and the utility.

Accordingly, supplement your response to address this concern.

This additional information should describe how current procedures will be modified and new ones initiated to meet the elements of this concern.

~Res ense ANPP has assigned responsibility for Operating Experience Information (OEI) management to the Independent Safety Engineering Group (ISEG).

Their corporate procedure identifies one source of OEI as vendor reports; including vendor technical bulletins and notices identifying problems and recommended corrective actions associated with vendor equipment, systems and services.

This ensures an active interface is maintained with all vendors of safety-related equipment.

The ISEG responsible engineer is required by procedure to forward the OEI to affected departments and to track each item to ensure proper resolution and close out.

This ensures that equipment technical information provided by vendors of safety-related equipment is incorporated into plant instructions and procedures.

ANPP is currently reviewing the process by which vendor information received by ANPP is distributed to ISEG.

After completion, ANPP will incorporate the results of the review into a more formalized process as appropriate.

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Information Re uest for Item 4.5.3 Provide the results of your review of the on-line functional testing intervals considering the concerns of 4.5.3.1 to 4.5.3.5 in

the, generic letter.

Proposed Technical Specification changes resulting from this review shall be submitted for review.

~Res oese A fault tree model for the postulated fault, "failure to trip the reactor" was constructed for PVNGS Units 1, 2

and 3.

This model explicitly addressed concerns 4.5.3.1 to 4.5.3.4 including, but not limited to, items such as random component

failures, operator errors and out-of-service time for testing.

Component failure rates were quantified using applicable operating experience data to perform a

Bayesian update of WASH 1400 failure rate distributions.

Common cause failure rates were quantified using operating experience data and the Vesely specialization of the Marshall-Olken algorithm.

The fault tree models were quantitatively evaluated using Monte Carlo simulation to derive a

system unavailability distribution.

A sensitivity analysis was also performed to determine how sensitive the system unavailability was to variations in the failure rates of individual components.

When Generic Letter 83-28 was

issued, the failure mode of the RTBs was partially attributed to breaker component wear.

During the course of the GEOG program, to investigate various options for enhancements of existing breakers, it became apparent that the primary mode of failure of the breakers was age related hardening of the lubricant in the breaker trip shaft bearings and latch roller assembly.

It was also determined that the breakers do not contain components subject to detrimental frictional wear; thus, the potential for breaker failure due to component wear is minimal.

Therefore, concern 4.5.3.5, component "wear-out" caused by testing, was not considered in this evaluation.

The results of this analysis show that the median probability that PVNGS Units 1,

2 and 3

RPS will fail to trip the reactor is 4.08 x 10 6 per demand with a

95th percentile confidence limit probability of 2.06 x

10 5 per demand.

This compares favorably to the NRC derived point estimate value of 2 x 10 5

per demand as the probability that the RPS would fail to trip the reactor for plants with a C-E supplied NSSS.

Based on this, it is concluded that the current RPS test intervals are consistent with maintaining the high degree of availability expected of the RPS.

Therefore, no proposed Technical Specification changes have resulted from this review.

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