ML20092C173

Amends 70 & 98 to Licenses DPR-71 & DPR-62,respectively, Revising Tech Specs Re Mgt Organization Structure & Reporting Requirements

ML20092C173
Person / Time
Site:	Brunswick
Issue date:	05/25/1984
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Carolina Power & Light Co
Shared Package
ML20092C179	List: ML20092C173 ML20092C184
References
DPR-62-A-098, DPR-71-A-070 NUDOCS 8406210041
Download: ML20092C173 (80)

Text

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o UNITED STATES

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g NUCLEAR REGUL ATORY COMMISSION

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i CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. DPR-71 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Carolina Power & Light Conpany (the licensee) dated July 28, 1981, as supplemented December 10, 1982, and December 29, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Connission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, Facility Operating License No. DPR-71 is hereby amended by deleting paragraphs 2.I and 2.J.

The license is further amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) is hereby amended to read as follows:

(2) Technical Specifications 4

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 70, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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This license amendment is effective as of the date of its issuance.

j FOR THE NUCLEAR REGULATORY COMMISSION l

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Domenic B. Vassallo, Chief Operating Reactors Branch #2 i

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i Attachment-Changes to the l

Technical Specifications Date of Issuance: May 25, 1984 f

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ATTACHMENT TO LICENSE AMENOMENT NO. In FACILITY OPERATING LICENSE NO. DPR-71 I

i DOCKET NO 50-325 Revise the Appendix A Technical Specifications as follows:

Remove Insert XIV XIV r

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i, INDEK ADtINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY.............................................

6-1 6.2 ORGANIZATION 6.2.1 0FFSITE................................................

6-1 6.2.2 FACILITY STAFF.........................................

6-1 6.2.3 ONSITE NUCLEAR S AFETY GROUP Function...............................................

6-8 Re s pon s ib i l i t i e s.......................................

6-8 Authority..............................................

6-8 6.2.4 S H I FT TE CHN I CAL A DV IS0R................................

6-8 6.3 F ACILITY STAFF 0 0kLIF ICATIONS..............................

6-8 6.4 TRAINING...................................................

6-8 6.5 REVIEW AND AUDIT 6.5.1 N UCLEAR S AFETY R EV I EWERS...............................

6-9 6.5.2 SAFETY EVALUATIONS AND NUCLEAR REVIEW CONTROL Sa f e t y Eva lua tions.....................................

6-9 P rocedures. Tes t s. and Expe riment s.....................

6-10 Modifications..........................................

6-10 Operating License / Technical Specifications.............

6-10 6.5.3 PLANT NUCLEAR S AFETY C'MMITTEE (PNSC)

Function...............................................

6-11 Co mp o s i t i on............................................

6-11 Al t e rna t e s.............................................

6-11 Neeting Frequency......................................

6-11 Quorum.................................................

6-11 Ac t iv i t i es.............................................

6-12 1

Authority..............................................

6-13 necords................................................

6-13 BRUNSWICK - UNIT 1 XIV Amendeont No. 70 w.

IND EK ADMINISTRATIVE CONTROLS S ECTION PAGE 6.5.4 CORPORATE NUCLEAR SAFETt SECTION Function...............................................

6-13 Organization...........................................

6-13 Review.................................................

6-14 Records................................................

6-15 6.5.5 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM Fu n c t i o n...............................................

6-16 Audits.................................................

6-16 Re c o r d s................................................

6-17 Authority..............................................

6-17 Personne1..............................................

6-17 l

6.5.6 0UTSID E AG ENCE INSP ECTION AND AUDIT P ROGRAM............

6-18 6.6 R EPOR TABL E OCCU RR ENCE ACTI0N...............................

6-18 6.7 S A F ETY L IM I T V I OL ATI ON.....................................

6-18 6.8 P RO C ED U RES AND P R0G RAM S....................................

6-19 6.9 REPORTING REQUIREMENTS Routine Re ports and Re portable Occurrences.............

6-20 i

S t a r t u p Re po r t s........................................

6-20 An n u a l Re po r t s.........................................

6-21 Pe rsonnel Exposure and Monitoring Re port...............

6-21 hinual Radiological Environmental Operating Re port.....

6-22 Semiannual Radioactive Ef fluent Re lease Re po r t.........

6-23 Ho n t hly Op e ra t ing Re po r t s..............................

6-24 Re po r t a b l e Oc c u r r e nc e s.................................

6-25 Promp t No tification With Written Fo llowup..............

6-25 Vi i r t y Da y Wr i t t e n Re po r t s.............................

6-27 S p e c i a l Re po r t s........................................

6-28 BRUNSWICK - UNIT 1 XV Amendment No. 70

g IND EX ADMINISTRATIVE CONTROLS SECTION PAGE l

l 6.10 RECO RD R ETEN TI ON..........................................

6-28 6.11 RAD I ATION P ROTECTION P R0G RAM..............................

6-30 6.12 HIG H RAD I ATION AREA.......................................

6-30 6.13 0 FFSITE D OSE CALQJLATION MANUAL (ODCM)....................

6-31 6.14 P ROCESS CONTROL P ROGRAM (PCP).............................

6-31 6.15 MMOR CHANGES TO LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTI!MS..........................

6-32 l

BRUNSWIQt - UNIT 1 XVI Amendment No. 70 l

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g 6.0 ADfINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Manager - Brunswick Plant shall be responsible for overall l

f acility operation and shall delegate in writing the succession to this responsibility during his absence.

6.2 ORGANIZATION l

OFFSITE 6.2.1 The offsite organization for f acility management and technical support shall be as shown on Figure 6.2.1-1.

FACILITY STAFF 6.2.2 The f acility organization shall be as shown on Figures 6.2.2-1 and 6.2.2-2 and:

i s.

Each f acility on duty shif t shall be composed of at least the mintaus facility shif t crew composition shown in Table 6.2.2-1.

b.

At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor.

j, c.

When either reactor is in OPERATIONAL CONDITION 1, 2, or 3, at least one licensed Senior Reactor Operator shall be in the control room.

d.

An individual qualified to implement radiation protection procedures shall be onsite when fuel is in either reactor.*

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e.

All CORE ALTERATIONS shall be directly supervised by either a j

licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

L f.

A Fire Brigade of at least five seabers shall be maintained onsite at all times.*

The Fire Brigade shall not include the minimum shif t l

crew shown in Table 6.2.2-1 or any personnel required for other essential functions during a fire emergency.

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• The individual qualified to implement radiation protection procedures and Fire Brigade composition may,be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected absence provided immediate action is taken to fill the required positions.

BRUNSWICK - UNIT 1 6-1

. Amendment No. 70

-ADMINISTRATIVE CONTROLS FACILITY STAFF (Continued) g.

Administrative procedures shall be developed and implemented to limit

.l the working hours of facility staff who perform safety-related l

functions; e.g., senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

1hese procedures shall meet the working hour guidelines published by the Commission in Generic Letter No. 82-12.

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BRUNSWIQt - UNIT 1 6-2 hendment No. 70

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VICE PRESIDENT RRUNSWICK NUCLEAP PROJECT GENERAL MAN 4GCR BRUNSWICK PLANT I

M4 NAGER Pl.' ANT OPER ATIONS I

MANAGER OPERATIONS I

I OPERATIONS SUPERINTENDENT I

I PRINCIPAL ENGINEER SHN'T OPERATING PRINCIPAL ENGINEER PRINCIPAL ENGINEER OPERATIONS (UNIT 1)

SUPE R VIS OR{,,

opgMATIONS (UNIT 2)

OPERATIONS I

I SHIFT FIRE BRIGADE 4

4X SENIOR SPECIAll8T (EMERGENCY DINATION)

FIRE P OTECTION I

SHIFT FIRE BRIGADE SUPPORT GROUP LEGENO V Nember of Srleede Fire Chiefe verlee with Shift Orgenlaalien 44 One Speelellet le Aselened The Dettee of The Plant Fire Chief Amendment No.

GRUNSWICN STEAM &

i ELECTRIC PLANT PLANT FIRE PROTECTION ORGANt2ATION C4eohne Powee & Light Company Pigeee S.2.2-3

.PeSe No. 4 8 Amon<i-ant 70

TABLE 6.2.2-1 MINIMLM FACILITY SHIFT CREW COMPOSITION WITH UNIT 2 IN CONDITION 1, 2, OR 3 POSITION NIMBER OF INDIVIDUALS REQUIRED TO FILL POSi. TION CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 1

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1 WITH UNIT 2 IN CONDITION 4 OR 5 POSITION NIMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 lD) 1 I

SOS (a) g(b) g SR{a)

RO 3

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WITH UNIT 2 DE-FUELED l

POSITION NIMBER OF IN1/IVIDUALS REQUIRED TO FILL POSITION CONDIT?oNS 1, 2, & 3 CONDITIONS 4 & 5 1

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t%ne BkJSWICK - UNIT 1 6-6 Amendment No. 70

i TABLE 6.2.2-1 (Gntinued)

MINL4IM FACILITY SHIFT CREW COMPOSITION TABLE NOTATION SOS - Stif t Operating Supervisor with a Senior Reactor Operators License SRO - Individual with a Senior Reactor Operators License RO - Individual with a Reactor Operators License AD - Auxiliary Operator (non-licensed individual)

STA - Stif t Technical Advisor (a) Assumes each individual is' licensed on both plants.

(b) Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS.

i lhe Shif t Crew Composition may be less than the minimum requirements for a period of time not to exceed 2 hodrs in order to accommodate unexpected absence of on-duty shif t crew rembers provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2.2-1.

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- BRUNSWIQt - UNIT 1 6-7 Amendment No. 70

_ _ = - _

ADMINISTRATIVE CONTROLS 6.2.3 ONSITE NUCLEAR SAFETY (0NS)

FUNCTION 6.2.3.1 The ONS Unit shall function to examine facility operating characterisitics, NRC issues, industry advisories, and other sources which may indicate areas for improving facility safety.

RESPONSIBILITIES 6.2.3.2 The ONS Unit shall be responsible for maintaining surveillance of facility activities to provide independent ve. ification* that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.3 The ONS Unit shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving facility safety to the Manager-Corporate Nuclear Safety Section.

6.2.4 SHIFT TECHNICAL ADVISOR 6.2.4.1 The Shif t Technical Advisor shall serve in an advisory capacity to the Shif t Operating Supervisor on matters pertcining to the engineering aspects assuring safe operation of the unit.

6.3 FACILITY STAFF QUALIFICATION 6.3.1 Each member of the facility staf f defined in Figure 6.2.2-1 shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Manager - Environmental & Radiation Control who l

shall meet or exceed the qualifications of-Regulatory Guide 1.8, September -1975 and (2) the Shif t 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 during transients and l

accidents.

6.4' TRAINING 6.4.1 A retraining and replacement training' program for the facility staff shall be maintained under the direction of the Director - Training and shall-l meet or exceed the -requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

I 6.4.2 A training program for the Fire Brigade _ shall be maintained under the direction of the Manager-Operations and shall meet or exceed the requirements l_

of Section 27 of the NFPA Code-1975.

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• Not responsible for sign-off function.

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-BRUNSWICK - UNIT I 6-8 Amendment No. 70.

AININISTRATIVE CONTROLS 6.5 REVIEW AND AUDIT 6.5.1 NUCLEAR S AFETY REVIEWERS 6.5.1.1 Individuals shall be designated / approved by the General Manager -

Brunswick Plant for performing nuclear safety reviews.

6.5.1.2 Individuals designated under Specification 6.5.1.1 above shall have an academic degree in an engineering or related field or equivalent and two years related experience.

6.5.1.3 A list shall be maintained of individuals qualified to perform nuclear safety reviews, including additional individuals whose expertise may be necessary during the reviews to assure that the reviewers collectively possess the background and qualfications in the disciplines necessary and important to q

the specific review.

6.5.1.4 The list specified in Specification 6.5.1.3 above shall include the disciplines for which each individual is qualified.

6.5.1.5 For those cases where interdisciplinary reviews are required, as many individuals as necessary shall be used to perform the nuclear review function.

6.5.1.6 One of the two nuclear safety reviewers shall be an individual other j

than the original preparer or the individual approving the action.

l 6.5.2 SAFETY EVALUATIONS AND NUCLEAR REVIEW CONTROL S AFETY EVALUATIONS 6.5.2.1 A safety evaluation shall be prepared for each of the following:

a.

Changes to procedures required by Specification 6.8, or changes to other procedures that affect nuclear safety.

b.

Proposed tests or experiments that affect nuclear safety.

c.

Proposed modifications to plant systems or equipment that affect nuclear safety.

d.

Proposed changes to the Technical Specifications.

e.

Proposed changes to the Operating License.

6.5.2.2 Two nuclear safety reviews of the item and safety evaluation (s) prepared in accordance with Specification 6.5.2.1 above shall be performed prior to approval and implementation.

6.5.2.3 The ites and associated safety evaluation (s) shall be examined in order to determine whether an interdisciplinary review is required in accordance with Specification 6.5.1.5 above.

BRUNSWICK - UNIT 1 6-9 Amendment No. 70

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A011NISTRATIVE CONTROLS i

l PROCEDURES, TESTS, AND EXPERIMENTS 1

6.5.2.4 The safety evaluation prepared in accordance with I

Specifications 6.5.2.1.a and 6.5.2.1.b above shall include a written determination, with basis, of whether or not the procedures, proposed tests and experiments, and changes thereto constitute an unreviewed safety question as defined in 10 CFR 50.59, or whether they involve a change to the Technical Specifications.

6.5.2.5 Following the nuclear safety review, the procedures required by Specification 6.8, other procedures that affect nuclear safety, proposed tests or experiments, and changes thereto (other than editorial or typographical) which have been determined to not involve an unreviewed safety question as defined in 10 CFR 50.59 or change to the Technical Specifications shall be approved prior to implementation by the General Manager - Brunswick Plant or his previously designated alternate.

MODIFICATIONS 6.5.2.6 The safety evaluation prepared in accordance with Specification 6.5.2.1.c above shall include a written determination, with basis, of whether or not the proposed modification is a change in the f acility as described in the safety analysis report, involves a change to the Technical Specifications, or constitutes an unreviewed safety question as defined in 10 CFR 50.59.

6.5.2.7 Following the nuclear safety review, proposed modifications which have been determined to not involve an unreviewed safety question as defined in 10 CFR 50.59 or a change to the Technical Specifications shall be approved by the General Manager - Brunswick Plant or his previously designated alternate.

OPERATING LICENSE / TECHNICAL SPECIFICATIONS 6.5.2.8 The safety evaluation prepared in accordance with Specifications 6.5.2.1.d and 6.5.2.1.e above shall include a written preliminary determination, with basis, of whether or not the proposed Operating License / Technical Specification change (s) is a change in the facility as described in the safety analysis report.

6.5.2.9 Following the nuclear safety review of the safety evaluation prepared in accordance with Specifications 6.5.2.1.d and 6.5.2.1.e above and the associated proposed action, the request shall be:

a.

Reviewed by the Plant Nuclear Safety Committee ia accordance with Specification 6.5.3.8.

b.

Reviewed by the Corporate Nuclear Safety Section in accordance with Specification 6.5.4.9.

BRUNSWICK - UNIT 1 6-10 Amendment No. 70

AININISTRATIVE CONTROLS i

6.5.3 PLANT NUCLEAR S AFETY COMMITTEE (PNSC)

FUNCTION 6.5.3.1 As an ef fective means for the regular review, overview, evaluation, and maintenance of plant. operational safety, a Plant Nuclear Safety Committee (PNSC) shall be established.

6.5.3.2 The PNSC shall function through the utilization of sd) committees, audits, investigations, reports, and/or performance of reviews as a group.

COMPOSITION 6.5.3.3 The PNSC shall be composed of the:

Chairman:

General Manager - Brunswick Plant

• Member:

Manager - Plant Operations Member:

Manager - Technical & Administrative Support Member:

Manager - Technical Support Member:

Manager - Operations Member:

Manager - Maintenance Member:

Manager - Environmental & Radiation Control Member:

Assistant to Plant General Manager Member:

Director - QA/QC Member:

Director - Regulatory Compliance Member:

Director - Administrative Support ALTERNATES 6.5.3.4 All alternate members shall be appointed in writing by the PNSC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as members at any one time.

6. 5. 3. 5 All alternates, shall as a minimum, meet equivalent qualification criteria as specified for professional-technical personnel in Section 4.4 of ANSI N18.1-1971.

MEETING FREQUENCY 6.5.3.6 - The PNSC shall meet at least once per calendar month and as convened by the PNSC Chairman or his designated alternate.

Q UORU4 6.5.3.7 The minimum quorum of the PNSC necessary for the performance of the PNSC activities of the Technical Specifications shall consist of the PNSC Chairman or his designated alternate and five members including alternates. No more them two alternates shall be counted toward meeting the minimum quorum require ent.

• Or designated alternate.

l BRUNSWICK -- UNIT l' 6-11 Amendment No. 70

ADMINISTRATIVE CONTROLS ACTIVITIES I

6.5.3.8 The PNSC activities shall include the following:

a.

Review of all procedures required by Specification 6.8 and changes thereto (and any other procedures and changes thereto), any of which constitute an unreviewed safety question or involve a change to the Technical Specifications, prior to implementation.

b.

Review of all proposed tests or experiments that constitute an unreviewed safety question or involve a change to the Technical Specifications, prior to implementation.

c.

Review of all proposed modifications that constitute an unreviewed safety question or involve a change to the Technical Specifications, prior to implementation.

d.

Review of all proposed changes to the Technical Specifications or Operating License, prior to implementation.

Review of reports on violations of Technical Specifications including e.

reports covering evaluation and recommendations to prevent recurrence to the Vice President - Brunswick Nuclear Project and to the Manager -

Corporate Nuclear Safety Section.

f.

Performance of special reviews, investigations (or analyses), and reports thereon as requested by the Manager - Corporate Nuclear Safety i

Section.

g.

Review of events requiring 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written notification to the Commission.

i h.

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

t 1.

Annual review of the Security Plan.

J.

Annual review of the Emergency Plan.

k.

Review of accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and dispositice of the corrective action to prevent i

recurrence and the forwarding of these reports to the Vice President -

Brunswick Nuclear Project and the Manager -- Corporate Nuclear Safety 1

Section.

1.

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

BRUNSWICK - UNIT 1 6 Amendment No. ' 70 w

AININISTRATIVE CONTROLS AUTHORITY

6. 5. 3. 9 If there is a disagreement between recommendations of a majority of the PNSC and the actions contemplated by the General Manager - Brunswick Plant, the I

PNSC shall provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President -

Brunswick Nuclear Project and the Vice President - Corporate Nuclear Safety and Re search. The course determined by the General Manager - Brunswick Plant to be the most conservative shall be followed.

RECORDS 6.5.3.10 The PNSC shall maintain written minutes of each PNSC meeting that, at a minimum, document the results of all PNSC activities performed under the provisions of these Technical Specifications.. Copies shall be provided to the Vice President - Brunswick Nuclear Project and the Manager - Corporate Nuclear l

Safety Section.

6.5.4 CORPORATE NUCLEAR SAFETY SECTION l

FUNCTION 6.5.4.1 The Corporate Nuclear Safety Section (CNSS) of the Corporate Nuclear Safety & Research Department shall function to provide independent review of significant plant changes, tests, and procedures; verify that REPORTABLE OCCURRENCES are investigated in a timely. manner and corrected in a mannar that reduces the probability of recurrence of such events; and detect trends that may not be apparent to a day-to-day observer.

ORGANIZATION 6.5.4.2 The individuals assigned responsibility for independent reviews shall be specified in technical disciplines. These individuals shall collectively have the experience and competence required to review activities in the following areas:

a.

nuclear power plant operations b.

nuclear engine ering c.

chemistry and radiochemistry d.

metallurgy e.

non-destructive testing f.

instrumentation and control g.

radiological safety h.

mechanical and electrical engineering i

1.

administrative controls BRUNSWICK - UNIT 1 6-13 Amendment No. 70

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AIMINISTRATIVE CONTROLS ORGANIZATION (Continued) j.

seismic and environmental k.

quality assurance practices l

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Other appropriate fields associated with the unique characteristics l

of the nuclear power plant.

l 6.5.4.3 The Manager - Corporate Nuclear Safety Section shall have an academic degree in an engineering or related field and, in addition, shall have a minimum of ten years related experience, of which a minimum of five years shall be in the operation and/or design of nuclear power plants.

6.5.4.4 The independent safety review program reviewers shall have an academic degree in an engineering or related field or equivalent and, in addition, shall have a minimum of five years related experience.

6.5.4.5 An individual may possess competence in more than one specialty

area, if sufficient expertise is not available within the Corporate Nuclear Safety Section, competent individuals from other Carolina Power & Light Company organizations or outside consultants shall be utilized in performing independent reviews and investigations.

6.5.4.6 At least three individuals, qualified as discussed in 6.5.4.4 above shall review each item submitted under the requirements of Section 6.5.4.9.

i 6.5.4.7 Independent safety reviews shall be performed by individuals not directly involved with the activity under review or responsible for the activity under review.

6.5.4.8 The Corporate Nuclear Safety Section independent safety review program shall be conducted in accordance with written, approved procedures.

f REVIEW 4

6.5.4.9 The Corporate Nuclear Safety Section shall perform reviews of the following:

a.

The safety evaluations for 1) changes to procedures required by Specification 6.8, 2) modifications of equipment or systems, and

3) tests or experiments that constitute a change to the safety analysis report to verify that such actions did not constitute an unreviewed safety question or involve a change to the Technical Specifications.

Implementation may proceed prior to completion of this review.

b.

Proposed changes to procedures required by Specification 6.8, and proposed modifications that constitute an unreviewed safety question as defined in 10 CFR 50.59 or a change co the Technical Specifications, prior to implementation.

BRUNSWICK - UNIT 1 -

6-14

' Amendment No. 70

AD11NISTRATIVE CONTROLS REVIEW (Continued)

Proposed tests or experiments that involve an unreviewed safety c.

question as defined in 10 CFR 50.59 or a change to the Technical Specifications, prior to implementation.

d.

Proposed changes to the Technical Specifications and Operating License.

e.

Violations, deviations, and events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written notification to the Commission, such as:

1.

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

I 2.

Significant operating dbnormalities or deviations from normal and expected performance of plant safety-related structures, systems, or components.

f.

Reports and minutes of the PNSC.

g.

Any other matter involving safe operation of the nuclear power plant that the Manager - Corporate Nuclear Safety Section deems appropriate f or consideration or which is referred to the Manager - Corporate Nuclear Safety Section by the on-site operating organization or other f unctional organizational units within Carolina Power & Light Company.

6.5.4.10 Review of items considered under 6.5.4.9(e) through (g) above shall include the results of any investigations made and the recommendations resulting f rom these investigations to prevent or reduce the probability of recurrence of the event.

RECORDS 6.5.4.11 Records of Corporate Nuclear Safety Section reviews, including recommendations and concerns, shall be prepared and distributed as indicated below:

a.

Copies of documented reviews shall be retained in the CNSS files.

b.

Recommendations and concerns shall be submitted to the General Manager - Brunswick Nuclear Project and Vice President - Brunswick '

Plant, within 14 days of completion of the review.

c.

A summation of Corporate Nuclear Safety recommendations and concerns shall be submitted to the Chairman / President and. Chief Executive Of ficer; Executive Vice President - Power Supply and Engineering and Construction; Vice President - Corporate Nuclear Safety and Research; Vice President - Brunswick Nuclear Project; General Manager. -

Brunswick Plant; and others, appropriate, on at least a bi-monthly frequency.

1 BRUNSWICK.- UNIT f1 6-15 haendment No. 70

AIMINISTRATIVE CONTROLS 6.5.5 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM FUNCTION 6 5.5.1 The Performanco Evaluation Unit (PEU) of the Corporate Quality Assurance Department shall function to perform audits of facility activities specified in Specification 6.5.5.2.

AUDITS 6.5.5.2 Audits of f acility activities shall be performed by the PEU.

These audits shall encompass:

a.

The conformance of f acility operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months.

b.

The training and qualifications of the entire f acility staff at least once per 12 months.

c.

The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems, or methods of operation that affect nuclear safety at least once per 6 months.

d.

The verification of compliance and implementation of the requirements of the Quality Assurance Program to meet the criteria of Appendix "B",

10 CFR 50, at least once per 24 months.

e.

The Emergency Plan and implementing procedures at least once per 12 months.

f.

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

g.

The Facility Fire Protection Program and implementing procedures at least once per 12 months.

h.

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

i.

The OFFSITE DOSE CALCULATIONAL MANUAL and implementing procedures at l

least once per 24 months.

j.

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

l 1

t l

l BRUNSWICK - UNIT 1 6-16 Amendment No. 70 l

\\

t i

ADMINISTRATIVE CONTROLS I

AUDITS (Continued) i k.

The perforpance 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 1

once per 12 months.

l 1.

Any other area of f acility operation considered appropriate by the i

l Manager - Quality Assurance Services Section.

6.5.5.3 Personnel performing the quality assurance audits shall have access to the plant operating records.

l RECORDS 6.5.5.4 Records of audits shall be prepared and retained.

6.5.5.5 Audit reports encompassed by 6.5.5.2 above shall be prepared, approved by the Principal QA Specialist - Performance Evaluation Unit, and forwarded to the Executive Vice President - Power Supply and Engineering and Construction; Vice President - Brunswick Nuclear Project; Vice President -

Corporate Nuclear Safety and Research; General Manager - Brunswick Plant; and others, as appropriate, within 30 days af ter completion of the audit.

AUTHORITY i

6.5.5.6 The Manager - Quality Assurance Services Section under the Manager -

Corporate Quality Assurance shall be responsible for the following-I t

a.

The administering of the Corporate Quality Assurance Audit Program.

l i

b.

The approval of the individual (s) selected to conduct quality assurance audits.

3 PERSONNEL 6.5.5.7 Audit personnel shall be independent of the area audited.

J 6.5.5.8 Selection of personnel for auditing assignments shall be based on 6

experience or training that establishes that their qualifications are commensurate with the complexity or special nature of the activities to be audited.

In selecting audit personnel, consideration shall be given to special abilities, specialized technical training, prior pertinent experience. l personal characteristics, and education.

i 6.5.5.9 Qualified outside consultants or other Ladividuals independent from l

those personnel directly involved in plant operation shall be used to augment i the audit teams when necessary.

I BRUNSWICK - UNIT 1 6-17 Amendment No. 70 1

_~

~.

i AIMINISTRATIVE CONTROLS i

6.5.6 0UISIDE AGENCY INSPECTION AND AUDIT PROGRAM l

6.5.6.1 An independent f tre protection and loss prevention inspection and l

audit shall be performed at least once per 12 months utilizing either qualified off site licensee personnel or an outside fire protection firm.

6.5.6.2 An inspection and audit of the fire protection and loss prevention l

program shall be performed by an outside igualified fire consultant at intervals no greater than 36 months.

6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 he 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission shall be reviewed by the General Manager - Brunswick Plant I

and submitted to the Manager - Corporate Nuclear Safety Section and the Vice President - Brunswick Nuclear Project.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The f ollowing actions shall be taken in the event a Safety Limit is vio' lated :

a.

he f acility shall be placed in at least HOT SHITfDOWN within two hours.

{,

b.

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

The Vice President -

Brunswick Nuclear Project and the Manager - Corporate Nuclear Safety Section shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l c.

A Safety Limit Violation Report shall be prepared.

The report shall be reviewed by the General Manager - Brunswick Plant. This report l

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 Vice President - Brunswick Nuclear Project, and the q

Manager - Corporate Nuclear Safety Section within 14 days of the violation.

BRUNSWICK - UNIT L 6-18

' Amendment No. 70

ADMINISTRATIVE CONTROLS j

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, November 1972.

b.

Refueling operations.

c.

Surveillance and test activities of safety related equipment.

d.

Security Plan implementation.

4 e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

g.

OFFSITE DOSE CALCULATION MANUAL implementation.

h.

PROCESS CONTROL PROGRAM implementation.

al 1.

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.

t 6.8.2 Temporary changes to procedures of Specification 6.8.1 above, any other procedures that affect nuclear safety, and proposed tests or experiments may be made provided:

a.

The intent of the original procedure, proposed test or experiment is not altered.

b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator License on the j

unit affected.

c.

The change is documented, reviewed pursuant to Specifications 6.5.2.1 and 6.5.2.2 and approved by the General Manager - Brunswick Plant or his previously designated alternate within 14 days of implementation.

6.8.3 - The following ' programs shall be established, implemented, and maintained :

I a..

Primary Coolant Sources Outside Containment Afrogram 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 program shall. include the following:

BRUNSWICK - UNIT 1.

6-19 Amendment No. 70'

ADMINISTRATIVE CONTROLS i

PROCEDURES AND PROGRAMS (Gantinued) 1._

Preventive maintenance and periodic visual inspection requirements, and 2.

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

b.

In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

Elis program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

c.

Post-Accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the following:

i 1.

Itaining o f personnel, 2.

Procedures for sampling and analysis, and 3.

Provisions for maintenance of sampling and analysis equipment.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office unless otherwise noted.

l STARTUP REPORTS 6.9.1.1 A summary report of plant startup and power. escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) -installation of fuel that has a different design or has been manufactured by a different fuel.

supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

BRUNSWICK'- UNIT 1 6-20 Amendment No. 70

l I

ADMINISTRATIVC CONTROLS STARTUP REPORTS ((butinued) 6.9.1.2 lie startup report shall address each of the tests identified in the l

FSAR and shall include a description of the measured values of the operating l

conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, (3) 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events, i.e.,

initial criticality, completion of startup test program, and resumption or commencement of commercial power operation, supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS./

I 6.9.1.4 hanual reports covering the activities of the unit as described below during the previous calendar year shall be submitted prior to March 1 of each year.

1he initial report shall be submitted prior to March 1 of the year following initial criticality.

~

2 l

PERSONNEL EKPOSURE AND MONITORING REPORTf o

a 6.9.1.5 Reports required on an annual basis shall include a tabulation of the number of s tation, utility, and other personnel, including contractors, receiving exposures greater than 100 aren/yr exposure according to work and job functiona,jnd their associated man-rem t, e.g., reactor o perations and surveillance, inservice inspection routine maintenance,'special maintenance (deset1 be maintenance), waste processing, and refueling.

1he dose assignments to various duty functions may be estimated, based on pocket dosimeter, ILD, or film badge measurements.

Small exposures totalling less than 20 percent of the individual total dose need not be accounted for.

In the aggregate, at least 80 percent of the total whole body dose received from external sources shall be assigned to specific major work functions.

4

- 1/ A single submittal may be made for a multiple unit station.

1he submittal-should combine those sections that are common to all units at the station.

2/ ~ 1 tis tabulation supplements the requirements of 520.407 of 10 CFR Part 20.-

BRUNSWICK - UNIT 1 6-21 heen'dment No. 70 y-m=-

w y-y r + + - -

ADMINISTRATIVE CONTROLS ANNUAL RADIOLCCICAL ENVIRONMENTAL OPERATING REPORh 4

6.9.1.6 Routine radiological environmental operating reports covering the operation of the facility during the previous calendar year shall be submitted prior to May 1 of each year.

6.9.1.7 The Annual Radiological Environmental Operating Re port shall include i

the following:

{

a.

Smamaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with pre-operational studies, with operational controls (as appropriate), and with previous environmental surveillance reports, and an assessment of the observed impact of the plant operation on the environment.

4 b.

Results of land uses censuses required by Specificaton 3.12.2.

c.

Results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the OFFSITE DOSE CALCULATION MANUAL, as well as summarized and tabulated results of these analyses and measurements in the format of Table 3 in the Radiological Assessment Branch Technical Position, Revisien 1, i

November 1979.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing j

resul ts.

The missing data should be submitted as soon as possible in a supplementary report.

d.

A summary description of the radiological environmental monitoring program.

least two legible mapsb./ covering all sampling locations keyed to e.

At a table giving distances and directions from the centerline of one reactor.

f.

The results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3.

i 3/ A single submittal may be made for a multiple unit station.

--%7 One map shall cover stations near the SITE BOUNDARY; a second map shall include the more distant stations.

l BRUNSWICK - UNIT 1 6-22~

- Amendment No. 70 h

,---v w,

c c,

'-r,

-e 9

ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (Continued) g.

Discussion of all deviations from the sampling schedule of Table 3.12.1-1.

h.

Discussion of all analyses in which the LLD required by Table 4.12.1-1 was not achievable.

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE. REPORT 6.9.1.8 Routine radioactive effluent release reports covering the operation l

of the facility during the previous 6 months of operation shall be submitted j

within the time periods specified in Specifications 669.1.9 and 6.9.1.10 below.

6.9.1.9 The portion of the Semiannual Radioactive Effluent Release Reports to be submitted within 60 days af ter January 1 and July 1 of each year shall include the following:

I A summary of the quantities of radioactive liquid and gaseous a.

i effluents and solid waste released from the facility as outlined in j

Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactivity Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly j

basis similar to the format of Appendix B thereof.

b.

The information specified below for each class of solid waste (as defined by 10 CFR Part 61, when implemented) shipped offsite during the report period:

1.

Container volume, 2.

Total curie quantity (specify whether determined by measurement l

or estimate),

3.

Principal radionuclides (specify whether determined by measurement or estimate),

4.

Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

5.

Type of container (e.g., LSA, Type A, Type B, Large Quantity),

and 5/ A single submittal may be made for a multiple unit station. The submittal should. combine those sections th' t are common to all units at the station.

a BRUNSWICK - UNIT 1 6-23 Amendment No. 70

4 i

ADiINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) 6.

Solidification agent or absorbent (e.g., cement, urea l

f ormaldehyde).

c.

A list of description of unplanned releases f rom the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

d.

Any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) or the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calcul' cions and/or a

environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

6.9.1.10 The portion of the Semiannual Radioactive Ef fluent Release Report to be submitted within 90 days af ter January 1 of each year shall include the following:

a.

An annual summary of hourly setorological data collected over the previous calendar year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint f requency distriptions of wind speed, wind direction, and atmospheric stability b.

An assessment of the radiation doses due to the radioactive liquid and gaseous effluents released f rom the station during the previous calendar year.

i t

j MONTHLY OPERATING REPORTS 6.9.1.11 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to main steam system safety / relief valves, shall be submitted on a monthly basis to the Director, Of fice of Resource Management, U. S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Administrator of the Regional Of fice no later than the 15th of each month following the calendar month covered by the report.

6/ In lieu of submission with the Semiannual Radioactive Ef fluent Release

~

Report, the licensee has the option of retaining this summary of required meteorological data in'a file that shall be provided to the NRC upon request.

1 BRUNSWICK - UNIT 1 6-24 Amendment No. 70

ADMINISTRATIVE CONTROLS i

REPORTABLE OCQlRRENCES 6.9.1.12 1he REPORTABLE OCCURRENCES of Specifications 6.9.1.13 and 6.9.1.14 below, including corrective actions and seasures to prevent recurrence, shall be reported to the NRC. Sapplemental reports may be required to fully

,'~

describe final resolution of occurrence.

In case of corrected or supplemental a

reports, a licensee event report shall be completed and reference shall be l

made to the original report date.

l PR0pt'T NOTIFICATION WITH WRIT'EN FOLLGiUP I

6.9.1.13 he types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed.by telegraph, asilgram, or facsimile transmission l

I to the Regional Administrator of the. Regional Of fice, or his designate no l

l later than the first working day following the event, with a written followup l

report within 14 days. 1he written followup report shall include, as a minimum, a completed copy of the licensee event report form.

Information 4

provided on the licensee event report form shall be supplemented, as needed, i

by additional narrative material to provide complete explanation of the ciretastances surrounding the event.

4 a.

Philure of the reactor protection system or other systems subject to j

limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint j

specified as the limiting safety system setting in the Technical i

Specifications or failure to complete the required protective function.

No te:

Instrsment drif t discovered as a result of testing need not be reported under this item (but see 6.9.1.13.e, 6.9.1.13.f, and 6.9.1.14.a below).

f b.

Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting

{

condition for operation established in the Inchnical Specifications.

l Note :

If -specified action is taken when a system is found to be orerating between the most conservative and least conservative aspects of a limiting condition for operation i

listed in the Inchnical $ecifications, the limiting condition for operation is not considered to have been violated and no report need be submitted under this section 1

(but see 6.9.1.14.b below).

I l'

i l

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BRUN5i1Qt - UNIT 1 6-25 Amendment No. 70

ADMINISTRATIVE CONTROLS i

PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP ((bntinued) c.

Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or prLaary containment.

2 i

Note:

Leakage of valve packing or gaskets within the limits for identified leakage set forth in 3achnical Specifications j

need not be reported under this section.

d.

Reactivity anomalies involving disagreement with predicted value of reactivity balance under steady-state conditions during power operation greater than or equal to 1% ak/k; a calculaced reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% ak/k; or occurrence of any unplanned criticality.

e.

Emilure or malfunction of one or more components that prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.

f.

Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.

Note :

For 6.9.1.13.e and 6.9.1.13.f, reduced redundancy that does not result in loss of system function need not be reported under this section (but see 6.9.1.14.b and 6.9.1.14.c below).

+

g.

02nditions arising from natural or man-made events that, as a direct result of the event, require plant shutdown, operation of safety systems, or other protective measures required by Technical Specifications.

4 i

h.

Berors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the Thchnical Specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

i.

Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a i

manner less conservative than assumed in the accident analyses in the safety analysis report or Ibchnical Specifications bases; or discovery i

BRUNSWICK - UNIT 1 6-26 Amendment No. 70

1 l

g l

ADMINISTRATIVE CONTROLS i

PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP ((bntinued) during plant life of conditions not specifically considered in the safety analysis report or Thchnical Specifications that require j

remedial action or corrective seasures to prevent the existence or i

development of an unsafe condition.

Note :

This item is intended to provide for reporting of potentially generic problems.

j.

Failure or malfunction of the main steam system safety / relief valves.

l THIRTY DAY WRITTEN REPORTS 6.9.1.14 1se types of events listed below shall be the subject of written reports to the Regional Administrator of the Regional Of fice within thirty l

days of occurrence of the event.

The written report shall include, as a minimum, a completed copy of the licensee event report form.

Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

a.

Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the Technical Specifications but which do not _ prevent the fulfillment of the functional requirements of affected systems (but see 6.9.1.13.a and 6.9.1.13.b above).

l b.

Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation (but see 6.9.1.13.b above).

l Note:

Routine surveillance testing, instrument calibration, or preventive maintenance that require configurations described in 6.9.1.14.a and 6.9.1.14.b above need not be reported except where test results themselves reveal a degraded mode as described above.

c.

Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems (but see 6.9.1.13.f above).

l d.

Abnormal degradation of systems other than *. hose specified in 6.9.1.13.c above designed to contain radioactive material resulting from the fission process.

No te:

Sealed ~ sources or calibration sources are not included under this item. Leakage of valve packing or gaskets within the iheits for identified leakage set forth in Technical Specifications need not be reported under this iten.

BRUNSWIQt - UNIT -1 6-27 Amendment No. 70

ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.

a.

Inoperable Saianic Monitoring Instrumentation, Specification 3.3.5.1.

b.

Seismic event analysis, Specification 4.3.5.1.2.

c.

Reactor coolant specific activity analysis, Specification 3.4.5.

d.

Fire detection instrumentation, Specification 3.3.5.7.

e.

Eire suppression systems, Specifications 3.7.7.1, 3.7.7.2, 3.7.7.3, and 3.7.7.5.

f.

ECCS actuation, Specifications 3.5.3.1 and 3.5.3.2.

g.

Fire barrier penetration, Specification 3.7.8.

h.

Liquid Ef fluents Dose, Specification 3.11.1.2.

i.

Liquid Radwaste Treatment, Specification 3.11.1.3.

l j.

Dose - Noble Gases, Specification 3.11.2.2.

k.

Dose - Iodine-131, Iodine-133, Tritum, and Radionuclides in Particulate Form, Specification 3411.2.3.

1.

Gaseous Radwaste Treatment, Specification 3.11.2.4.

m.

Ventilation Exhaust Treatment, Specification 3.1'1.2.5.

n.

Total Dose, Specification 3.11.4.

o.

Monitoring Program, Specification 3.12.1.b.

6.10 RECORD RETENTION Eacility records shall be retained in accordance with ANSI-N45.2.9-1974.

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

a.

Records and logs of f acility 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.

BRUNSWICK - UNIT 1 6-28 Amendment No. 70

ADMINISTRATIVE CONTROLS RECORDS RETENTION (03ntinued) c.

All REPORTABLE OCCURRENCES 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 and fission detector leak tests and results.

h.

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

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

I a.

Records and drawing changes reflecting f acility 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 f acility radiation and contamination surveys.

d.

Records or radiation exposure for all individuals entering radiation control areas.

i e.-

Records of gaseous and liquid radioactive material released co the environs.

f.

Records of transient or operational cycles for those facility components identified in Table 5.7.1-1.

g.

Records of reactor tests and experiments.

h.

Records of training and qualification for current members of the plant staff.

1.

Records of inservice inspections performed pursuant to these.

2-Technical Specifications.

j.

Racords of Quality Assurance activities required by. che QA Manual.

k.

Records of l reviews performed - for changes made to procedures or equipment or reviews of tests and experheents pursuant to

'LO CFR 50.59.

BRUNSWICK - UNIT 1 6-29 Amendment No. 70 L__

l AD(INISTRATIVE dONTROLS RECORDS RETENTION (Continued) i 1.

Records of the service lives of all hydraulic and mechanical snubbers listed in Table 3.7.5-1 including the data at which the service life commances and associated installation and maintenance records.

m.

Records of analyses required by the radiological environmental monitoring program.

n.

Records of (1) meetings of the PNSC, (2) meetings of the previous off-site review organization, the Company Nuclear Safety Committee (CNSC), (3) the independent reviews performed by the Corporate Nuclear Safety Section, and (4) the independent reviews performed by j

the Corporate Quality Assurance Audit Program, Performance Evaluation Unit.

6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " Control Device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is 1000 ares /hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*.

Any l

individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b.

A radiation monitoring device which continuously integrates the i

radiation dose race in the area and alarms when a preset integrated 4

dose is received.

Entry into such areas with this monitoring device may be made af ter the dose rate levels in the area have been l

established and personnel have been made knowledgeable of them.

c.

An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation i

surveillance at the f requency specified by the f acility Health Physicist in the Radiation Work Permit. -

j Realth Physics personnel or personnel _ escorted by Health Physics personnel l

l shall be exempt from the RWP issuance.*equirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.

BRUNSWICK - UNIT 1 6-30 Amendment No. 70

)

I ADMINISTRATIVE CONTROLS HIGH RADIATION AREA (Continued) 6.12.2 The requirements of 6.12.1 above shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr.

In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Operations Shif t Foreman on duty and/or the Radiation Control Supervisor.

6.13 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.13.1 The OFFSITE CALCULATION MANUAL (ODCM) shall be approved by the i

Commission prior to implementation.

6.13.2 Licensee initiated changes to the ODCM:

a.

Shall be submitted to the Commission in the Semiannual Radioactive Ef fluent Release Report for the period in which the change (s) was made effective.

This submittal shall contain:

1.

Sufficiently detailed information to totally support rationale without benefit of additional of supplemental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change (s);

2.

A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations;

and, 3.

Documentation of the f act that the change has been reviewed and found acceptable by the PNSC.

b.

Shall become effective upon review and acceptance by the PNSC.

6.14 PROCESS CONTROL PROGRAM (PCP) 6.14.1 The PROCESS CONTROL PROGRAM (PCP) shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the PCP:

a.

Shall be submitted to the Commission in the Semiannual Radioactive Ef fluent Release Report for the period in which the change (s) was made. This submittal shall contain:

1.

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

'IC:;52ICA - UNIT 1 6-31 Amendment No. 70

l. ~ -.

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AININISTRATIVE CONTRutS PROCESS CONTROL PROGRAM (PCP) (Continued) 2.

A determination that the change did not reduce the overall conformance of the solidification waste product to existing criteria for solid wastes; and 3.

Docusentation of the f act that the change has been reviewed and found acceptable by the PNSC.

b.

Shall become ef fective upon review and acceptance by the PNSC.

6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID I4ASTE TREATMENT SYSTEMS 3

6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous, a.nd4olid) :

,,i a.

Shall be reported to the' Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the PNSC. The discussion of each change shall contain:

1.

A cummary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59; 2.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

/

/

3.

. A detailed description of the equipment, components, and processes involved and the' interfaces with other plant systees; 4.

An evaluation of the change that shows the predicted release of t

radioactive materials in'tiquid and gaseous effluents and/or quantity of solid waste that differ f rom those previously predicted in the license application and amendments thereto; 5.

An evaluation of the change that shows the expected maximum exposure to an individual in the UNRESTRICTED AREA and to the general population that differ f rom those previously estimated in the license application and amendments thereto; 6.

A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the s

actual releases for the period prior to when the changes are to be made; 2

'/ /

1/. Licensees may choose to submit the inforpat' ion called for in this Specification as part of the annual FSAR update.

l 9

Amendesnt No. 70 3RUNSwim - UNIT 1 if 6-32 "

I L

1

l l

AININISTRATIVE CONTROLS MMOR CHANCES TO LIQUID, CASEOUS, AND SOLID WASTE TREATMENT SYSTEMS

( Co ntinued) 7.

An estimate of the exposure to plant operating personnel as a result of the change; and 8.

Doctmentation of the f act that the change was reviewed and found acceptable to the PNSC.

b.

Shall become effective upon review and acuptance by the PNSC.

I BRUNSWICK - UNIT 1 6-33 Amendment No. 70

,4fp no,Iof v

UNITED STATES

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g NUCLEAR REGULATORY COMMISSION

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.. t j WASHING TON, D. C. 20555

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CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 98 License No. DPR-62 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Carolina Power & Light Company (the licensee) dated July 28, 1981, as supplemented December 10, 1982, and December 29, 1983, complies with the standards and requirements of the ' Atomic Energy Act of 1954, as amended (the Act) and the Comi'ssion's rules and re~gulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the-application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

Th? issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all ap'plicable requirements have been satisfied.

2.

Accordingly, Facility Operating License No. DF9 42 is hereby amended by deleting paragraphs 2.F and 2.G.

The lir.nu is further amended by changes to the Technical Specificati e

t itcated in the' attachment to this license amendment, and paragrara

~('

s hereby amended to read as follows:

(2) Technical Specifications

~

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 94 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

s

\\

s

l

.. 3.- This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f{py:x : ;--

~

Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance:

May 25, 1984 L

,7 6

b

)

ATTACHMENT'TO LICENSE AMENDMENT NO. 98 FACILITY OPERATING LICENSE N0. DPR-62 DOCKET N0 50-324 Revise the Appendix A Technical Specifications as follows:

Remove Insert XIV XIV XV XV XVI XVI 6-1 Thru 6-26 6-1 Thru 6-33 1

I w e 1

w r

w

i l

l I

INDEX ADtINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY.............................................

6-1 6.2 ORGANIZATION 6.2.1 0FFSITE................................................

6-1 6.2.2 FACILITY STAFF.........................................

6-1 6.2.3 ONSITE NUCLEAR S AFETY GROUP Function...............................................

6-8 Re s ponsib ilitie s.......................................

6-8 Authority..............................................

6-8 6.2.4 SHIFT TECHNICAL ADVIS0R................................

6-8 6.3 F ACILITY STAFF Q UALIFICATIONS..............................

6-8 6.4 TRAINING...................................................

6-8 6.5 REVIEW AND AUDIT 6.5.1 N UCLEAR S AFETY REVIEWERS...............................

6-9 6.5.2 SAFETY EVALUATIONS AND NUCLEAR REVIEW CONTROL Sa f e ty Evaluations.....................................

6-9 Procedures, Tests, and Experiments.....................

6-10 Mo d i f ica t i on s..........................................

6-10 Operating License / Technical Specifications.............

_6-10 6.5.3 PLANT NUCLEAR SAFETY COMMITTEE (PNSC)

Function...............................................

6-11 Co mp o s i t i on............................................

6-11 Alternates.............................................

6-11 Me e t ing Fr e quency......................................

6-11

]

Quorum.................................................

6-11 Ac t iv i t i e s.............................................

6-12 Authority..............................................

6-13' Re c o r d s................................................

6-13 BRUNSWICK - UNIT 2 XIV Amendment No. 98

l 1

INDEX AD(INISTRATIVE CONTROLS i

SECTION PAGE 6.5.4 CORPORATE NUCLEAR SAFETY SECTION Function...............................................

6-13 Organization...........................................

6-13 Review.................................................

6-14 Records................................................

6-15 6.5.5 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM Function...............................................

6-16 Audits.................................................

6-16 Records................................................

6-17 Authority..............................................

6-17 Personnel..............................................

6-17 6.5.6 0UTSIDE AGENCY INSPECTION AND AUDIT PROGRAM............

6-18 6.6 REPO RTABLE OCCURRENCE ACTI0N...............................

6-18 6.7 S AF ETY L U41T V IO LATION.....................................

6-18 6.8 P ROCEDURES AND P R0 GRAMS....................................

6-19 6.9 REPORTING REQUIREMENTS Ro utine Re ports and Re portable Occurrences.............

6-20 S ta r tu p Re p o r t s........................................

6-20 An nua l Re po r t s.........................................

6-21 Personnel Exposure and Monitoring Report...............

6-21 Annual Radiological Environmental Operating Report.....

6-22 Semiannual Radioactive Ef fluent Release Report.........

6-23 Mo n thly Op e ra ting Re po r t s..............................

6-24 Re po rtable Occurrence s.................................

6-25 Promp t No tification With Written Fo llowup..............

6-25 Thirty Day Writ ten Re ports.............................

6-27 S p e c ia l Re p o r t s........................................

6-28 BRUNSWICK - UNIT 2 XV Amendment No. 98

INDEK ADMINISTRATIVE CONTROLS i

t SECTION PAGE 4

6.10 RECORD RETENTION..........................................

6-28 6.11 RADIATION PROTECTION PR0 GRAM..............................

6-30 6.12 HIGH RADIATION AREA.......................................

6-30 6.13 0FFSITE DOSE C ALCULATION MANUAL (0DCM)....................

6-31 6.14 PROCESS CONTROL PROGRAM (PCP).............................

6-31 6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS..........................

6-32 1

i

+

BRUNSWICK - UNIT 2 XVI~

Amendment No.98

1 6.0 A3fINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Manager - Brunswick Plant shall be responsible for overall I

facility operation and shall delegate in writing the succession to this responsibility during his absence.

6.2 ORGANIZATION OFFSITE 6.2.1 The of fsite organization for f acility management and technical support shall be as shown on Fi'gure 6.2.1-1.

FACILITY STAFF 6.2.2 The f acility organization shall be as shown on Figures 6.2.2-1 and 6.2.2-2 and:

a.

Each f acility on duty shift shall be composed of at least the minimum facility shif t crew composition shown in Table 6.2.2-1.

5.

At least one licensed Reactor Operator shall be in the control room wr.ea tuel is in the reactor.

c.

'anen either reactor is in OPERATIONAL CONDITION 1, 2, or 3, at least one licensed Senior Reactor Operator shall be in the control room.

d.

An 1.12ividnsi qualified to implement radiation protection procedures shall be onsite when fuel is in either reactor.*

l e.

All CORE ALTERATIONS shall be 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.

f.

A Fire Brigade of at least five members shall be maintained onsite at all times.*

The Fire Brigade shall not include the minimum shif t l

crew shown in Table 6.2.2-1 or any personnel required for other essential functions during a fire emergency.

• The individual qualified to implement radiation protection procedures and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed two hours in order to accommodate unexpected i

absence provided immediate action is taken to fill the required positions.

BRUNSWICK - UNIT 2 6-1 Amendment No.98

s ADilNISTRATIVE CONTROLS FACILITY STAFF (Continued) g.

Administrative procedures shall be developed and implemented to ILnit the working hours of facility staff who perform safety-related functions; e.g., senior reactor operators, reactor operators, health phyeicists, auxiliary operators, and key maintenance personnel.

These procedures shall meet the working hour guidelines published by the Consission in Generic Letter No. 82-12.

BRUNSWICK - UNIT 2 6-2 Amendment No.98

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i VICE PRESIDENT BRUNSWICK NUCLEAR PROJECT s

i GENERAL MANAGER BRUNSWICK PLANT I

MANAGER i

PL' ANT OPERATIONS 1

MANAGER OPERATIONS I

i I

OPERATIONS l

SUPERINTENDENT I

I PRINCIPAL ENGINEER SHIFT OPERATING PRINCIPAL ENelNEER PRINCIPAL ENGINEER OPERATIONS (UNIT 1)

SUPERVISOR (SRO)OPERATIONS (UNIT 2)

OPERATIONS I

I SHIFT FIRE BRIGADE 4

4X SENIOR SPECIALIST (EMERGENCY COORDINATION)

FIRE PROTECTION l

I

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SHIFT FIRE BRIGADE FIRE PROTECTION SUPPORT GROUP l

i LEGEND 5 NomSer of BriSede Fire Chiefe verlee with Shift OrSeelastion 44 One Speelellet le AselSned The Duties of The Plant Fire Chief Amendment No.

SRUNSWICK STEAM &

E CTRIC PLANT PLANT FIRE PROTECTION ORGANIZATION Carohna Powet & Light Company pleure S.s.s-s Pese No. e.e Amendmnt'!!o. 98

- - --. - - _ _ _ - - - _.. ~ -.., -. - - _ -. -. -

TABLE 6.2.2-1 l

MINIMUM FACILITY SHIFT CREW COMPOSITION l

WITH UNIT L IN CONDITION 1, 2, OR 3 i

POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 i

SOS 1

1(D)

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1 I(b)

RO 3

3 A0 3

3 STA 1

1

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i WITH UNIT 1 IN CONDITION 4 OR 5 I

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POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION 1

CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 j'

SOS 1

IlDI SR{a(a) g g(b)

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RO 3

2 i

A0 3

3 i

STA 1

None WITH UNIT 1 DE-FUELED I

l POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION l

CONDITIONS 1, 2, & 3 CONDITIONS 4 & 5 4

k SOS 1

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1(b)

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2 A0 3

3 STA 1

None 4

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- BRUNSWICK - UNIT 2 6-6

' Anendment No. 98 l

1

.=.

d TABLE 6.2.2-1 (Continued)

MINIMIN FACILITY SHIFT CREW COMPOSITION TABLE NOTATION SOS - Shif t Operating Supervisor with a Senior Reactor Operators License SRO - Individual with a Senior Reactor Operators License RO - Individual with a Reactor Operators License AD - Auxiliary Operator (non-licensed individual) 1 STA - Shif t Technical Advisor

• i (a) Assumes each individual is licensed on both plants.

(b) Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS.

The Shif t Crew Composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shif t crew members provided immediate action is taken to restore the Shif t Crew Composition to within the minimum requirements of Table 6.2.2-1.

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d BRIMSWICK - UNIT 2 6-7 Amendment No. 98

AININISTRATIVE CONTROLS l

l 6.2.3 ONSITE NUCLEAR S AFETY (ONS) l FUNCTION l

6.2.3.1 The ONS Unit shall function to examine f acility operating j

characterisitics, NRC issues, industry advisories, and other sources which may indicate areas for improving f acility safety.

RESPONSIBILITIES 6.2.3.2 The ONS Unic shall be responsible for maintaining surveillance of facility activities to provide independent verification

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

AUTHORITY l

6.2.3.3 The ONS Unit shall make detailed recommendations for revised

]

procedures, equipment modifications, or other means of improving facility safety to the Manager-Corporate Nuclear Safety Section.

6.2.4 shirr TECHNICAL ADVISOR 6.2.4.1 The Shif t Technical Advisor shall serve in an advisory capacity to the j

Shif t Operating Supervisor on matters pertaining to the engineering aspects l

assuring safe operation of the unit.

j 6.3 FACILITY STAFF QUALIFICATION 6.3.1 Each member of the f acility staff defined in Figure 6.2.2-1 shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Manager - Environmental & Radiation Control who l

shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) che Shif t Technical Advisor who shall have a bachelor's degree or i

equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant during transients and l

accidents.

1 6.4 TRAINING i

6.4.1 A retraining and replacement training program for the f acility staff shall be maintained under the direction of the Director - Training and shall l

meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix *A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Manager-Operations and shall meet or exceed the requirements l

of Section 27 of the NFPA Code-1975.

• Not responsible for sign-off function.

i 1

BRUNSWICK - UNIT 2 6-8 Amendment No. 98

l

[

s ADMINISTRATIVE CONTROLS 6.5 REVIEW AND AUDIT l

6.5.1 NUCLEAR S AFETY REVIEWERS 6.5.1.1 Individuala shall be designated / approved by the General Manager -

Brunswick Plant for performing nuclear safety reviews.

6.5.1.2 Individuals designated under Specification 6.5.1.1 above shall have an academic degree in an engineering or related field or equivalent and two years related experience.

6.5.1.3 A list shall be maintained of individuals qualified to perform nuclear safety reviews, including additional individuals whose expertise may be necessary during the reviews to assure that the reviewers collectively possess the background and qualfications in the disciplines necessary and important to the specific review.

6.5.1.4 The list specified in Specification 6.5.1.3 above shall include the disciplines for which each individual is qualified.

6.5.1.5 For those cases where interdisciplinary reviews are required, as many individuals as necessary shall be used to perform the nuclear review function.

6.5.1.6 One of the two nuclear safety reviewers shall be an individual other than the original preparer or the individual approving the action.

6.5.2 SAFETY EVALUATIONS AND INDEPENDENT REVIEW CONTROL S AFETY EVALUATIONS 6.5.2.1 A safety evaluation shall be prepared for each of the following:

a.

Changes to procedures required by Specification 6.8, or changes to 1

other procedures that affect nuclear safety.

b.

Proposed tests or experiments that affect nuclear safety.

c.

Proposed modifications to plant systems or equipment that af fect nuclear safety.

d.

Proposed changes to the Technical Specifications.

e.

Proposed changes to the Operating License.

6.5.2.2 Two nuclear safety reviews of the item and safety evaluation (s) prepared in accordance with Specification 6.5.2.1 above shall be performed prior to approval and implementation.

6.5.2.3 The item and associated safety evaluation (s) shall be examined in order to determine whether an interdisciplinary review is required in accordance with Specification 6.5.1.5 above.

BRUNSWICK - UNIT 2 6-9 Amendment No.98

ADMINISTRATIVE CONTROLS PROCEDURES, TESTS, AND EXPERIMENTS 6.5.2.4 The safety evaluation prepared in accordance with Specifications 6.5.2.1.a and 6.5.2.1.b above shall include a written determination, with basis, of whether or not the procedures, proposed tests and experiments, and changes thereto constitute an unreviewed safety question as defined in 10 CFR 50.59, or whether they involve a change to the Technical Specifications.

6.5.2.5 Following the nuclear safety review, the procedures required by Specification 6.8, other procedures that affect nuclear safety, proposed tests or experiments, and changes thereto (other than editorial or typographical) which have been determined to not involve an unreviewed safety question as defined in 10 CFR 50.59 or change to the Technical Specifications shall be approved prior to implementation by the General Manager - Brunswick Plant or l is previously designated alternate.

MODIFICATIONS l

6.5.2.6 The safety evaluation prepared in accordance with Specification 6.5.2.1.c above shall include a written determination, with basis, of whether or not the proposed modification is a change in the facility 1

as described in the safety analysis report, involves a change to the Technical Specifications, or constitutes an unreviewed safety question as defined in 10 CFR 50.59.

6.5.2.7 Following the nuclear safety review, proposed modifications which have been determined to not involve an unreviewed safety question as defined in 10 CFR 50.59 or a changs to the Technical Specifications shall be approved by 4

the General Manager - Brunswick Plant or his previously designated alternate.

OPERATING LICENSE / TECHNICAL SPECIFICATIONS 6.5.2.8 The safety evaluation prepared in accordance with Specifications 6.5.2.1.d and 6.5.2.1.e above shall include a written preliminary determination, with basis, of whether or not the proposed Operating License / Technical Specification change (s) is a change in the facility as described in the safety analysis report.

j 6.5.2.9 Following the nuclear safety review of the safety evaluation prepared in accordance with Specifications 6.5.2.1.d and 6.5.2.1.e above and the associated proposed action, the request shall be:

4.

Reviewed by the Plant Nuclear Safety Committee in accordance with Specification 6.5.3.8.

b.

Reviewed by the Corporate Nuclear Safety Section in accordance with Specification 6.5.4.9.

}

4 1

1 BRUNSWICK - UNIT 2 6-10 Amendment No. 98

ADfINISTRATIVE CONTROLS

)

6.5.3 PLANT NUCLEAR S AFETY COMMITTEE (PNSC)

FUNCTION 6.5.3.1 As an ef fective means for the regular review, overview, evaluation, and maintenance of plant operational safety, a Plant Nuclear Safety Committee (PNSC) shall be established.

1 6.5.3.2 The PNSC shall function throsqgh the utilization of sthcommittees, j

audits, investigations, reports, and/or performance of reviews as a group.

COMPOSITION 6.5.3.3 The PNSC shall be composed of the:

Chairman:

General Manager - Orunswick Plant

• Member:

Manager - Plant Oper stions Member:

Manager - Technical & Administrative Support Member:

Manager - Technical Supporc Member:

Manager - Operations Member:

Manager - Maintenance Member:

Manager - Environmental & Radiation Control 1

Member:

Assistant to Plant General Manager Member:

Director - QA/QC l

Member:

Director - Regulatory Compliance Member:

Director - Administrative Support j

ALTERNATES j

6.5.3.4 All alternate members shall be appointed in writing by the PNSC Chairman to serve on a temporary basis; however, no more than two alternates

]

shall participate as members at any one time.

6.5.3.5 All alternates, shall as a minimum, meet equivalent qualification i

criteria as specified for professional-technical personnel in Section 4.4 of ANSI N18.1-1971.

3Property "ANSI code" (as page type) with input value "ANSI N18.1-1971.</br></br>3" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. 1

MEETINC FREQUENCY 6.5.3.6 The PNSC shall meet at least once per calendar month and as convened by the PNSC Chairman or his designated alternate.

QUORIM 6.5.3.7 The minimum quorum of the PNSC necessary for the performance of the PNSC activities of the Technical Specifications shall consist of the PNSC Chairman or his designated alternate and five members including alternates. No i

more than two alternates shall be counted toward meeting the minimum quortsa requirement.

I

• Or designated alternate.

BRtMSWICK - UNIT 2 6-11 Amendment No.'98

J 4DtINISTRATIVE CONTROLS ACTIVITIES 6.5.3.8 The PNSC activities shall include the following:

a.

Review of all procedures required by Specification 6.8 and changes thereto (and any other procedures and changes thereto), any of which constitute an unreviewed safety question or involve a change to the Technical Specifications, prior to implementation.

b.

Review of all proposed tests or experiments that constitute an unreviewed safety question or involve a change to the Technical Specifications, prior to implementation.

i Review of all proposed modifications that constitute an unreviewed c.

safety question or involve a change to the Technical Specifications, prior to implementation.

d.

Review of all proposed changes to the Technical Specifications or Operating License, prior to implementation.

e.

Review of reports on violations of Technical Specifications including reports covering evaluation and recommendations to prevent recurrence to the Vice President - Brunswick Nuclear Project and to the Manager -

Corporate Nuclear Safety Section.

f.

Performance of special reviews, investigations (or analyses), and reports thereon as requested by the Manager - Corporate Nuclear Safety 1

Se ction.

3 Review of events requiring 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written notification to the Commission.

h.

Review of facility operations to detect potential nuclear safety hazards.

1 1.

Annual review of the Security Plan.

j.

Annual review of the Energency Plan.

k.

Review of accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence and the forwarding of these reports to the Vice President -

Brunswick Nuclear Project and the Manager - Corporate Nuclear Safety Section.

1.

Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE

(

CALCULATION MANUAL.

i BRUNSWICK - UNIT 2 6-12 Amendment No. 98

AD(INISTRATIVE CONTROLS 1

AITrHORITY 6.5.3.9 If there is a disagreement between recommendations of a majority of the PNSC and the actions contemplated by the General Manager - Brunswick Plant, the r

PNSC shall provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President -

Brunswick Nuclear Project and the Vice President - Corporate Nuclear Safety and Re search. Se course determined by the General Manager - Brunswick Plant to be the most conservative shall be followed.

RECORDS 6.5.3.10 he PNSC shall maintain written sinutes of each PNSC meeting that, at a minimum, document the results of all PNSC activities performed under the provisions of these Technical Specifications. Copies shall be provided to the Vice President - Brunswick Nuclear Project and the Manager - Corporate Nuclear Safety Section.

4 6.5.4 CORPORATE NUCLEAR SAFETY SECTION FUNCTION 6.5.4 1 The Corporate Nuclear Safety Section (CNSS) of the Corporate Nuclear Safety & Research Department shall function to provide independent review of significant plant changes, tests, and procedures; verify that REPORTABLE j

OCCURRENCES are investigated in a timely manner and corrected in a manner that reduces the probability of recurrence of such events; and detect trends that say not be apparent to a day-to-day observer.

ORGANIZATION i

i 6.5.4.2 The individuals assigned responsibility for independent reviews shall be i

specified in technical disciplines. These individuals shall collectively have the experience and cogatence required to review activities in the following areas:

I l

a.

nuclear power plant operations i

b.

nuclear engineering l

c.

chemistry and radiochemistry i

d.

metallurgy e.

non-destructive testing 1

f.

instnmentation and control g.

radiological safety 4

j h.

mechanical and electrical engineering 1.

administrative controls.

i BRtf88WIOC - UNIT 2 6-13 Amendment No. 98

. ~.

l ADtINISTRATIVE CONTROLS 1

ORGANIZATION (Continued) j.

seismic and environmental k.

quality assurance practices 1.

Other appropriate fields associated with the unique characteristics l

of the nuclear power plant.

I f

l 6.5.4.3 The Manager - Corporate Nuclear Safety Section shall have an academic l

j degree in an engineering or related field and, in addition, shall have a i

sinimum of ten years related experience, of which a minimum of five years j

shall be in the operation and/or design of nuclear power plants.

6.5.4.4 The independent safety review progran reviewers shall have an j

academic degree in an engineering or related field or equivalent and, in l

addition, shall have a minimum of five years related experience.

6.5.4.5 An individual may possess competence in more than one specialty j

area.

If sufficient expertise is not available within the Corporate Nuclear i

Safety Section, competent individuals from other Carolina Power & Light j

Company organisations or outside consultants shall be utilized in performing independent reviews and investigations.

I 6.5.4.6 At least three individuals, qualified as discussed in 6.5.4.4 above i

shall review each itse submitted under the requirements of Section 6.5.4.9.

)

6.5.4.7 Independent saf ety reviews shall be performed by individuals not directly involved with the activity under review or responsible for the activity under review.

6.5.4.8 The Corporate Nuclear Safety Section independent safety review program shall be conducted in accordance with written, approved procedures.

REVIEW j

6.5.4.9 The Corporate Nuclear Safety Section shall perform reviews of the j

fallowing:

l a.

The safety evaluations for 1) changes to procedures required by Specification 6.8, 2) modifications of equipment or systems, and

3) tests or experiments that constitute a change to the safety d

analysis report to verify that such actions did not constitute an unreviewed safety question or involve a change to the Technical I

Specifications.

Implementation any proceed prior to completion of 1~

this review, I

b.

Proposed changes to procedures required by Specification 6.8, and proposed modifications that constitute an unreviewed safety question as defined in 10 CFR 50.59 or a change to the Technical Specifications, prior to implementation.

BRIM 8 WICK - UNIT 2 6-14 Amendesnt No.98

i ADlINISTRATIVE CONTR01.S i

REVIEW (Continued) 4 c.

Proposed tests or experiments that involve an unreviewed safety question as defined in 10 CFR 50.59 or a change to the Technical Specifications, prior to implementation.

l t

d.

Proposed changes to the Technical Specifications and Operating I

j I.icense.

l I

e.

Violations, deviations, and events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> written j

j notification to the Commission, such as l

1.

Violations of applicable codes, regulations, orders, Technical

{

Specifications, license requirements, and internal procedures or t

j instructions having nuclear safety significance.

2.

Significant operating abnormalities or deviations from normal and expected performance of plant safety-related structures, i

systees, or components.

1 f.

Reports and minutes of the PNSC.

i j

g.

Any other matter involving safe operation of the nuclear power plant that the Manager - Corporate Nuclear Safety Section deems appropriate for consideration or which is referred to the Manager - Corporate s

I Nuclear Safety Section by the on-site operating organisation or other f unctionni organisational units within Carolina Power & !.ight Company.

i 6.5.4.10 Review of itses considered under 6.5.4.9(e) throtch (g) above shall

{

include the results of any investigations made and the recommendations j

resulting f rom these investigations to prevent or reduce the probability of

?

recurrence of the event.

i i

I RECORDS i

i j

6.5.4.11 Records of Corporate Nuclear Safety Section reviews, including recommendations and concerns, shall be prepared and distributed as indicated below j

a.

Copies of doctmented reviews shall be retained in the CNSS files.

l b.

Recommendations and concerns shall be submitted to the General 3

1 Manager - Brunswick Nuclear Project and Vice President - Brunswick 1

l Plant, within 14 days of completion of the review.

c.

A simmation of Corporate Nuclear Safety recommendations and concerne shall be submitted to the Chairman / President and Chief Executive 4

Of ficer; gaecutive Vice President - Power Supply and Engineering and i

Construction; Vice President - Corporate Nuclear Safety and Research; i

Vice President - Brunswick Nuclear Project; General Manager -

Brunswick Plant; and others, appropriate, on at least a bi-monthly frequency.

--- 0 RET 28 - EB9 fL

- AO i -

m & M ---------------

l AIMINISTRATIVE CONTROLS l

6.5.5 CORPORATE QUALITY ASSURANCE AUDIT PROGRAM FUNCTION 6.5.5.1 he Performance Evaluation thit (PEU) of the Corporate Quality Assurance Department shall function to perform audits of facility activities specified in Specification 6.5.5.2.

AUDITS 6.5.5.2 Audits of f acility activities shall be performed by the PEU. These audits shall encompass:

We conformance of f acility operation to provisions contained within a.

l the Technical frecifications and applicable license conditions at i

least once per 12 aooths.

l b.

De training and qualifications of the entire f acility staf f at least i

once per 12 months.

j c.

He results of actions taken to correct deficiencies occurring in facility equipment, structures, systees, or methods of operation that 7

affect nuclear safety at least once per 6 months.

d.

The verification of compliance and implementation of the requirements of the Quality Assurance Program to meet the criteria of i

Appendix "B", 10 CFR 50, at least once per 24 months.

I De Energency Plan and implementing procedures at least once per 12 e.

i months.

I f.

Se Security Plan and implementing procedures at least once per 12 months.

I g.

De Facility Fire Protection Program and implementing procedures at least once per 12 months.

h.

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

1 2

1.

Se OFFSITE DOSE CALCULATIONAL MANUAL and implementing procedures at taast once per 24 months.

J.

Se PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive westes at least once per 24 months.

4 3Rt388WIQt - UNIT 2 6-16 Amendment No. 98

)

i ADf1NISTRATIVE CONTROLS i

AUDITS (Continued) k.

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.

l 1.

Any other area of f acility operation considered appropriate by the Manager - Quality Assurance Services Section.

6.5.5.3 Personnel performing the quality assurance audits shall have access to the plant operating records.

RECORDS 6.5.5.4 ascords of audits shall be prepared and retained.

i 6.5.5.5 Audit reports encompassed by 6.5.5.2 above shall be prepared, approved by the Principal QA Specialist - Performance Evaluation Unit, and forwarded to the Executive Vice President - Power Supply and Engineering and Construction; Vice President - Brunswick Nuclear Project; Vice President -

i Corporate Nuclear Safety and Research; General Manager - Brunswick Plant; and l

others, as appropriate, within 30 days af ter completion of the audit.

ALTfHORITY i

j 6.5.5.6 The Manager - Quality Assurance Services Section under the Manager -

Corporate Quality Assurance shall be responsible for the following:

1

)

a.

The administering of the Corporate Quality Assurance Audit Program.

i J

b.

The approval of the individual (s) selected to conduct quality assurance audits.

PERSONNEL l

6.5.5.7 Audit personnel shall be independent of the area audited.

6.5.5.8 Selection of personnel for auditing assignments shall be based on experience or training that establishes that their qualifications are commensurate with the complexity or special nature of the activities to be a ud ited.

In selecting audit personnel, consideration shall be given to special abilities, specialised technical training, prior pertinent experience, personal characteristics, and education.

o 6.5.5.9 Qualified outside consultants or other individuals independent from I

those personnel directly involved in plant operation shall be used to augment l

the audit teams when necessary.

l BRUNSWICK - UNIT 2 6-17 Amendment No. 98

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l l

h ADtINISTRATIVE CONTROLS 6.5.6 - OUrSIDE AGENCY INSPECfION AND AUDIT PROGRAM l

6.5.6.1 An independent f tre protection and loss prevention inspection and l

audit shall be performed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection fire.

6.5.6.2 An inspection and audit of the fire protection and loss prevention l

program shall be performed by an outside qualified fire consultant at intervals no greater than 36 months.

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 seraltted pursuant to the requirements of Specification 6.9.

b.

Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to.the Commission shall be reviewed by the General Manager - Brunswick Plant and submitted to the Manager - Corporate Nuclear Safety Section and the Vice President - Brunswick Nuclear Project.

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

The f acility shall be placed in at least HOT SHUTDOWN within two a.

hours, b.

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

Brunswick Nuclear Project and the Manager - Corporate Nuclear Safety Section shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

A Saf ety Limit Violation Report shall be prepared.

The report shall be reviewed by the General Manager - Brunswick Plant. This report j

shall describe (1) applicable ciretastances 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 Vice President - Brunswick Nuclear Project, and the l

Manager - Corporate Nuclear Safety Section within 14 days of the violation.

l BRIMSWICK - UNIT 2 6-18 Amendment No. 98

=_

_.=_-. _.. -. --

i ADMINISTRATIVE CONTROLS 1

6.8 PROCEDURES AND PROGRAMS 3

6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

The applicable procedures recommended in Appendix "A" of Regulatory a.

Guide 1.33, November 1972.

b.

Refueling operations.

c.

Surveillance and test activities of safety related equipment.

d.

Security Plan implementation.

e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

i j

g.

OFFSITE DOSE CALCULATION MANUAL implementation.

1 h.

PROCESS CONTROL PROGRAM implementation.

1.

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 Temporary changes to procedures of Specification 6.8.1 above, any other procedures that affect nuclear safety, and proposed tests or experiments may j

be made provided:

The intent of the original procedure, proposed test or experiment is a.

a not altered.

b.

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

c.

The change is documented, reviewed pursuant to Specifications 6.5.2.1 i

and 6.5.2.2 and approved by the General Manager - Brunswick Plant or his previously designated alternate within 14 days of implementation.

6.8.3 The following prograns'shall be established, implemented, and i

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 program shall include the following:

BRUNSWICK - UNIT 2 6-19 Amendment No. 98.

=

ADtINISTRATIVE CONTROLS I

PROCEDURES AND PROGRAMS (Continued) i 1.

Preventive maintenance and periodic visual inspection requirements, and 2.

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

b.

In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident i

conditions. Thi,s program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

c.

Post-Accident Sampling I.

A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous i

ef fluents, and containment atmosphere samples under accident i

conditions. The program shall include the following 1.

Training of personnel, 4

2.

Procedures for sampling and analysis, and 3.

Provisions for maintenance of sampling and analysis equipment.

4 6.9 REPORTING REQUIREMENTS RottrINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, code i

of Federal Regulations, the following reports shall be submitted to the l

Regional Administrator of the Regional Office unless otherwise noted.

l STARTUP REPORTS 6.9.1.1 A stannary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license. (2) amenddent to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manuf actured by a different fuel l

supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

3RUNSWICK - UNIT 2 6-20 Amendment No. 98

_ _ _ _ = _ _

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t 4

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I ADf1NISTAATIVE CONTROLS r

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i

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STARTUP REPORTS (Continued)'

i 6.9.1.2 he startup' report shall address each of the cents identified in the l

(

l FSAR and shall include a description of the measured values of the operating i

conditions or characteristics obtained during the test program and a cogarison 1

j of these values with design predictions and specifications. Any corrective l

i actions 'that were required to obtain satisfactory operution shall also be j

described. Any additional specific details required in license conditions i

based on other commitments shall be included in this report.

I 6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days fellowing resumption or commencement of commercial power operation,' (3) 9 mon:he following initial i

criticality', whichever is earliest. If the Startup F.eport does not cover all j

three events, L'.e.,

initial criticality, completion of startup test program, 1

and resumptior, or commencement of consorcial power o,peration, supplementary i

reports shall be submitted at least every three mont.hs until all three events have been coupleted.

i l

l ANNUAL REPORTS.l/

t I

i 6.9.1.4 Ann (aal reports covering the activitier of the unit as described below i

i during the previous calendar year shall be r t dtted prior to March 1 of each year.

De initial report shall be submitts arior to March 1 of the year following initial criticality.

/

)

PERSONNgL EXP08tRg AMD MONITORING Rgp0tT) 2 6.9.1.5 Reports required on an annual basis shall include a tabulation of the I

number of station, utility, and other personnel, including contractors, receiving exposures greater than 100 aree/yrgnd their associated man-ree exposure according to work and job functione-., e.g., reactor operations and surveillance, inservice inspection, routine enintenance,;special maintenance (describe maintenance), waste processing, and refueling. He dose assignments j

to various duty functions may be estiented, based on pocket dosimeter, TLD, or i

film badge measurements. Seali exposures totalling less than 20 percent of the j

individual total dose need not be accounted for.

In the aggregate, at least i

80 percent of the total'whole body dose received f rom external sources shall be

)

assigned to specific 'sajor work functioud.

1 j

i s.

1 t

1/ A single admittal may be ande for a multiple unit' station.

Se submittal should cochine those7ections that are common to all yntes at the station.

4 t

p y

s,i, 1/ 21s tabulation shp'plements the requirements of $20.407 of 10 CFR part 20.

i

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j i

att41SWIOC UNIT 2 6-21 Amendesnt No. M i

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___-_______--____m,

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ADlINISTRATIVE CONTROLS i

ANNIRL RADIOLOGICAL ENVIR0lGIENTAL OPERATING REPOR 6.9.1.6 Routine radiological environmental operating reports covering the l

operacian of the f acility during the previous calendar year shall be submitted l

prior to May 1 of each year.

l l

6.9.1.7 The Annual Radiological Environmental Operating Report shall include t

the following:

a.

Summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with pre-operational studies, t

I with operational controls (as appropriate), and with previous

[

environmental surveillance reports, and an assessment of the observed l

)

impact of the plant operation on the environment.

[

I b.

Results of land uses censuses required by Specificaton 3.12.2.

i l

l c.

Results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the 1

0FFSITE DOSE CALCULATION MANIEL, as well as smearised and tabulated j

results of these analyses and measurements in the format of Table 3 in the Radiological Assessment Branch Technical Position, Revision 1, i

November 1979.

In the event that some individual results are not j

available for inclusion with the report, the report shall be submitted i

noting and explaining the reasons for the missing results. The l

l sissing data should be submitted as soon as possible in a supplementary report.

l 1

1 i

i d.

A sumasry description of the radiological environmental monitoring progras.

L i

e.

At least two legible napeO covering all sampling locations keyed to a j

table giving distances and directions from the centerline of one

{

reactor.

i f.

The results of licensee participation in the Intertaboratory Comparison Program, required by Specification 3.12.3.

i 1

1/ A single s4atttal may be made for a multiple unit station.

j4 One map shall cover stations near the SITE BOUNDARY; a second esp shall include the more distant stations.

l BRUNSWIQt - UNIT 2 6-22 mnendment No. 98

AD(INISTRATIVE CONTROL 3 ANNUAL RADIOLOGICAL ENVIROPMENTAL OPERATING REPORT (Continued) g.

Discussion of all da riations from the sampling schedule of Tabla 3.12.1-1.

h.

Discussion of all analyses in which the LLD required by Table 4.12.1-1 was not achievable.

SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPOR 6.9.1.8 Routine radioactive effluent release reports covering the operation of the f acility during the previous 6 months of operation shall be submitted within the time periodt specified in Specificati'ons 6.9.1.9 and 6.9.1.10 below.

6.9.1.9 The portion of the Sealannual Radioactive Ef fluent Release Reports to be submitted within 60 days af ter January 1 and July 1 of each year shall include the following:

a.

A stmaary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the facility as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Westes and Releases of Radioactivity Materials in Liquid and Gaseous Ef fluents f rom Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis similar to tifs format of Appendix B thereof.

b.

The information specified below for each class of solid waste (as defined by 10 CFR Part 61, when Laplemented) shipped offsite during the report period:

1.

Container volume, 4

2.

Totial curie quanEity (specify whether determined'hy measurement or estimace),

3.

Principe radionuclides (specify Nether determined by measurement or estimate),

4.

Source of waste and processing employed (e.g., dewatered spent.

resin, compacted dry waste, evaporator bottoes),

s Type of container (e.g., LSA, Type A,, Typ'e B, Large Quantity),

5.

and

- _ z z ___ __

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' '

• f

,5) A aingle' submittal may be made for a mitiple unit station. The submittal j

should combine those. secticas that are common to all units at the station.

)

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a

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s Amendment No. 98

.BRUNSWICKs-UNIT 2 6-23 4

ul 4

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i 1

AININISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) 6.

Solidification agent or absorbent (e.5., cement, urea f o rmaldehyd e).

c.

A list of description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

d.

Any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) or the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

6.9.1.10 The portion of the Semiannual Radioactive Ef fluent Release Report to be submitted within 90 days af ter January 1 of each year shall include the following:

a.

An annual summary of hourly motorological data collected over the previous calendar year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint f requency distri ytions of wind speed, wind direction, and atmospheric stability b.

An assessment of the radiation doses due to the radioactive liquid and gaseous ef fluents released f rom the station during the previous calendar year.

MONTHLY OPERATING REPORTS 6.9.1.11 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to main steam 'systas safety / relief valves, shall be sdsmitted on a monthly basis to the Director, Of fice of

-Resource Management, U. S. Nuclear Regulatory Commission, Washington, D.C.

20555, with a copy to the Regional Administrator of the Regional Of fice no later than the 15th of each month following the calendar month covered by the report.

6/ in lieu of sdamission with the Semiannual Radioactive Ef fluent Release Report, the licensee has the option of retaining this summary of required meteorological data in a file that shall be provided to the NRC upon

' request.

BRUNSWICK - UNIT 2 6-24

. Amendment No. 98

i ADMINISTRATIVE CONTROLS REPORTABLE OCCURRENCES j

6.9.1312 The REPORTABLE OCCURRENCES of Specifications 6.9.1.13 and 6.9.1.14 below, including currective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence.

In case of corrected or supplemental reports, a licensen event report shall be completed and reference shall be made to the original report date.

PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP-6.9.1.13 31e types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Regional Administrator of the Regional Of fice, or his designate no later l

than the first working day following the event, with a written followup report i

within 14 days. The written followup report shall include, as a minimum, a i

completed copy of the licensee event report form.

Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

]

l a.

Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint l

specified as the limiting safety system setting in the Technical Specifications or f ailure to complete the required protective

. function.

No te:

Instrument drif t discovered as a result of testing need not be reported under this item (but see 6.9.1.13.e, 6.9.1.13.f.

and 6.9.1.14.a below).

b.

Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the Technical Specifications.

Note:

If specified action is taken when a system is found to be operating between the most conservative and least conservative aspects of a limiting condition for operation listed in the Technical Specifications, the limiting J

condition for operation is not considered to have been violated and no report need be sdomitted under this section i

(but see 6.9.1.14.b below).

4 BRUNSWICK - UNIT 2 6-25 Amendment No.- 98

AD4INISTRATIVE CONTROLS PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP (Continued) c.

Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

Note:

Leakage of valve packing or gaskets within the limits for identified leakage set forth in Technical Specifications need not be reported under this section.

d.

Reactivity anomalies involving disagreement with predicted value of reactivity balance under steady-state conditions during power operation greater than or equal to 1% ok/k; a calculated reactivity

. balance indicating a SHUTDOWN MARGIN less conservative than specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcritical, an unplanned reactivity insertion of more than 0.5% ok/k; or occurrence of any unplanned criticality.

e.

Failure or malfunction of one or more components that prevents or could prevent, by itself, the f ulfillment of the functional requirements of systems required to cope with accidents analyzed in the S AR.

i f.

Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.

Note:

For 6.9.1.13.e and 6.9.1.13.f. reduced redundancy that does not result in loss of system function need not be reported under this section (but see 6.9.1.14.b and 6.9.1.14.c b elow).

g.

Conditions arising f rom natural or man-made events that, as a direct result of the event, require plant shutdown, operation of safety systems, or other protective measures required by Technical Specifications.

h.

Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in.the safety-analysis report or in the bases for the Technical Specifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

i.

Performance of structures, systems, or coeponents that requires remedial action or corrective measures to pr4 vent operation in a

)

manner less conservative than assumed in the accident analyses in the i

safety analysis report or Technical Specifications. bases; or discovery BRUNSWICK - UNIT 2

'6-26 Amendment No. 98

l AIMINISTRATIVE CONTROLS PROMPT NOTIFICATION WITH WRITTEN FOLLNUP (Continued) during plant life of conditions not specifically considered in the safety analysis report or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

Note:

This item is intended to provide for reporting of potentially generic problems.

j.

Failure or malfunction of the main steam system safety / relief valves.

l THIRTY DAY WRITTEN REPORTS 6.9.1.14 The types of events listed below shall be the subject of written reports to the Regional Administrator of the Regional Office within thirty days l

of occurrence of the event. The written report shall include, as a minimum, a completed copy of the licensee event report form.

Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

a.

Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by ' the Technical Specifications but which do not prevent the fulfillment of the functional requirements of affected. systems (but see 6.9.1.13.a and 6.9.1.13.b above).

l b.

Conditions leading to operation in a degraded mode permitted by a l

limiting condition for operation or plant shutdown required by a limiting condition for operation (but see 6.9.1.13.b above).

l No te:

Routine surveillance testing, instrument calibration, or preventive maintenance that require configurations described in 6.9.1.14.a and 6.9.1.14.b above need not be reported except where test results themselves reveal a degraded mode as described above.

c.

Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems (but see 6.9.1.13.f above).

l d.

M> normal degradation of systems other than those specified in 6.9.1.13.c above designed to contain radioactive material resulting f rom the fission process.

No te:

Sealed sources or calibration sources are not included under-this11 tem.

Leakage of valve packing or gaskets within the limits for identified leakage set forth in Technical Specifications need not be reported under this ~ item.

- BRUNSWIOC - UNIT 2 6-27 Amendeent No._93

1 ADtINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be sdamitted to the Regional Administrator of the l

Regional Office within the time period specified for each report. These reports shall be sd)mitted covering the activities identified below pursuant to the requirements of the applicable reference specification.

a.

Inoperable Seismic Monitoring Instrumentation, Specification 3.3.5.1.

b.

Seismic event analysis, -Specification 4.3.5.1.2.

c.

Reactor coolant specific activity analysis, Specification 3.4.5.

, d.

Fire detection instrumentation, Specification 3.3.5.7.

e.

Fire suppression syrtems, Specifications 3.7.7.1, 3.7.7.2, 3.7.7.3, and 3.7.7.5.

f.

ECCS actuation, Specifications 3.5.3.1 and 3.5.3.2.

g.

Fire barrier penetration, Specification 3.7.8.

h.

Liquid Ef fluents Dose, Specification 3.11.1.2.

1.

Liquid Radwaste Treatment, Specification 3.11.1.3.

J.

Dose - Noble Gases, Specification 3.11.2.2.

k.

Dose - Iodine-131, Iodine-133, Tritum, and Radionuclides in Particulate Fo rm, Specification 3.11.2.3.

1.

Gaseous Radwaste Treatment, Specification 3.11.2.4.

i m.

Ventilation Exhaust Treacaent, Specification 3.11.2.5.

n.

Total Dose, Specification 3.11.4.

o.

Monitoring Program, Specification 3.12.1.b.

6.10 RECORD RETENTION Facility records shall be retained in accordance with ANSI-N45.2.9-1974.

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

. a.

Records and logs of f acility 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.

. BRUNSWICK - UNIT 2 6-28 Amendment No. 98

AININISTRATIVE CONTROLS RECORDS RETENTION (Continued) c.

All REPORTABLE OCCURRENCES 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 and fission detector leak tests and results.

h.

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

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

a.

Records and drawing changes reflecting f acility 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 f acility radiation and contamination surveys.

d.

Records or radiation exposure for all individuals entering radiation control areas.

e.

Records of gaseous and liquid radioactive material released to the environs.

f.

Records of transient or operational cycles for those facility components identified in Table 5.7.1-1.

g.

Records of reactor tests and experiments.

h.

Records of training and qualification for current members of the plant staff.

1.

Records of inservice inspections performed pursuant to these Technical i

Specifications.

J.

Records of Quality Assurance activities required by the QA Manual.

i k.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to.

10 CFR 50.59.

BRUNSWICK - UNIT 2 6-29 Amendment No. 98 1

w a

_-A_

M AININISTRATIVE CONTROLS l

RECORDS RETENTION (Continued) 1.

Records of the service lives of all hydraulic and mechanical snubbers listed in Table 3.7.5-1 including the data at which the service life commences and associated installation and maintenance records.

Records of analyses required by the radiological environmental m.

monitoring program.

Records of (1) meetings of the PNSC, (2) meetings of the previous n.

off-site review organization, the Company Nuclear Safety Comeittee (CNSC), (3) the independent reviews performed by the Corporate Nuclear Safety Section, and (4) the independent reviews performed by the Corporate Quality Assurance Audit Program, Performance Evaluation Unit.

J 6.11 RADIATION PROTECTION PROGRAM 6.11.1 Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HICH RADIATION AREA 6.12.1 In lieu of the " Control Device" or "alara signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is 1000 aren/hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*.

Any l

individual or group of individuals permitted to enter such areas shall be l

provided with or accompanied by one or more of the following:

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. - Entry into such areas with this monitoring device may be made af ter the dose rate levels in the area have been l

established and personnel have been made knowledgeable of them.

c.

An individual qualified in radiation protection procedures who is equipped with a radiation dose rate monitoring device.

This individual shall be responsible for providing positive control over i

the activities within the area and shall perform periodic radiation j

surveillance at the f requency specified by the f acility Health Physicist in the Radiation Work Permit.

Health Physics personnel or personnel escorted by Health Physics personnel

-l shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation protection procedures for entry into high radiation areas.

BRUNSWICK - UNIT 2 6-30 Amendment No. 98

1 ADMINISTRATIVE CONTROLS ~

MIGH RADIATION AREA (Continued)

-6.12.2 The requirements of 6.12.1 above shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 meem/hr.

In addition, locked doors shall be provided to prevent unauthorized entry into such areas an'd the keys shall be maintained under the administrative control of the Operations Shif t Foreman on duty and/or the Radiation Control Supervisor.

i 6.13 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.13.1 The OFFSITE CALCULATION MANUAL (ODCM) shall be approved by the Commission prior to implementation.

I' 6.13.2 Licensee initiated changes to the ODCM:

a.

Shall be submitted to the Commission in the Semiannual Radioactive j

Ef fluent Release Report for the period in which the change (s) was made effective.

This submittal shall contain:

1.

Sufficiently detailed information to totally support rationale without benefit of additional of supplemental information.

Information submitted should consist of a package:of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate

~

analyses or evaluations justifying the change (s);

1 2.

A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations;

and, i

3.

Documentation of the f act that the change has been reviewed 'and i

found acceptable by the PNSC.

b.

Shall become effective upon review and acceptance by.the PNSC.

l 6.14 PROCESS CONTROL PROGRAM (PCP)-

1 4-6.14.1 The PROCESS CONTROL PROGRAM-(PCP)'shall be approved by the Commission:

i prior to. implementation.

6.14.2 Licensee initiated changes to the PCP:

a.

Shall be submitted to the Commission in the Semiannual Radioactive-Ef fluent Release-Report for the period in which the change (s),was.

made.. This submittal'shall contain:

- 1.:

Suf ficiently detailed information to totally support the rationale for:the-change without benefit of. additional or 3

supplemental information; BPUNSOICKL-UNITf2" 6-31 Amond$cnt No.~- 98' 4i-y-

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g

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g P 9-

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e

AININISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (PCP) (Continued) 2.

A determination that the change did not reduce the overall conformance of the solidification waste product to existing criteria for solid wastes; and 3.

Documentation of the f act that the change has been reviewed and found acceptable by the PNSC.

b.

Shall become effective upon review and acceptance by the PNSC.

6.15 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS 6.15.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):

i a.

Shall be reported to the Commission in the Semiannual Radioactive Ef fluent Release Report for the period in which the evaluation was reviewed by the PNSC. The discussion of each change shall contain:

l 1.

A summary of the evaluation that led to the determination that i

che change,could be made in accordance with 10 CFR Part 50.59; 2.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; 3.

A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems; 4.

An evaluation of the change that shows the predicted release of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; l

5.

An evaluation of the change that shows the expected maximum exposure to an individual in the UNRESTRICTED AREA and to the general population that differ f rom those previously estimated in the license application and amendments thereto; 6.

A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases _for the period prior to when the changes are to be made; 4

e 7/. Licensees may choose to sdbmit the information called for in'this Specification as part of the annual FSAR update.

I BRUNSWICK - UNIT 2 ~

6-32 Am ndment No. 98 e

i 4

ADf1NISTRATIVE CONTROLS MMOR CHANGES TO LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS l

( Co ntinued) 7.

An estimate of the exposure to plant operating personnel as a result of the change; and 8.

Docunertation of the f act that the change was reviewed and found accep:.able to the PNSC.

b.

Sh..n %cosa ef fective upon review and acceptance by the PNSC.

1 r

BRUNSWICK - UNIT 2 6-33 Amendment No. 98

.- -,. -