Text

Requests Assistance in Reviewing Encl Util 841102,850505 & 0919 Applications to Amend Licenses NPF-4 & NPF-7 Re Organization Changes,Entries to Airlock for Door Repair & Visual Insp of Snubbers,Respectively
	ML20133G206
Person / Time
Site:	North Anna
Issue date:	10/04/1985
From:	Thompson H; Office of Nuclear Reactor Regulation
To:	Walker R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
	TAC-56347, TAC-56350, TAC-59762, TAC-59763, TAC-59764, TAC-59765, NUDOCS 8510150249
	Download: ML20133G206 (2)
	v • d • e

( b d a October 4, 198s O[G Docket Nos. 50-338 and 50-339 MEMORANDUM FOR: Roger D. Walker, Director Division of Reactor Projects Region II FROM: Hugh L. Thompson, Jr., Director Division of Licensing Office of Nuclear Reactor Regulation

SUBJECT:

LICENSING ACTION REVIEWS FOR THE NORTH ANNA POWER STATION, UNITS N0. 1 AND N0. 2 (NA-182)

Your assistance is requested in conducting a review of the three enclosed submittals from the Virginia Electric and Power Company (VEPCO) for revising the NA-182 Technical Specifications (TS). The first submittal, dated November 2,1984 (Serial No. 578), would revise the TS to reflect current organizations within the Nuclear Operations Department, Quality Assurance Department, Maintenance and Performance Service Departments, and the Security Department. The second submittal, dated May 2, 1985 and September 19,1985 (Serial Nos.85-162 and 85-162 A), would revise the NA-182 TS by clarifying the action statement to permit entries into the air lock for repair of an inoperable inner air lock door. The third submittal, dated September 24, 1985 (Serial No.85-686), would revise the NA-182 TS Surveillance Requirements for Snubbers regarding visual inspection acceptance criteria and would establish functional test methods for large bore snubbers as specified in VEPCO's Interim Program as reviewed and approved by Region II (IR 50-338/83-29 and 50-339/83-29).

The products expected from you as a result of your reviews are completed safety evaluations to support VEPC0's proposed amendment requests. Your

, reviewers should use the Standard Review Plan (SRP) and Standard TS, as

, applicable, for guidance in determining acceptance criteria (recognizing,

! of course, that for operating reactors, the criteria in these documents are not requirements). It is requested that the safety evaluations be l written in the third person and that it be transmitted over the 5520 net-

! work to NRC DL, Lead OP, attention L. Engle. This is subject to change j and should be verified by your reviewer just prior to transmittal.

l In accordance with NRR Office Letter No. 44, each safety evaluation performed

by a technical division shall have separate SALP input provided. For purposes i of these reviews, the Regional personnel involved are considered part of the g'o

! technical divisions. Therefore, we are requesting that your forwarding memo-randum contain a SALP input for each safety evaluation performed.

i Work for the VEPC0 submittals concerning the proposed changes have been dis-cussed with S. Elrod and D. Gruber of your staff. The TAC numbers for the 8510150249 851004 l

i fDR ADOCM 05000338 PDR

J submittal dated November 2, 1984 are 56347 and 56350 and the requested com-pletion date is November 30, 1985. The TAC numbers for the submittal dated May 2 and September 19,1985 are 59762 and 59763 and the requested completion date is also November 30, 1985. Finally, the TAC numbers for the submittal dated September 24,1985 are 59764 and 59765 and the requested completion date is October 25, 1985. As discussed with your staff, the requested com-pletion date of October 25, 1985 is necessary in order that VEPC0 can imple-ment the large bore snubber tests for NA-1 during the forthcoming refueling outage scheduled to conrnence November 1,1985.

Please notify me as soon as possible if the completion dates are not accept-able. Any contact with the licensee concerning these reviews or any additional information deemed necessary should be obtained through the NRR Project Manager for NA-1&2, Leon B. Engle, who can be reached at FTS 492-7094. During the course of these reviews, you should encourage the reviewers to work closely with the NA Project Manager.

Frank Mir2511"If #

Hugh L. Thompson, Jr., Director Division of Licensing Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures K. Landis, Region II S. Elrod, Region II D. Gruber, Region II DISTRIBUTION:

DOCKET FILE ORB #3 Rdg NRC PDR L PDR PMXreutzer LBEngle EJButcher GClainas HThompson r

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" VINGINIA ELECTRIC AND Powen COMPANY 2

Ricnwoxu,VinotxiA const V.L.SrswAar Yaca Persinest November 2, 1984 xtcas . o,..uion, 4

Mr. Harold R. Denton, Director Serial No. 578 i

Office of Nuclear Reactor Regulation N0/JHL:acm Attn: Mr. James R. Miller, Chief Docket Nos. 50-338 Operating Reactors Branch No. 3 50-339 Division of Licensing License Nos. NPF-4 U. S. Nuclear Regulatory Commission NPF-7 k'ashington, D. C. 20555

Gentlemen

4 VIRGINIA ELECTRIC AND P0k'ER COMPANY NORTH ANNA P0k'ER STATION UNIT NOS. 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGES Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company requests an j

amendment, in the form of changes to the Technical Specifications, to Operating Licenses NPF-4 and NPF-7 for North Anna Unit Nos. I and 2.

The proposed Technical Specification changes reflect the current organizations within the Nuclear Operations Department. Quality Assurance Department, Maintenance and Performance Services Department and Security Department. In

' addition, the changes reflect 1) the title change of the Executive Vice President - Power to the Executive Vice President and Chief Operating Officer,

2) the requirement to retain the records for at least five years ' when the Station Emergency Plan and implementing procedures and the Station Security Plan and implementing procedures are annually audited, and 3) the reference change of the ANSI standard on Facility Staff Qualification and Training.

i In order to meet the requirements of ANS 3.1-(12/79 Draft) we are planning a

" phased" verification of the specific qualification requirements of- the standard. The task analysis required by paragraph 5.3.1 of ANS 3.1-(12/79

, Draft) will be performed. Individual background and experience will be.

considered to assure that the intent of the standard is met.

From an organizational point of view, the proposed changes will provide assurance that the plant is operated safely and all issues which have potential for impacting the continued safe operation of the plant will be

reviewed by the individuals best qualified to render judgment and advise the 1

Station Manager.

)

The proposed Technical Specification changes for North Anna Unit No. I are provided in Attachment 1.

The proposed Technical ' Specification changes for North Anna Unit No. 2 are provided in Attachment 2. A discussion of the proposed Technical Specification changes is provided in Attachment 3.

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i vanotwta ELrcraic axo Porra Cowemr to Harold R. Denton 2

This request has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control staff;. It has been determined that this request does not pose an unreviewed safety question as defined in 10 CFR 50.59 and does not pose a significant hazards consideration as defined in 10 CFR 50.92.

We have reviewed this request in accordance with the criteria in 10 CFR 170.12. A check in the amount of $150 is enclosed as an application fee.

4 These proposed Technical Specification changes supercede our previous reorganization Technical Specification changes. Please cancel our proposed Technical Specification change requests dated June 24, 1982 (Serial No. 264),

July 1, 1983 (Serial No. 264A) and October 13, 1983 (Serial No. 264B).

Vr truly yogrs, r- t i

W. L. Stewart Attachments

1. Proposed Technical Specification Changes - Unit 1

2. Proposed Technical Specification Changes _ Unit 2

3. Discussion of Proposed Technical Specification Changes

4. Voucher Check for $150 cc: Mr. James P. O'Reilly Regional Administrator Region II

. , .Nr. M. W. Branch NRC Resident Inspector North Anna Power Station Mr. Charles Price ~~

Department of Health 109 Governor Street Richmond, Virginia 23219 I

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COMMONWEALTH OF VIRGINIA )

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CITY OF RICHMOND )

The foregoing document was acknowledged before me, in and for the City and Connonwealth aforesaid, today by W. L. Stewart who is Vice President Nuclear Operations, of the Virginia Electric and Power Company. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this 4 d day of N m ' - , 19 9'/ .

My Concission expires: A-M , 19 95' .

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ADMINISTRATIVE CONTROLS 6.2.3 SAFETY ENGINEERING STAFF (SES) ]

FUNCTION

! 6.2.3.1 The SES shall function to examine plant operating characteristics, i NRC issuances, industry advisories, Licensee Event Reports, and other sources which may indicate areas for improving plant safety.

COMPOSITION 6.2.3.2 The SES shall be composed of at least five dedicated, full-time l

engineers located onsite.

RESPONSIBILITIES 6.2.3.3 The SES shall be responsible for maintaining surveillance of plant activities to provide independent verification

that these activities are performed correctly and that human errors are reduced as much as practical.

I AUTHORITY 6.2.3.4 The SES shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving plant safety to the Assistant Station Manager (Nuclear Safety and Licensing).

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall serve in an advisory capacity to Shift Supervisor on matters pertaining to the engineering aspects of assuring i l .:. cafe operation of the unit.

6.2.4.2 The Shift Technical Advisor shall disseminate relevant operational experience identified by the SES.

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, 6.3 FACILITY STAFF QUALIFICATIONS ,

6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifi-4 cations of ANS 3.1-(12/79 Draft) for comparable positions and the supplemental l 1

requirements specified in the March 28, 1980 NRC letter to all licensees, except for (1) the Supervisor - Health Physics who shall meet or exceed the qualifications of Regulatorr Guide 1.8, September 1975 and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 The Station Manager is responsible for ensuring that retraining and replacement training programs for the facility staff are maintained and that such programs meet or exceed the requirements and recommendations of Section 5 of ANS 3.1-(12/79 Draf t) and Appendix "A" of 10 CFR Part 55 and the supple- l

mental requirements specified in the March 28, 1980 NRC letter to all

! licensees, and shall include familiarization with relevant industry operational experience identified by the SES.

6.5 REVIEW AND AUDIT 6.5.1 STATION NUCLEAR SAFETY AND OPERATING COMMITTEE (SNSOC)

FUNCTION

6.5.1.1 The SNSOC shall function to advise the Station Manager on all matters related to nuclear safety.

,' COMPOSITION 6.5.1.2 The SNSOC shall be composed of the:

~~

Chairman: Assistant Station Manager (Nuclear Safety and Licensing)

Vice Chairman: Assistant Station Manager (Operations and Maintenance)

, !! amber: Superintendent-Operations Member: Superintendent-Maintenance Member: Superintendent-Technical Services

Member: Supervisor-Health Physics

, ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the SNSOC Chairman to serve on a temporary basis; however, no more than one alternate shall participate as a voting member in SNSOC activities at any one time.

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NORTH ANNA - UNIT 1 6-5

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s ADMINISTRATIVE CONTROLS MEETING FREQUENCY , ,

6.5.1.4 The SNSOC shall meet at least once per calendar mont$ ind as convened by the SNSOC Chairman or his designated alternate.

QUORLH

.5.1.5 A quorum of the SNSOC consists of the Chairman or Vice-Chairman and two members including alternates.

RESPONSIBILITIES 6.5.1.6 The SNSOC shall be responsible for:

a. Review of 1) all procedures required by Specification 6.8.1, 6.8.2 and 6.8.3 and changes thereto, 2) all programs required by Specification 6.8.4 and changes thereto and 3) any other proposed procedures or changes thereto as determined by the Station Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

c. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

d. Review of all proposed changes to Appendix "A" Technical Specifications and Appendix "B" Environmental Protection Plan.

Recommended changes shall be submitted to the Station Manager.

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e. Investigation of all violations of the Technical Specifications

. . including the preparation and forwarding of reports covering a - ' -

evaluation and recommendations to prevent recurrence to the Vice President - Nuclear Operations and the Director-Safety Evaluation and Control.

~~

f. Review of events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the Commission.

g. Review of facility operations to detect potential nuclear safety hazards.

h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Station Nuclear Safety and Operating Committee or Station Manager.

1. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Station Manager.

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l j. Review of the Emergency Plan and implementing pro'cedures and shall

! submit recommended changes to the Station Manager.

NORTH ANNA - UNIT 1 6-6

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ADMINISTRATIVE CONTROLS

k. Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports cove'ing r evaluation, recommendations and disposition of the corrective act' ion to prevent recurrence and the forwarding of these reports to .the Vice

' President-Nuclear Operations and to the Director-Safety Evaluation and Control. ,

1. Review changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL.

2 AUTHORITY 1

1 6.5.1.7 The SNSOC shall:

a. Provide written approval or disapproval of items considered . under

] 6.5.1.6(a) through (c) above.

b. Render determinations in writing with regard to whether or not each g

item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

t' I c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President-l Nuclear Operations and the Director-Safety Evaluation and Control of disagreement between the SNSOC and the Station Manager; however, the ,

Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

RECORDS 1

i

) 6.5.1.8 The SNSOC shall maintain written minutes of each meeting and i 1-

,! copies shall be provided to the Station Manager. Vice President-Nuclear

-, Operations and the Director-Safety' Evaluation and Control.

6.5.2 SAFETY EVALUATION AND CONTROL (SEC) i FUNCTION i

6.5.2.1 SEC shall function to provide independent review of designated I activities in the areas of:

a. Nuclear power plant operations

b. Nuclear engineering l

c. Chemistry and radiochemistry

d. Metallurgy

e. Instrumentation and control

f. Radiological safety l Mechanical and electrical engineering 3

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h. Administrative controls and quality assurance practices ~

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! 1. Other appropriate fields associated with the unique characteristics of I the nuclear power plant

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l NORTH ANNA - UNIT 1 6-7

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s ADMINISTRATIVE CONTROLS 1

! d. Violations and reportable occurrences such as: .

1. Violationsofapplicablecodes, regulations,ordArs, Technical Specifications, license requirements or internal

procedures or instructions having safety significance;

2. Significant operating abnormalities or deviations from normal l or expected performance of station safety-related structures, j

systems, or components; and

3. Reportable occurrences as defined in the station Technical Specification 6.9.1.8.

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Review of events covered under this paragraph shall include the results of any investigations made and recommendations resulting f rom such investigations to prevent or reduce the probability of recurrence of the event.

e. The Quality Assurance Department audit program at least once per 12 months and audit reports.

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} f. Any other matter involving safe operation of the nuclear power j

stations which is referred to the Director-Safety Evaluation and 8 Control. l

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2 3 Reports and meeting minutes of the Station Nuclear Safety and Operating Committee.

i AUTHORITY

, 6.5,.2.9 The Director-Safety Evaluation and Control shall report to and advise j -, - -the ' Manager-Nuclear Programs and Licensing, who shall advise tL? Vice l

- President- Nuclear Operations on those areas of responsibility specified in

- Section 6.5.2.7.

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l RECORDS 6.5.2.10 Records of SEC activities required by Section 6.5.2.7 shall be 4

prepared and maintained in the SEC files and ~ a summary shall be disseminated l

as' indicated below each calendar month.

! 1. Vice President-Nuclear Operations

2. Nuclear Power Station Managers 4

3. Manager-Nuclear Operations Support

4. Manager-Nuclear Programs and Licensing '

! 5. Executive Manager-Quality Assurance

6. Others that the Director-Safety Evaluation and Control may designate.

i NORTH ANNA - UNIT 1 6-9 f

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s ADMINISTRATIVE CONTROLS

m. The PROCESS CONTROL PROGRAM and implementing procedures; for processing and packaging of radioactive wastes at least once per 24 months.

n. The performance of activities required by the Quality Assurance Program to meet the provisions of Regulatory Guide 1.21. Revision 1 June 1974, and Regulatory Guide 4.1, Revision 1, April 1975 at least once per 12 months.

AUTHORITY 6.5.3.2 The Quality Assurance Department shall report to and advise the Executive Manager-Quality Assurance, who shall advise the Senior Vice President-Power Operations on those areas of responsibility specified in Section 6.5.3.1.

RECORDS 2

6.5.3.3 Records of the Quality Assurance Department audits shall be prepared and maintained in the department files. Audit reports shall be disseminated as indicated below:

1. Vice President - Nuclear Operations

2. Nuclear Power Station Manager

3. Manager-Nuclear Operations Support

4. Manager-Nuclear Programs and Licensing

5. Executive Manager - Quality Assurance 3 e .

- 6. Director - Safety Evaluation and Control

7. Nuclear Power Station Manager Quality Assurance

8. Supervisor of area audited NORIN ANNA - UNIT 1 6-11

a ADMINISTRATIVE CONTROLS 6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission shall be reviewed by the SNSOC and submitted to the Director-Safety Evaluation and Control and the Vice President -

Nuclear Operations.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The facility shall be placed in at least HOT STANDBY vithin one hour.

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President-Nuclear Operations, and the Director-Safety Evaluation and Control shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SNSOC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

. d. The Safety Limit Violation Report shall be submitted to the Conunission, the Director-Safety Evaluation and Control and the Vice President-Nuclear Operations within 14 days of the violation. l 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978.

b. Refueling operations.

3 NORTH ANNA - UNIT 1 6-12

ADMINISTRATIVE CONTROLS

c. Surveillance and test activities of safety related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 1.21, Revision 1 June 1974 and Regulatory Guide 4.1, Revision 1, April 1975.

6.8.2 Each procedute of 6.8.1 above, except 6.8.1.d and 6.8.1.e and changes thereto, shall be reviewed and approved by the SNSOC prior to implementation and reviewed periodically as set forth in administrative procedures.

Procedures of 6.8.1.d and 6.8.1.e shall be reviewed and approved as per 6.5.1.6.1 and J.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

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c. The change is documented, reviewed, and approved by the SNSOC within

.. 14 days of implementation.

6.8.4 The following programs shall be established, implemented., and maintained: *

a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the recirculation spray, safety injection, chemical and volume control, gas stripper, and hydrogen recombiners. The program shall include the following:

(1) Preventive maintenance and periodic visual inspection require-j ments and

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(ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

NORTH ANNA - UNIT I 6-13

ADMINISTRATIVE CONTROLS

1. For any abnormal degradation of the structural integrity of the reactor vessel or the Reactor Coolant System pressure boundary detected during the performance of Specification 4.4.(0, an initial report shall be submitted within 10 days after detection and a detailed report submitted within 90 days after the completion of Specification 4.4.10.

J. For any abnormal degradation of the containment structure detected during the performance of Specification 4.6.1.6, an initial report shall be submitted within 10 days af ter completion of Specification 4.6.1.6. A final report, which includes (1) a description of the condition of the liner plate and concrete. (2) inspection procedure, (3) the tolerance on cracking and (4) the corrective actions taken, shall be submitted within 90 days after the completion of Specification 4.6.1.6.

k. Inoperable Fire Detection Instrumentation, Specification 3.3.3.7.

1. Inoperable Fire Suppression Systems, Specifications 3.7.14.1, 3.7.14.2, 3.7.14.3, 3.7.14.4 and 3.7.14.5.

6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

a. Records and logs of facility operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipme'nt related to nuclear safety.

. c. Each REPORTABLE OCCURRENCE submitted to the Consission.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of changes made to Operating Procedures,

f. Records of radioactive shipments.

g. Records of sealed source leak tests and results.

h. Records of annual physical inventory of all sealed source material of record.

i. Records of the annual audit of the Station Emergency Plan and implementing procedures. 1 l

j. Records of the annual audit of the Station Security Plan and l implementing procedures.

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

NORTH ANNA - UNIT 1 6-22

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Station Manager shall be responsible f or overall f acf;11ty operation.

In his absence, the Assistant Station Manager (Operations and Maintenance) l shall be responsible for overall f acility operation. During i:he absence of both, the Station Manager shall delegate in writing the succession to this responsibility.

i 6.1.2 The Shift Supervisor (or during his absence from the Control Raiom, a cesignated individual) shall be responsible for the Control Room command i function and shall be the only individual that may direct the licensed activ-ites of licensed operators. A management directive to this effect, signed by the Senior Vice President - Power Operations, shall be reissued to all station l personnel on an annual basis.

6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a. Each on duty shif t shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in MODES 1, 2, 3 or 4, at least one licensed Senior Reactor

Operator shall be in the Control Room.

c. A health physics technician # shall be on site when fuel is in the reactor. ,

d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

e. A site Fire Brigade of at least 5 members shall be maintained onsite at all timesf. The Fire Brigade shall not include the s

minimum shift crew shown in Table 6.2-1 or any personnel required for other essential functions during a fire emergency.

fIhe health physics tecnnician and Fire Brigade composition may be less than

' the minimum requirement for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order

' to accommodate unexpected absence provided inunediate action is taken to fill l/ the required positions.

l NORTH ANNA - UNIT 2 6-1 l

~ - -

ADMINISTRATIVE CONTROLS 6.2.3 SAFETY ENGINEERING STAFF (SES) ;

Ft'NCTION 6.2.3.1 The SES shall function to examine plant operating characteristics.

NRC issuances, industry advisories, Licensee Event Reports, and other sources which may indicate areas for improving plant safety.

COMPOSITION 6.2.3.2 The SES shall be composed of at least five dedicated, full-time engineers located onsite.

RESPONSIBILITIES 6.2.3.3 The SES shall be responsible for maintaining surveillance of plant activities to provide independent verification

that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.4 The SES shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving plant safety to the Assistant Station Manager (Nuclear Safety and Licensing).

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shift Technical Advisor shall serve in an advisory capabity ta.

Shift Supervisor on matters pertaining to the engineering aspects of assuring '

safe operation of the unit.

-- 6.'2.4.2 The Shift Technical Advisor shall disseminate relevant operational experience identified by the SES.

1 ll l

! 4 i ,

Not responsible for sign-off function.

NORTH ANNA - UNIT 2 6-la ,,

'I .s Figure 6.2-1 Offsite Orgianization.for Facility Management and Technical Support l

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ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or excee'd the minimum qualifications of ANS 3.1-(12/79 Draft) for comparable poritions and the supplemental requirements specified in the March 28, 1980 NRC letter to all l licensees, except for (1) the Supervisor - Health Physics who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering dis:1pline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 The Station Manager is responsible for ensuring that retraining and replacement training programs for the facility staff are maintained and that such programs meet or exceed the requirements and recommendations of Section 5 of ANS 3.1-(12/79 Draft) and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in the March 28, 1980 NRC letter to all licensees, and shall include far.iliarization with relevant industry operational experience identified by the SES.

6.5 REVIEW AND AUDIT 6.5.1 STATION NUCLEAR SAFETY AND OPERATING COMMITTEE (SNSOC)

FUNCTION 6.5.1.1 The SNSOC shall function to advise the Station Manager on all matters related to nuclear safety.

COMPOSITION i

[ .

6)5.1.2 The SNSOC shall be composed of the:

Chairman: Assistant Station Manager (Nuclear Safety and Licensing)

Vice Chairman, Assistant Station Manager (Operations and Maintenance) ,_

Member: Superintendent-Operations Member: Superintendent-Maintenance Member: Superintendent-Technical Services Member: Supervisor-Health Physics ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the SNSOC Chairman to serve on a temporary basis; however, no more than one alterna*;e shall participate as a voting member in SNSOC activities at any one time.

4 NORTH ANNA - UNIT 2 6-6 l

ADMINISTRATIVE CONTROLS i

MEETING FREQUENCY 6.5.1.4 TheSNSOCshallmeetatleastoncepercalendarmon(h-andasconvened by the SNSOC Chairman or his designated alternate.

t QUORUM i 6.5.1.5 A quorum t f the SNSOC consists of the Chairman or Vice-Chairman and two members including alternates.

RESPONSIBILITIES 6.5.1.6 The SNSOC shall be responsible for:

a. Review of 1) all procedures required by Specification 6.8.1, 6.8.2 and 6.8.3 and changes thereto, 2) all programs required by Specification 6.8.4 and changes thereto, 3) any other proposed procedures or changes thereto as determined by the Station Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety,

c. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

d. Review of all proposed changes to Appendix "A" Technical Specifications and Appendix "B" Environmental Protection Plan.

Recommended changes shall be submitted to the Station Manager.

= . .

l : -

- . e. Investigation of all violations of the Technical Specifications

~

including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Vice President-Nuclear Operations and the Director-Safety Evaluation l ~~

and Control.

l i f. Review of events requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months written notification to the l Comnission.

I

g. Review of facility operations to detect potential nuclear safety hazards.

! h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Station Nuclear l Safety and Operating Committee or Station Manager. l l 1. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Station Manager. l

j. Review of the Emergency Plan and taplementing procedures and shall submit recommended changes to the Station Manager.

l, l NORTH ANNA - UNIT 2 6-7 l ,

I l

I ADMINISTRATIVE CONTROLS

k. Review of every unplanned onsite release of radioactive material to the environs including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these feports to the Vice President-Nuclear Operations and to the Director-Safety Evaluation and Control.

1. Review changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE (

CALCULATION MANUAL.

AUTHORITY 6.5.1.7 The SNSOC shall:

a. Provide written approval or disapproval of items considered under 6.5.1.6(a) through (c) above.

b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

c. Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Vice President -

Nuclear Operations and the Director-Safety Evaluation and Control of disagreement between the SNSOC and the Station Manager; however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

RECORDS 6.5.1.8 The SNSOC shall maintain written minutes of each meeting and copies shall be provided to the Station Manager, Vice President - Nuclear Operations, l and the Director-Safety Evaluation and Control.

=.

~'

[ 1 6. 5.'2 SAFETY EVALUATION AND CONTROL (SEC)

FUNCTION 6.5.2.1 SEC shall function to provide independent review of designated, activities in the areas of:

a. Nuclear power plant operations

b. Nuclear engineering

c. Chemistry and radiochemistry

d. Metallurgy

e. Instrumentation and control

f. Radiological safety

g. Mechanical and electrical engineering l h. Administrative controls and quality assurance practices

1. Other appropriate fields associated with the unique characteristics of the nuclear power plant l

NORTH ANNA - UNIT 2 6-8

ADMINISTRATIVE CONTROLS l

REVIEW (Cont'd) !

d. Violations and reportable occurrences such as: }

1. Violations of applicable codes, regulations, orders, Technical Specifications, license requirements or internal procedures or instructions having safety significance;

2. Significant operating abnormalities or deviations from normal or expected performance of station safety-related structures, systems, or components; and

3. Reportable occurrences as defined in the station Technical Specification 6.9.1.8.

Review of events covered under this paragraph shall include the results of any investigations made and recommendations resulting from such investigations to prevent or reduce the probability of recurrence of the event.

e. The Quality Assurance Department audit program at least once per 12 months and audit reports.

f. Any other matter involving safe operation of the nuclear power stations which is referred to the Director-Safety Evaluation and Control.

g. Reports and meeting minutes of the Station Nuclear Safety and Operating Committee.

AUTHORITY i

~

~ ~ : 6 .'5 .'2 . 9 The Director-Safety Evaluation and Control shall report to and advise the Manager-Nuclear Programs and Licensing, who shall advise the Vice President- Nuclear Operations on those areas of responsibility specified in Sectioc 6.5.2.7. .

RECORDS 6.5.2.10 Recore of SEC activities required by Section 6.5.2.7 shall be prepared and maintained in the SEC files and a summary shall be disseminated as indicated below each calendar month.

1. Vice President-Nuclear Operations

2. Nuclear Power Station Managers

3. Manager-Nuclear Operations Support NORTH ANNA - UNIT 2 6-10 l

t 1 - . . _ . - . . _ . .

ADMINISTRATIVE CONTROLS RECORDS (Cont'd)

4. Manager-Nuclear Programs and Licensing .

~~

5. Executive Manager-Quality Assurance

6. Others that the Director-Safety Evaluation and Control may

designate.

6.5.3 QUALITY ASSURANCE DEPARTMENT FUNCTION 6.5.3.1 The Quality Assurance Department shall function to audit station activities. These audits shall encompass:

a. The conforrance of facility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b. The performance, training and qualifications of the entire facility staff at least once per 12 months, c.

The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that

affect nuclear safety at least once per 6 months.

i d. The performance of activities required by the Operational Quality i

Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months.

I "

. . e. The Station Emergency Plan and implementing procedures at least once

.' ' per 12 months.

f. The Station Security Plan and implementing procedures at least once per 12 months. -

g. Any other area of facility operation considered appropriate by the Executive Manager-Quality Assurance or the Senior Vice President-Power Operations.

i h. The Station Fire Protection Program and implementing procedures at least once per 24 months.

i. An independent fire protection and loss prevention program inspec-tion and . audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.

J. An inspection and audit of the fire protection a d loss prevention program shall be performed by a qualified outside fire consultant at least once per 36 months.

I NORTH ANNA - UNIT 2 6-11 l

j

, ADMINISTRATIVE CONTROLS

k. The radiological environmental monitoring program and the results thereof at least once per 12 months.

1. The OFFSITE DOSE CALCULATION MANUAL and implementihg procedures at least once per 24 months. -

~

m. The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months.

n. The performance of activities required by the Quality Assurance Program to meet the provisions of Regulatory Guide 1.21, Revision 1 June 1974 and Regulatory Guide 4.1, Revision 1, April 1975 at least once per 12 months.

AUTHORITY 6.5.3.2 The Quality Assurance Department shall report to and advise the Executive Manager-Quality Assurance, who shall advise the Senior Vice President-Power Operations on those areas of responsibility specified in Section 6.5.3.1.

RECORDS 6.5.3.3 Records of the Quality Assurance Department audits shall be prepared and maintained in the department files. Audit reports shall be disseminated as indicated below:

1. Vice President - Nuclear Operations

2. Nuclear Power Station Manager

, 3. Manager-Nuclear Operations Support

4. Manager-Nuclear Programs and Licensing

5. Executive Manager - Quality Assurance _-

6. Director-Safety Evaluation and Control

7. Nuclear Power Station Manager-Quality Assurance l 8. Supervisor of area audited l

i 4

l NORLI ANNA - UNIT 2 6-12 Il

~

l l

. . ADMINISTRATIVE CONTROLS l 6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES:

l

a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

b. Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Commission shall be reviewed by the SNSOC and submitted to the Director-Safety Evaluation and Control and the Vice President -

Nuclear Operations.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The facility shall be placed in at least HOT STANDBY vithin one hour.

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President -

Nuclear Operations and the Director-Safety Evaluation and Control shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

c. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the SNSOC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission, the Director-Safety Evaluation and Control and the

~

Vice President - Nuclear Operations within 14 days of the violation. l 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33 Revision 2 February 1978.

b. Refueling operations.

NORTH ANNA - UNIT 2 6-13 i

I f

__ .. ~ , * -

)

' . ADMINISTRATIVE CONTROLS

c. Surveillance and test activities of safety related equipment. ,

1

d. Security Plan implementation.

~.

e. Emergency Plan implementation.

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation.

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. Quality Assurance Program for effluent and environmental monitoring, using the guidance in Regulatory Guide 1.21, Revision 1, June 1974

, and Regulatory Guide 4.1, Revision 1. April 1975.

6.8.2 Each procedure of 6.8.1 above, except 6.8.1.d and 6.8.1.e and changes thereto, shall be reviewed and approved by the SNSOC prior to implementation and reviewed periodically as set forth in administrative procedures.

Procedures of 6.8.1.d and 6.8.1.e shall be reviewed and approved as per 6.5.1.6.1 and 6.5.1.6.J.

6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant supervisory staff, at least one of whom holds a Senior Reactor Operator's i

License on the unit affected.

c. The change is documented, reviewed and approved by the SNSOC and j within 14 days of implementation.

j

= . ..

.- 6.8.4 The following programs shall be established, implemented, and maintained:

~

a. Primary Coolant Sources Outside Containment

! A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the recirculation spray, safety injection, chemical and volume control, gas stripper, and hydrogen recombiners. The program shall include the following:

(i) Preventive maintenance and periodic visual inspection l requirements and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less. I NORTH ANNA - UNIT 2 6-14

ADMINYSTRATIVE CONTROLS 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10 Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. ;

6.10.1 The following records shall be retained for at least five years:

a. Records and logs of facility operation covering time interval at each power level.

b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c. Each REPORTABLE OCCURRENCE submitted to the Commission.

d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

e. Records of changes made to Operating Procedures.

f. Records of radioactive shipments.

g. Records of sealed source leak tests and results.

h. Records of annual physical inventory of all dealed rource material of record.

1. Records of the annual audit of the Station Emergency Plan end implementing procedures.

j. Records of the annual audit of the Station Security Plan and implementing procedures.

= * ~

. 6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a. Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c. Records of facility radiation and contamination surveys.

d. Records of radiation exposure for all individuals entering radiation control areas.

e. Records of gaseous and liquid radioactive material release to the environs.

f. Records of transient or operational cycles for those facility components identified in Table 5.7-1.

NORTH ANNA - UNIT 2 6-22

i .- . _

DISCUSSION OF PROPOSED TECHNICAL SPECIFICATION CHANCES The enclosed changes reflect the recent reorganizationalkchanges in the Nuclear Operations Department. These changes are reflected on Technical Specifications Figures 6.2-1 and 6.2-2.

The changes in the Nuclear Operations Department begin with the creation of the Manager, Nuclear Programs and Licensing. The Manager, Nuclear Programs and Licensing reports to the Vice President-Nuclear Operations. His responsibilities and authorities are in the areas of emergency planning, licensing and independent review. Another newly created position is the Assistant Station Hanager (Nuclear Safety and Licensing). He reports to the Station Manager and assumes certain responsibilities and authorities previously held by the Station Manager and Assistant Station Manager (Operations and Maintenance) in the area of the Station Nuclear Safety and Operating Committee. He also has responsibility and authority for emergency planning, the safety engineering staff and licensing. The title of the other Assistant Station Manager before the creation of the Assistant Station Manager (Nuclear Safety and Licensing) has been changed to the Assistant Station Manager (Operations and Maintenance). His responsibility and authority have not changed.

In addition, the following organizational changes were previously in effect:

} The Superintendent of Projects reports to the Station Manager and is 4 responsible for the management of plant modifications / maintenance i activities which are performed as capital projects.

The Licensing Coordinator reports to the Assistant Station Manager (Nuclear Safety and Licensing) and is responsible for the coordination of

.=

=

. licensing activities and regulatory requirements for the power station.

The Coordinator Emergency Planning reports to the Assistant Station Manager (Nuclear Safety and Licensing) and is responsible for implementation of the Energency Plan at the station and the duties

! described therein.

The Supervisor Engineering (Safety Engineering Staff) now reports to the Assistant Station Manager (Nuclear Safety and Licensing).

1 The Supervisor -

Maintenance Services reports to the Superintendent Maintenance. He is responsible for evaluating maintenance procedures and methods to ensure compliance with regulatory requirements and company policy, developing a maintenance management program for preventive and

! scheduled routine maintenance, developing and monitoring a spara and replacement parts program, developing plans for impending outages, and recommending requirements for new and revised maintenance training programs. -

l l

l

, _ _ - _ _ . _ . _ _ _ , _ _ _ __,...._.,--..-_.--.____.__-m _- . _ _ - _ _ _ _ _

2 The Station Security Supervisor reports directly to the Director Nuclear Security (Off-site) and is responsible for the , direction of the activities of the Station Security Organization and'" ensures that the provisions of the Security Program are carried out.

Several new positions now report to the Supervisor. Administrative Services. These are briefly discussed below.

The Safety Supervisor coordinates station training prograns/or occupational safety needs with the Station Training Department and/or station department heads.

Systems Supervisor is responsible for the The Business administration of accounting activities and monitoring cost control and budget activities.

Supervisor Personnel Services is responsible for the The coordination of station personnel policies.

The Supe rvisor, Records Management, maintains station files in accordance with applicable regulations and develops and implements a station records management program.

The title of the Fire Marshall has been changed to the Loss Prevention Supervisor. His responsibilities have not changed.

The Supervisor Engineering (Performance and Testing) reports to the Superintendent Technical Services and is responsible for providing engineering technical support for station operations, administering the periodic test program to insure compliance with the Technical Specifications, optimizing plant performance, and providing 5 - - continuous technical evaluation of reactor operation.

The Supervisor Engineering (D/C and Projects) reports to the Superintendent-Technical Services and is responsible for providing the administrative control and technical evaluation of all station modifications and design changes.

The Supervisor Engineering (Planning) reports to the Assistant Station Manager (Operations and Maintenance) and is responsible for the preparation and administration of schedules for all maintenance and design change activities during unit outages.

Supervisor Quality Control QA Activities reports to the Manager, Quality Assurance. In this capacity, the Supervisor Quality Control QA Activities shall develop, maintain and implement an improved quality assurance auditing program for North Anna' Power Station to assure that technical requirements including the design bases, applicable regulatory requirements and specified codes and standards are cor, r ectly translated into specifications, drawings, procedures or instructions. He may also serve in an advisory capacity to the Station Nuclear Safety and Operating Committee.

l i

I 3

Specification 6.2.3.4 currently refl'ects that the Safety Engineering Staff (SES) makes detailed recommendations to the Station Manager and the Director-Safety Evaluation and Control. Now the SES wilh' make detailed recommendations to the Assistant Station Manager (Nuclear Safety and Licensing).

I Specification 6.5.1.2 deals with the comporition of the Station Nuclear Safety and Operating Committee (SNSOC). Currently, the Chairman is the Station Manager and the Vice Chairman is the Assir. tant Station Manager. The proposed changes are to have the Assistant Station Manager (Nuclear Safety and Licensing) be the Chairman of SNSOC and the Assistant Station Manager (Operations and Maintenance) be the Vice Chairman of SNSOC. This change does 4

not create any new authorities or responsibilities within the Nuclear j Operations Department; rather by reducing the span of control of the affected managers (both new and existing management), control and effectiveness in the areas of concern have been enhanced. Thus more management attention will be focused on the significant issues in the areas of nuclear safety, licensing and emergency prepardness.

Specification 6.5.1.6 deals with the responsibilities of SNSOC. The proposed changes will allow the SNSOC to recommend Appendix "A" Technical Specification

, changes and Appendix "B" Environmental Protection Plan changes to the Station I

Manager instead of just reviewing proposed changes. Previously, the Chairman SNSOC reviewed and received recommended changes to the Plant Security Plan and implementing procedures and Emergency Plan and implementing procedures. The proposed changes are to have SNSOC review changes to the Plant Security Plan and implementing procedures and Emergency Plan and implementing procedures but

they shall submit recommended changes to these plans to the Station Manager 1 instead of the Chairman SNSOC. The SNSOC currently reviews and approves

, changes to the Process Control Program (PCP) and the Offsite DOSE Calculation Manual (ODCM). The proposed change would have SNSOC review changes to the PCP

. and ODCM but the Station Manager would approve any changes. These are more

. .I

. programmatic changes and the Station Manager will be responsible for them.

i Specification 6.5.1.7 currently gave SNSOC the authority to only recommend to the Station Manager written approval or disapproval of items considered ynder

}

6.5.1.6(a) through (d) . The proposed changes would allow SNSOC to pr' ovide written approval or disapproval of items considered under the current 6.5.1.6(a), 6.5.1.6(b) and 6.5.1.6(d). There is also a reordering of the

first four items in 6.5.1.6. '

1 Specification 6.5.1.8 currently tells SNSOC to provide copies of the written j minutes of the SNSOC meeting to the Manager-Nuclear Operations and Maintenance

and the Director-Safety Evaluation and Control. The _ proposed changes will l provide copies of the SNSOC meeting minutes to the Station Manager, Vice f

President-Nuclear Operations and the Director-Safety Evaluation and Control.

Specification 6.5.2.9 currently has the Director-Safety Evaluation and Control (SEC) reporting to and advising the Manager-Nuclear Technical Services. The title of the Manager-Nuclear Technical Services is outdated'and should be the

! Manager-Nuclear Programs and Licensing.

I

4 Specification 6.5.2.10 should be revised to have SEC prepare and maintain in

' the SEC files records of SEC activities. These records should be disseminated monthly to the; 1) Vice President-Nuclear Operations, 2) Nuclear: Power Station Managers, 3) Manager-Nuclear Operations Support, 4) Manager-Nuclear Programs and Licensing, 5) Executive Manager-Quality Assurance and 6) Others that the Director-Safety Evaluation and Control may designate.

! Specification 6.5.3.3 should be revised to have the Quality Assurance Department audits be prepared and maintained in the department files. Audit reports should be disseminated to the; 1) Vice President-Nuclear Operations,

2) Nuclear Power Station Manager, 3) Manager-Nuclear Operations Support,

4) Manager-Nuclear . Programs and Licensing, 5) Executive Manager-Quality Assurance 6) Director-Safety Evaluation and Control 7) Nuclear Power Station Managers Quality Assurance, and 8) Supervisor of areas audited.

6.8.1, and changes Specification 6.8.2 previously had each procedure in j

thereto, be reviewed by the SNSOC and approved by the Station Manager prior to implementation and reviewed periodically as set forth in administrative 6.8.1 d 2

procedures. The proposed changes are to have each procedure, except and e be reviewed and approved by the SNSOC prior to implementation and reviewed periodically as set forth in administrative procedures. Procedures of 6.8.1 d and e shall be reviewed and approved as per 6.5.1.6.1 and j. This would keep the programmatic responsibility with the Station Manager.

Specification 6.8.3 has temporary changes made to procedures of 6.8.1 being provided to and reviewed by SNSOC and approved by the Station Manager within 4 14 days of implementation. The proposed change would allow SNSOC to review and approve temporary changes made to procedures in 6.8.1 within 14 days of implementation.

Also, reorganizations in the Nuclear Operations Department have occurred in 2 the past few years. The title of the Manager, Nuclear Operations and l ,

~

, Maintenance has been upgraded to the Vice President-Nuclear Operations. The i

- position of the Manager, Nuclear Operations and Maintenance has been renamed the Manager Nuclear Operations Support and the title of the Manager, Nuclear Technical Services has been renamed the Manager, Nuclear Programs and Licensing. The Technical Analysis and Control Group has been deleted ~ The Section Supervisor. Administrative Services has been renamed Director, Administrative Services and reports to the Manager, Nuclear Operations Support. The title of the Director, Operations and Maintenance Services has been changed to the Director, Operations and Maintenance Support. The titles of the Section Supervisor, Training and Section Supervisor, Operation and Maintenance Support have been deleted from Figure 6.2-1. The function of emergency planning has been added in the Nuclear Operations Department. The Director, Emergency Planning will report to the Manager, Nuclear Programs and Licensing. The title of the Director, Chemistry and Health Physics has been revised to Director. Health Physics. The chemistry function will be the responsibility of the Director, Operations and Maintenance Support.

Because the reorganization only results in a redistribution of existing authorities and responsibilities to enhance management controls in selected areas, this change is considered administrative in nature. Thus, because the change is administrative in nature, no unreviewed safety question is involved.

5 Reorganizations in the Quality Assuran'ce Department have occurred in the past

' few years. To reflect the current organization, the following changes must be made in Section 6.0 of the Technical Specifications. The; title of the Manager-Quality Assurance, Operations has been revised to ;be the Executive Manager-Quaility Assurance. The title of the Nuclear Power- Resident Quality Control Engineer has been changed to the Nuclear Power Station Manager Quality Assurance and he reports to the Executive Manager-Quality Assurance. In addition, the title Director-Quality Assurance, Nuclear Operations and Director-Quality Asserance, Operations have been deleted.

Having the Nuclear Power Station Manager Quality Assurance report directly to the Executive Manager-Quality Assurance will enhance the Quality Assurance Program of the Company.

A new department called Maintenance and Performance Services has been created.

The creation of the Maintenance and Performance Services Department will aid in the quality of training activities at the power stations. The Superintendent. Nuclear Training reports directly to the Director, Nuclear Training offsite.

He also has communication with the Station Manager. The Director, Nuclear Training reports to the Manager Power Training Services and he reports to the Manager, Maintenance and Performance Services.

The Maintenance and Performance Services Department will plan, organize, direct and control nuclear training, so that, effective and efficient technical training is provided to the Nuclear Operations staff. They will assess and recommend specific training requirements for regulatory agencies as applicable and coordinate program offerings as necessary.

The title of the Executive Vice President-Power has changed to the Executive Vice President and Chief Operating Officer. The Executive Vice President-Power previously issued a management directive, on an annual basis to all station personnel, the responsibilities of the Control Room command

=

function of the Shift Supervisor. The Senior Vice President-Power Operations w

. ;~ ill~ sign the management directive on the Shift Supervisors responsibilities and-issue this to all station personnel on an annual basis.

The Security Department has also had a reorganization. The Station SecGrity Supervisor reports to the Director, Nuclear Security at the corporate office.

The Station Security Supervisor continues to have communications with the

Supervisor Administrative Services at the Station.

i In addition, there is a change to Technical Specification 6.10 which adds Technical Specifications 6.10.11 and 6.10.1j. The reason for the addition of Technical Specification 6.10.11 is because 10 CFR 50.54 (t) requires the retention of records for at least five years when the Station Emergency Plan and implementing procedures are audited annually. The reason for the addition of Technical Specification 6.10.1j is because 10 CFR 73.46g(6) requires the retention of records for at least five years when the Station Security Plan and implementing procedures are audited annually.

t The proposed change to the referenced ANSI standard on Facility Staff l Qualifications (Section 6.3) and Training (Section 6.4) reflects the ANS standard specified in Vepco's QA Topical Report, " Quality Assurance Program Operations Guide 1.8 Phase". Amendment 4, regarding Vepco's position on NRC Regulatory

" Personnel Qualification and Training". The QA Topical Report was

c

. 6 o ,

approved on October 6, 1982. Thus, the change amends the Technical Specifications to make them consistent with the NRC approved QA Topical Report. The specific change replaces ANSI N18.1-1971 with ANS 3.1-(12/79 Draft). ANS 3.1-(12/79 Draft) meets or exceeds the requirements of the older ANSI standard.

These proposed changes do not pose a significant hazards consideration as defined in the Federal Register, 48 FR 14870 Example (i); a purely administra:ive change to the technical specifications; for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.

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l May 2, 1985

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Mr. Harold R. Denton, Director Serial No.85-162 Office of Nuclear Reactor Regulation N0/JHL:acm Attn: Mr. James R. Miller, Chief Docket Nos. 50-338 50-339 Operating Reactors Branch No. 3 Division of Licensing License Nos. NPF-4

NPF-7 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 I Gentlemen:

' VIRGINIA P0k'ER NORTH ANNA P0kTR STATION UNIT NOS.1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGES Pursuant to 10 CFR 50.90, Virginia Power requests an amendment in the form of changes to the Technical Specifications to Operating License Nos. NPF-4 a

and NPF-7, for North Anna Unit Nos.1 and 2.

i The proposed Technical Specification char.ges apply to Technical Specification 3.6.1.3, Containment Air Locks. The changes expand and 2 clarify the action requirements, change the allowabis seal leakage from zero to a small measurable amount and add a footnote to surveillance requirement 4.6.1.3b to make the provisions of Spcification 4.0.2 not applicable. These changes will provide consistency between the North Anna

!, 1 and 2 Technical Specifications and also the Standard Technical

'; Specifications (STS) for Westinghouse Pressurized h*ater Reactors,

! Revision 4 and Revision 5, Draft. The addition of the footnote also l

provides consistency with the requirements of 10 CFR 50, Appsndix J.

i l

The proposed Technical Specification changes for North Anna Unit No.1 are ,

provided in Attachment 1. The proposed Technical Specification changes for North Anna Unit No. 2 are provided in Attachment.2. A discussion of the proposed Technical Specification changes is provided in Attachment 3.

This request has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control Staf f. It

- has been determined that this request does not involve an unreviewed safety question as defined in 10 CFR 50.59 or significant hazards i

consideration as defined in 10 CFR 50.92, i

l f, n 06 P P m s m 8s enn .

G 4SI A

1 We have reviewed this request in accordance with the criteria in 10 CFR

- 170.12. A voucher check in the amount of $150 is enclosed as an application fee.

Very truly yours,

, -~~ ..

- E.L . $ w W. L. Stewart Attachments

1. Proposed Technical Specification Changes - Unit 1

2. Proposed Technical Specification Changes - Unit 2

3. Discussion of Proposed Technical Specification Changes

4. Voucher Check for $150 cc: Dr. J. Nelson Grace Regional Administrator Region II Mr. M. W. Branch NRC Resident inspector

= . . North Anna Power Station

~ Mr. Charles Price Department of Health

. 109 Governor Street .

~

Richmond, Virginia 23219 t

1 .

i

COMMONVEALTH OF VIRGINIA )

)

CITY OF RICHMOND )

The foregoing document was acknowledged before me, in and for the City and Commonwealth aforesaid, today by W. L. Stewart who is Vice President -

Nuclear Operations, of Virginia Power. He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before as this a day of b}. 19 $ .

My Commission expires: f ..

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ATTAC10fENT 1 6

I MDR SO70367 850502

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CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION :

3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to 0.05 L at a

P,, greater than or equal to 40.6 psig.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2. Operation may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days. -

3. Otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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4. The provisions of Specification 3.0.4 are not applicable.

b. With a containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door closed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the fo,11owing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

t l

NORTH ANNA - UNIT I 3/4 6-4 l

I

1 SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERA 5LE,:

a. *'.lf thin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following closing, except when the alt lock is being used for multiple entries, then at least once per 72 - hours, by verifying precisionthat themeasurements flow seal leakage when is less than 0.01 measured forL,leastas at determined 30 secondsby with the volume between the seals at a pressure of greater than or equal to 40.6 psig.

b. At least once per 6 months by conducting an overall air lock leakage i test at greater than or equal to P 40.6 psig, and by verifying that the overall air lock leakage rate Is, within its limit , and

c. At least once per 18 months during shutdown by verifying that only

, one door in each air lock can be opened at a time.

1 1

7 3 .

i j

4 Exempt to Appendix "J" of 10 CFR 50.

lTheprovisionsofSpecification4.0.2arenotapplicable ..

l NORTH ANNA - UNIT 1 3/4 6-4a l

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CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CORDITION FOR OPERATION .

3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to 0.05 L* at P,, greater than or equal to 40.6 psig.

APPLICABILITT: MODES 1, 2, 3 and 4 ACTION:

a. With one containment air lock door inoperable:

1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.

2. Operation- may then continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock door is verified to be locked closed at least once per 31 days.

3. Otherwise, be in at least HOT SIANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4. The provisions of Specification 3.0.4 are not applicable.

b. With a containment air lock inoperable, except as the result of an inoperable air lock door, maintain at least one air lock door c1,osed; restore the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I NORTH ANNA - UNIT 2 3/4 6-4 l

l

l SURVEILLANCE REQUIREMENTS

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4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a. *Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following closing, except when the hir- lock is being used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , by verifying that the seal leakage is less than 0.01 L as determined by precision flow censurements when measured for at*1 east 30 seconds with the volume between the seals at a pressure of greater than or equal to 40.6 psig.

b. At least once per 6 months by conducting an overall air lock leakage test at greater than or equal to P the overall air lock leakage rate Is, within its limit , and40.6 psig, and by

c. At least once per 18 months during shutdown by verifying that only one door in each air lock can be opened at a time.

l 9

, = .

l ,

Exempt to Appendix "J" of 10 CFR 50.

l 1

lTheprovisionsofSpecification4.0.2arenotapplicable.

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n NORTH ANNA.- UNIT 2 3/4 6-5

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i ATTACINENT 3 O

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DISCUSSION OF PROPOSED TECHNICAL SPECIFICATION CHANGES The proposed changes to the North Anna Unit 1 Technical Specifications are toexpandandclarif[theACTIONrequirementsofTechnicalSpeification 3.6.1.3, Containment Air Locks. In addition, a footnote has been added to 4

surveillance requirement 4.6.1.3b to make the provisions of Specification 4.0.2 not applicable. 'nese revisions will provide consistency between

'the North Anna 1 and 2 Technical Specifications and also with the

, Standardized Technical Specifications (STS) for Westinghouse PWR's, i Revision 4 and Revision 5, Draft. The addition of the footnote also provides consistency between the Unit 1, Unit 2, STS and the requirements of 10 CFR 50, Appendix J.

,)

i i

{ A change to surveillance requirement 4.6.1.3a is being proposed for North i

Anna Unit Nos. l'and 2. The proposed change is in the surveillance .

requirement for door seal, leakage. The proposed change will reduce the amount of maintenance required on the containment air locks. By reducing the amount of maintenance required, the amount of personnel radiation

= . .

exposure will be reduced.

i i Surveillanc tosting of the air lock' seals will continue to assure that i

the overa'11' air lock leakage will not b' come e excessive due to seal damage during the ' intervals between air lock leakage tests. The overall air lock

/

leakage limit of 5 0.05 L , and the total type B and C leakage limit of 5 0.60-L , will be unchanged. The proposed surveillance requirement for door seal leakage limit of 5 0.01 L , is 'a small part of either of the above limits and therefore does not coAs'itutet a significant change to the l _ O >l !

Technical Specifications.

~

It has been determined that an "unreviewed safety question," as[ defined in 10CFR50.59, does not exist because:

a. This change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the final safety Analysis i

Report. The surveillance frequency remains unchanged, and the l

Air Lock leakage limits insure overall air lock integrity.

4

b. This change does not create a possibility for an accident or malfunction of a different type than any evaluated previously -

in the Final Safety Analysis Report. No physical or operational feature of the Air Lock is altered.

c. This change does not reduce the margin of safety as defined in j ,

. - .the basis of any Technical Specification. The overall air lock leakage and the total type B and C test leakage limits established in 10CFR50 Appendix J remain unchanged.

~

The proposed changes to Technical Specification 3.6.1.3 have been determined not to pose a significant hazards consideration as stated in the Federal Register dated April 6, 1983. Example vil of examples'of amendments that are considered not likely to involve a significant hazards consideration states, "A change to make a license conform to changes in

the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations". The propesed changes are consistent with the Standardized ;{ Technical Specifications for Westinghouse PWR's, Revision 4 and Revision 5, Draft.

O i

l 1

i

ha bq)7/90 ?% < dy y/p VIROINIA EI.ECTRIC AND Powen COMPANY RIC11MOND.VIHOINI A 2 02 61 W.L. STEWART vics Passionwt Nt ctsan Orsaattons September 24, 1985 Mr. Harold R. Denton, Director Serial No.85-686 Office of Nuclear Reactor Regulation N0/EJL:acm Docket Nos. 50-338 Attn: Mr. Edward J. Butcher, Acting Chief 50-339 Operating Reactors Branch No. 3 Division of Licensing License Nos. NPF-4 NPF-7 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGES Pursuant to 10 CFR 50.90, Virginia Electric and Power Company requests an to amendment, in the form of changes to the Technical Specifications, Operating License Nos. NPF-4 and NPF-7 for North Anna Power Statior. Units 1 and 2.

We are proposing changes to Section 4.7.10, Surveillance Requirements for Snubbers, that would modify portions of the visual inspection acceptance criteria, establish separate sampling methods for the functional tests for small bore and large bore snubbers, and establish functional test methods for large bore snubbers.

a.

The ' proposed changes to the Technical Specifications are provided in Attachments 1 and 2 for Units 1 and 2, respectively. A discussion of the proposed changes to the Technical Specifications is provided in Attachment 3. In accordance with 10 CFR Part 170 an Application Fee of . -

$150 is also enclosed.

Thb request has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control Staff. It has been determined that this change does not pose an unreviewed safety question as defined in 10 CFR 50.59 and does not pose a significant hazards consideration as defined in 10 CFR 50.92.

As we have discussed with members of your staff, Unit 1 is scheduled to begin a refueling outage November 1,1985. We would like to implement these requirements with the snubber work that will be performed during this outage. We would appreciate any assistance that you and your staff can provide in expediting this request.

$EM lty -

4pote4%

1

VIROINIA ELacT2:e axo Powra Courawy to If you have any questions, or need additional information to process this request, please contact us.

Very truly yours, ta\ m W. L. Stewart Attachments:

1. Proposed Technical Specification Change for North Anna Unit 1

2. Proposed Technical Specification Change for North Anna Unit 2

3. Discussion of Proposed Changes

4. Application Fee cc: Dr. J. Nelson Grace Regional Administrator h7C Region II Mr. M. W. Branch h7C Resident Inspector

, . , North Anna Power Station Mr. C. Price Department of Health 109 Governor Street ~

Richmond, Virginia 23219 I

n a

COMMONWEALTH OF VIRGINIA )

)

CITY OF RICHMOND ) _

The foregoing document was acknowledged before me, in and for the City and Commonwealth aforesaid, today by W. L. Stewart who is Vice President He is duly Nuclear Operations, of Virginia Electric and Power Company.

authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

l Acknowledged before me this aq day of E qAe,,L , 19 5 5 1

My Commission expires: SAwc w to ,19 iS .

ke, .\c%h Notary Public

i. .

l (SEAL) l l

I S/001

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ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGE FOR NORTH ANNA UNIT 1 a , ,

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1 PLANT SYSTEMS i . , SURVEILLANCE REQUIREMENTS (Continutd)

, b. Visual Inspection Acceptance Criteria 9 1

' Visual inspections shall verify (1) that there are no visual indica-tions of damage or impaired OPERABILITY, (2) attachments to the <

l l,

foundation or supporting structure are secure, and (3): in those

' locations where snubber movement can be manually induced without disconnecting the snubber, that the snubber has freedom of movement /

and is not frozen up. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the - rejection is clearly established and remedied for -

that particular snubber and for other snubbers that may be generic-ally susceptible; and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specifications j - 4.7.10.d and 4.7.10.e. When hydraulic snubbers which have uncovered fluid ports are tested for operability, the tests shall be performed 2

' by starting with the piston at the as-found setting and extending the piston rod in the tension mode direction. Snubbers which have been determined to be inoperable as a result of unexpected transients, isolated damage, or other random events, and can not be proven operable by functional testing for the same reasons, shall not be counted in determining the next visual inspection period when the

' provision in 4.7.10.c that failures are subject to an engineering evaluation of component structural integrity has been met and equipment has been restored to an operable state via repair and/or

' replacement as necessary.

! c. Functional Tests At least once per 18 months during shutdown, a representativ sample of small bore snubbers which follows the expression 35 ( 1 + ) ,

t where c=2 is the allowable number of small bore snubbers not meeting l by the ' operator, shall be

the acceptance criteria selected i

functionally tested either in-place or in a bench test. For each l

! number of . small bore snubbers above c which does not meet the functional test acceptance criteria of Specification 4.7.10.d or l

4.7.10.e an additional sample selected according~to the expression l

35 (1 + y) ( c 1 (a - c) i shall be functionally tested where a is the total number of small bore snubbers found inoperable during the functional testing of l I

l the representative sample.

t

! Functional testing shall continue according to the expression 1

b 2) where b is the number of snubbers (35 (1 + *y)ble in f

found inopera the previous re-sample, until no additional l

inoperable enubbers are found within a sample or.until all small bore snubbers in Tables 3.7.4a and 3.7.4b have'been functionally tested.

t l

l.

NORTH ANNA - Unit 1 3/4 7-29

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continutd)

At least once per 18 months during shutdown, 10% of the large bore snubbers (snubbers greater than 50 kips) shall be functionally tested full snubber bench test, or in a snubber valve either in place, in a block bench test. For each large bore snubber that does not meet the an functional test acceptance criteria of Specification 4.7 10.d If the engineering evaluation is required to determine the failure mode.

failure is determined to be generic, an additional 10% of that type of snubber shall be functionally tested. If the failure is determined to be non-generic, an additional 10% of that type of snubber will be tested during the next functional test period.

- The representative samples selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of snubbers. At least 25% of the snubbers in the representative samples shall include snubbers from the following three

- categories:

1. The first snubber away from each reactor vessel nozzle

2. Snubbers within 5 feet of p heavy equipment (valve, pump, turbine, motor, etc.).

3. Snubbers within 10 feet of the discharge from a safety relief valve.

to Snubbers identified in Tables 3.7.4a and 3.7.4b as "Especially Difficult Remove" or in "High Radiation Zones During Shutdown" shall also be included in the representative sa=ples.* Tables 3.7.4a and 3.7.4b may be used jointly or separately as the basis for the sampling plan.

In addition to the regular sample, snubbers which failed theIfprevious a spare functional test shall be retested during the next test period.

snubber has been installed in place of a failed snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare

,i . snubber shall be retested. Test results of these snubbers may not be included

'for the re-sampling.

If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by to manufacturer or design deficiency all snubbers of the same design subject the same defect shall be functionally tested. This fortesting requirement snubbers not meeting shall the be independent of the requirements stated above functional test acceptance criteria.

The requirement to functionally test large snubbers greater than 50 kips around the steam generators and reactor coolant pumps, is exempt from functional testing for the representative sample of snubbers selected for testing during the Cycle 3 refueling and maintenance outage.

Permanent or other exemptions from functional testing for individual snubbers in these categories may be granted by the Commission only if a justifiable basis for exemption is presented and/or snubber life destructive testing at either was performed to qualify snubber operability for all design conditione the completion of their fabrication or at a subsequent date.

NORTH ANNA - UNIT 1 3/4 7-29a !

4 0 e t

ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGE FOR NORTH ANNA UNIT 2 e

o O

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continutd)

b. Visual Inspection Acceptance Criteria Visual inspections shall verify (1) that there are no visual indica-tions of damage or impaired OPERABILITY, (2) attachm'nts e to the foundation or supporting structure are secure, and'.(3) in those locations where snubber movement can be manually induced without disconnecting the snubber, that the snubber has freedom of movement and is not frozen up. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the cause of the the next visual inspection interval, providing that (1) rejection is clearly established and remedied for that particular snubber and for other snubbers that may be generic-ally susceptible; and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specifications 4.7.10.d and 4.7.10.e. When hydraulic snubbers which have uncovered fluid ports are tested for operability, the tests shall be performed by starting with the piston at the as-found setting and extending the piston rod in the tension mode direction. Snubbers which have been determined to be inoperable as a result of unexpected transients, isolated damage, or other random events, and can not be proven operable by functional testing for the same reasons, shall not be counted in determining the next visual inspection period when the provision in 4.7.10.c that failures are subject to an engineering evaluation of component structural integrity has been met and equipment has been restored replacement as necessary, to an operable state via repair and/or

c. Functional Tests At least once per 18 months during shutdown, a representative sample ofsmallboresnubberswhichfollowstheexpression35(1+y),

where c=2 is the allowable number of small bore snubbers not meeting

the acceptance criteria selected by the operator, shall be function-ally tested either in-place or in a bench test. For each number of 2 -

small bore snubbers above c which does not meet the functional test l acceptance criteria of Specification 4.7.10.d or 4.7.10.e, an addi-

' tional sample selected according to the expression 35 (1 + y) ( c 1 (a - c) i shall be functionally tested, where a is the total number of small bore snubbers found inoperable during the functional testing of the representative sample.

Functional testing shall continue according to the expression b(35(1+y)2)wherebisthenumberofsnubbersfound inoperable in the previous re-sample, until no additional inoperable snubbers are found within a sample or until all small bore snubbers in Tables 3.7.4a and 3.7.4b have been functionally tested.

l .

NORTH ANNA - Unit 2 3/4 7-26a

PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continusd)

At least once per 18 months during shutdown, 10% of the large bore snubbers (snubbers greater than 50 kips) shall be functionally tested either in place, in a full snubber bench test, or in a snubber valve 41ock bench test. For each large bore snubber that does not meet the Specification 4 7 10.d an functional test acceptance criteria of If the engineering evaluation is required to determine the failure mode.

failure is determined to be generic, an additional 10% of that type of snubber shall be functionally tested. If the failure is determined to be non-generic, an additional 10% of that type of snubber will be tested during the next functional test period.

The representative samples selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of snubbers. At least 25% of the snubbers in the representative samples shall include snubbers from the following three categories:

1. The first snubber away from each reactor vessel nozzle

2. Snubbers within.5 feet of heavy equipment (valve, pump, turbine, motor, etc.).#

3. Snubbers within 10 feet of the discharge from a safety relief valve.

to Snubbers identified in Tables 3.7.4a and 3.7.4b as "Especially Difficult Remove" or in "High Radiation Zones During Shutdown" shall also be included in the representative samples.* Tables 3.7.4a and 3.7.4b may be used jointly or separately as the basis for the sampling plan.

In addition to the regular sample, snubbers which failed theIf previous a spare functional test shall be ratested during the next test period.

stubber has been installed in place of a failed snubber, then both the failed 5 : snubbe~ri(if it is repaired and installed in another position) and the spare snubber shall be retested. Test results of these snubbers may not be included for the re-sampling.

If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen in place, the cause will be evaluated and if caused by manufacturer or design deficiency all snubbers of the same design subject to the same defect shall be functionally tested. This l

testing requirement shall be independent of the requirements stated above for snubbers not meeting the functional test acceptance criteria.

! #The requirement to functionally test large snubbers greater than 50 kips, around the steam generators and reactor coolant pumps, is exempt from f functional testing for the representative sample of snubbers selected for testing during the Cycle 1 refueling and maintenance outage.

Permanent or other exemptions from functional testing for individual snubbers in these categories may be granted by the Comunission only if ja justifiable l basis for exemption is presented and/or snubber life destructive testing was l

performed to qualify snubber operability for all design conditions at either the completion of their fabrication or at a subsequent date.

NORTH ANNA - UNIT 2 3/4 7-26b

D e g

a b

e ATTACHMENT 3 DISCUSSION OF PROPOSED CHANGES

= .

O O

e

i i ,

DISCUSSION OF PROPOSED CHANGES The proposed change will modify the method in which the operability of

! hydraulic snubbers with uncovered fluid ports is determined. The Technical Specifications presently require a hydraulic snubber to be declared inoperable if its fluid port has been found uncovered and that the snubbdr cannot be ;

determined operable via functional testing for the purpose of establishing the next visual inspection interval.

j The proposed Technical Specification will ,

permit the functional testing of snubbers which have been found with uncovered fluid ports as a method of determining snubber operability for the purpose of j

establishing the next visual inspection interval. When testing hydraulic

) snubbers which have uncovered fluid ports, the tests shall be performed by j

starting with the piston at the as-found setting and extending the piston rod in the tension mode direction.

]

i

' Functionally testing a hydraulic snubber with an uncovered fluid port from the as-found condition will determine if the snubber would have functioned as

{ designed. . Testing a snubber from the as-found condition in the tension mode

is a conservative method of testing because this requires fluid to be supplied to the snubber valve block and cylinder to accommodate piston rod movement. If

! it can be shown by a functional test from the as-found setting that a snubber i with an uncovered fluid port functions properly, the snubber should have functioned in fact operable.

properly, if required, when it was installed in the plant and was a

4 The probability of an accident is not increased and the

margin of safety is not decreased by this change because functionally testing anubbers with uncovered fluid ports from the as-found condition is an accept-able method of determining snubber operability. This change is similar to l

another . f acility's Technical Specification requirements for snubbers with uncovered fluid ports.

l The proposed change will- add a statement concerning snubbers found to be

{

inoperable 'as a result of physical . damage sustained as a result of random events.

I The Technical Specifications presently allow a hydraulic snubber which was determined to be inoperable by visual inspection, to be determined i

f , operable for the purpose of establishing the next visual inspection interval

' if the ~cause of the rejection is clearly understood and a functional test determines the snubber to be operable. The proposed change will permit j inoperable snubber that an

cannot be determined operable by functional to be declared ' operable for the purpose of establishing a new inspection testing, j

interval if it can be determined that the snubber was rendered inoperable as a result of unexpected transients, isolated damage or other' random events.

Examples of events which would be considered random , or isolated include an j object on inadvertently dropped on a snubber or a chainfall accidentally anchored a snubber. An engineering evaluation of component structural integrity j

would still be performed after esch failure.

l If it can be determined that a snubber was rendered inoperable as a result of unexpected transients, isolated damage or other random events, similar failures would not be anticipated. Additional inspections of snubbers would therefore

! not be needed to determine overall snubber operability. The probability of an accident is not increased and the margin of safety is not decreased by this change because snubber failures which are determined to be isolated in nature do not affect overall snubber operability.

Page 2 l Ths propossd changs also codifies and clarifies the number of small and large bore snubbers that require functional testing. Snubbers will be broken into two groups, small bore and large bore. The formula [35 (1 + C/2)] will be

used to calculate only the initial sample sire of small bore snubbers to be

functionally tested. The initial sample size for large bore snubbers will be

ten percent of the total number of large bore snubbers. The present Technical l Specification does not separate small bore and large bore snubbe a into groups j

or specify the number of large bore snubbers which must be ; tested. By i separating small bore and large bore snubbers into groups and celculating a

sample size for each group, a specific number of large bore snubbers will be tested. The total number of snubbers requiring testing would increase by the j number of large bore snubbers tested because the formula 35 (1 + C/2) would be j used to determine only the initial sample size for small bore snubbers.

j Generic Letter 84-13 recommends both the formula 35 (1 + C/2) and ten percent i of the total number of each type snubber in the plant as methods to determine l an initial sample size for functional testing. The method of selecting a ten

] percent sample size for large bore snubbers was selected because the formula j ,

35 (1 + C/2) for initial sample size and the formula currently in the j Technical Specification for determining an additional sample can not be i applied to a small test group such as large bore snubbers.

l In the event of small bore snubber failures, the proposed change will not 5~ alter the current method of testing additional small bore snubbers. The !

I formula now present in the Technical Specifications will be used to calculate

I the number of snubbers in the additional sample to be functionally tested.

I In the event of a large bore snubber failure. an engineering evaluation will be performed to determine if the failure is generic in nature. If the failure ,

is determined to be generic in nature, an additional sample of ten percent of j the large bore snubbers will be functionally tested during the current functional test period. If the failure is determined to be non-generic in

{

- nature, functional testing of an additional sample of ten percent of the large '

- bore snubbers will be postponed until the next functional test period. A large maintenance effort is required to remove large bore snubbers which are then sent off site for testing. Testing an additional sample of ten percent of the large bore snubbers during the current test period is not warranted for i : a. specific non-generic failure. Testing during the next test interval allows j 2

for scheduling of work to support the maintenance effort and will minimize the i impact of snubber testing on plant operation. The probability of an accident is not increased and the margin of plant safety is not decreased by postponing functional testing until the next functional test period because the overall '

operability of Isrge bore snubbers will not be affected.

j The proposed change adds snubber valve block testing as a method of

{

functionally testing large bore snubbers. A large maintenance effort is }

} required to' remove large bore snubbers because of their location. Piping, i conduit, and power supply cables are often located in close proximity to large j bore snubbers. This equipment must be cut and then reassembled when the entire snubber is removed to perform functional testing. Working in the areas

, where large bore snubbers are located is a radiological hazard. Most large i bore snubbers are located in high radiation areas. Testing large bore snubber

valve blocks would require the removal of only the valve block. This would .

1 greatly reduce personal exposure in addition to significantly reducing the l effort required to test large bore snubbers. -

i l Testing snubber valve blocks is an acceptable method of functionally testing

! snubbers. Technical Specifications require that bleed rate and lock-up be !

j. determined by a functional test. These parameters are based on fluid flow l l )

i i

Peg 2 3

- . through tha valve block. The ASME O&M Working Group on Inservice Performance Testing of Snubbers has recommended subcomponent (i.e., valve block) testing as an acceptable method of functionally t,esting snubbers. Examination and Performance Testing of Nuclear Power Plant Dynamic Restraints (snubbers).

O&M-4, Revision 1. Draft 4 dated 1-84 with changes from 12-84 meeting, Section 3.2.6.e states: "Where the physical size of the snubber, test equipment limitations or inaccessibility of location prevent inplace testing' and bench testing, the snubber subcomponents shall be examined and tested ~in accordance with approved procedures." Additionally, North Anna has an Interim Program for testing large bore snubbers that has been reviewed by the staff of NRC Region II (IR 50-338/83-29 and 50-339/83-29). The margin of plant safety is not decreased by functionally testing only snubber valve blocks because the overall operability of large bore snubbers will not be affected.

50.59 Safety Review Pursuant to 10 CFR 50.59, we have reviewed the proposed Technical Specification changes and have concluded that no unreviewed safety question exists since (1) the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased by these proposed changes; (ii) the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report is not being created by these proposed changes; (iii) the margin of safety as defined in the basis for any technical specification is not reduced by these proposed changes because the operability and performance of the snubbers has not been affected by these changes.

50.92 Significant Hazards Review The proposed changes do not pose a significant hazards consideration as defined in 10 CFR 50.92. The Commission has provided examples of changes that constitute no significant hazards consideration in Federal Register, Volume 48, page 14870. Example (ii) is a change that constitutes an additional

. limitation, restriction, or control not presently included in the technical specifications; for example, a more stringent surveillance requirement.

Example (vii) is a change to make a license conform to changes in the regu-lations, where the license change results in very minor changes to f acility operations clearly in keeping with the regulations. The proposed changes are similar to example (ii) in that more snubbers will initially undergo func-tional testing during each test period. The proposed changes are also similar to example (vii) in that they are in accordance with the guidance provided in Generic Letter 84-13 and a draft industry standard (ASME 0&M-4), and similar to the Technical Specifications requirements for another facility.

Based on these examples, it has been concluded that the proposed changes do not pose a significant hazards consideration.

015/NATS85

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9 ATTACHMENT 4 APPLICATION FEE 2 .

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' NUCLEAR REG C DC 20555 97003l 08/20/85 3301l $150.00I 03301 25 0097003 00000015000 .

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v t e Letter Sequence Other
TAC:56347, (Open) TAC:56350, (Open) TAC:59762, (Open) TAC:59763, (Open) TAC:59764, (Open) TAC:59765, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML20133G206
MONTHYEAR1985-10-04 [Table View]

ML20133G206

Text

SUBJECT:

Enclosures:

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