ML18030B034

Proposed Tech Specs Re Administrative Control.Description, Justification & Determination of NSHC Encl

ML18030B034
Person / Time
Site:	Browns Ferry
Issue date:	01/17/1986
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18030B032	List: ML18030B031 ML18030B034
References
TAC-56027, TAC-56028, TAC-56029, NUDOCS 8601290112
Download: ML18030B034 (98)

Text

ENCLOSURE 1

PROPOSED TECHNICAL SPECIFICATION REVISIONS BROWNS FERRY NUCLEAR PLANT UNIT ls 2o AND 3 (TVA BFNP TS 201 SUPPlEMENT I)

SECTION 6 - ADMINISTRATIVECONTROLS 860i290ii2 860ii7 PDR ADOPT 05000>>9I P

PDR I~

ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY...........................................

6-1 6.2 6.2.1 orporate...................,............................

C 6-1 ORGANIZATION

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6 1 6.2.2 6.3 6.4 6.5 6.5.1.

6.5.2 6.5.3 6.6 6.7 lant Staff......................................,,.'...,.

P 6-1 PLANT STAFF UALIFICATIONS...............................

6-4 TRAINING ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

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~ 6-4 PLANT REVIEW AND AUDIT........:..........................

6-4 Plant Operation Review Committee (PORC)..................

6-4 Nuclear Safety Review Board (NSRB).......................

6-9 REPORTABLE EVENT ACTIONS.................................

6-15 SAFETY LIMITVIOLATION...................................

6-16 Technical Review and Approval of Procedures..............

6-14 6.8 PROCEDURES/INSTRUCTION AND PROGRAMS...........'...........

6-17 6.8.1 6.8.2 6.8.3 6.9'.9.1 rocedures...................'............................

P 6-17 ri11s

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D 6-18 Radiation Control Procedures................

REPORTING RE UIREMENTS......................

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6 20 6-20 6-20 6-21 outine Reports..............,...........................

R S tartup Reports............................,.........,...

Annual Operating Report..................................

Monthly Operating Report.................................

6-22 Reportable Events.......................,.................

6-22 Radioactive Effluent Release Report......................

6-23 Source Tests.............................................

6-23 6.9.2 6.10 Special Reports.............................

STATION OPERATING RECORDS AND RETENTION.....

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E 4

BFN TECHNICAL SPECIFICATIONS 6.0.

ADMINISTRATIVE CONTROLS

,'. 6.0, ADMINISTRATIVECONTROLS

- 6. 1

~ RESPONSIBILITY The Plant Manager has onsite responsibilities for the safe operation of the facility and shall report to the Browns Ferry Site Director.

In the absence of the Plant Manager",

a Plant Superintendent will assume his responsibilities.

6.2 ORGANIZATION CORPORATE 6.2.1 The portion of TVA management which relates to the h

operation of the plant is shown in Figure 6.2-1.

PLANT STAFF 6.2.2 The functional organization for the operation of the

, plant shall be shown in Figure 6.2-2.

a. Shift manning requirements, shall as a minimum, be as described in Table 6.2.A and below.

b.

A licensed senior reactor operator shall be present at the site at all times. when there is fuel in the reactor.

6-1

6.2.2 (Cont.)

c.

A licensed reactor operator shall be in the control room whenever there is fuel in the reactor.

d.

Two licensed reactor operators shall be in the control room during any cold startups, while shutting down the reactor, and during recovery from unit trip.

In addition, a person holding a senior operator license shall be in the control room for that unit whenever it is in an operatiopal mode other than cold shutdown or refueling.

e.

A Health Physics Technician" shall be present at the facility at all times when there is fuel in the reactor.

f.

A person holding a senior operator'license or a senior operator license limited to fuel handling, shall be present during alteration of the core to directly supervise the activity and during this time shall not be assigned other duties.

g.

A site fire brigade of at least five members shall be maintained onsite at all times.~

The fire brigade shall not include the Shift Engineer and the other members of the minimum shift crew necessary for safe shutdown of the unit, nor any personnel required for other essential functions during a fire emergency.

<<The Health Physics Technician and fire brigade 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, provided immediate action is taken to fillthe required positions.

6-2

Table 6.2.A Minimum Shift Crew Re uirementsb Position Units in 0 eration T

e of License 0

1 2d Senior Operator Senior Operator Licensed Operators Additional Licensed Operators 1

1 1

1 0

1 2

2 3

3 3

3 0

1 2

2 SRO SRO RO or SRO RO or SRO Assistant Unit Operators (AUO) 4 4

5 5

None Shift Technical Advisor (STA)

Health Physics Technician 0

1:

1 1

1 1

1 1

'Nonce None "Note for Table 6.2.A

' a.,

A senior operator will be assigned responsibility for overall plant operation at all times there is fuel in any unit.

Except for the senior operator discussed in note "a", the shift crew composition may 'be one less than the minimum requirements of Table 6.2.A for a period of time not to exceed two hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within, the minimum requirements of Table 6.2.A.

This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

co

' d.

One of the Additional Licensed Operators must be assigned to each control room with an operating unit.

The number of required.licensed personnel, when the operating units are,'ontrolled from a common control room, are two senior operators and four operators.

e, The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and transient and accident response and analysis.

6-3

6.3 PLANT STAFF UALIFICATIONS Qualifications of the Browns Ferry Nuclear Plant management and operating staff shall meet the minimum acceptable levels as described in ANSI-N18.1, Selection and Training of Nuclear Power Plant Personnel, dated March 8, 1971.

The qualifications of the Health Physics Supervisor will meet or exceed the minimum acceptable levels as described in Regulatory Guide 1.8, Revision 1, dated September 1975.

6.4 TRAINING A retraining and replacement training program for station personnel shall be in accordance with ANSI - N18.1, Selection and Training of Nuclear Power Plant Personnel, dated March 8, 1971.

The minimum frequency of the retraining program shall be every two years.

6.5 PLANT REVIEM AND AUDIT 6.5.1 Plant 0 erations Review Committee PORC 6.5.1.1 Function a.

The PORC shall function to advise the Plant Manager in all matters related to nuclear safety.

b. This advisory function shall be performed by the PORC acting in a formal meeting or by members acting individually without a formal meeting.

6-4

Composition

'=';5;1.2 The PORC shall be composed of the:

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. a. Chair~n:

Member:

Member:

Member:

Member:

Member:

Member:

Member:

Plant Manager or PlantSuperintendent Operations 6 Engineering or Plant Superintendent, Maintenance Electrical Maintenance 1

Gr oup Supervisor Mechanical Maintenance Group Supervisor Instrument Maintenance Group Supervisor Operations Group Supervisor Engineering Group Supervisor Quality Assurance Staff Supervisor Health Physics Group Supervisor 1

b. All alternate chairmen and alternate members shall be appointed in writing by the PORC chairman.

'EETXHG FREQUENCY 6,,'5.1.3 The PORC shall convene in a formal meeting at least once a month and as directed by the chairman.

Other, PORC meetings may be requested by the chairmen or members as required.

6-5 '

6.5.1.4 For expedited meetings, when it is not practical to convene as a

group, the chairman or alternate chairman may conduct committee business by polling the members individually (by telephone or in person) or via a serialized review.

glfORUN 6.5.1.5 The quorum necessary for the PORC to act in a formal meeting shall consist of the chairman or alternate chairman and at least five members or their alternates.

Members shall be considered present if they are in telephone communication with the committee.

RESPONSIBILITIES 6.5.1.6 The PORC shall be responsible for the activities listed below.

The PORC may delegate the performance of reviews, but will maintain cognizance over and responsibility for them, e.g.,

subcommittees.

a.

Review of administrative procedures for the control of the technical and cross-disciplinary review of (1) all procedures required by specification 6.8.1.1, and changes

thereto, (2) any other procedures and changes thereto determined by the Plant Manager to affect nuclear safety.

b.

Review of the administrative procedures required by specification 6.8.l.l.a.

and changes thereto.

c.

Review of emergency operating procedures and changes thereto.

d.

Review of the Radiological Emergency Plans and the implementing procedures.

6-6

e.

Review of all proposed changes to the Technical Specifications.

f.

Review of safety evaluation for proposed tests or experiments to be completed under the provisions of 10 CFR 50.59 g.

Review of all safety evaluations for modifications to safety related structures, systems or components to verify that such actions did not constitute an unreviewed safety question as defined in 10 CFR 50.59, or requires a change to these Technical Specifications.

h.

Review of reportable events, unusual events, operating anomalies, and abnormal performance of plant equipment.

i.

Investigate reported or suspected incidents involving safety questions or violations of the Technical Specifications.

j.

Review of'nit operations to detect potential hazards to nuclear safety.

Items that may be included in this review are NRC inspection reports, QA audit, NSRB audit results, American Nuclear Insurer (ANI) inspection results, and significant corrective action reports (CARs).

k.

Performance of special reviews, investigations, or analysis, and report thereon as requested by the Plant Manager or the Nuclear Safety Review Board.

6-7

AUTHORITY 6.5.1.7 The PORC shall:

a.

Recommend to the Plant Manager in writing, approval, or disapproval of items considered under 6.5.1.6.a through f above.

1.

The recommendation shall be based on a majority vote of the PORC at a formal meeting.

2.

The recommendation shall be based on a unanimous vote of the PORC when the PORC members are acting individually.

3.

Each member or alternate member shall have one vote.

b.

Furnish for consideration a determination in writing with regard to whether or not each item considered under 6.5.1.6.f above constitutes an unreviewed.safety question.

c.

Nake recommendations to the Plant Manager in writing, that action reviewed under 6.5.1.6.g above did not constitute an unreviewed safety question,

'I d.

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 Site Director and the Nuclear Safety Review Board of disagreements between the PORC and the Plant Manager.

However, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to specification 6.1.

j 6-8

RECORDS The PORC shall maintain wvitten minutes of each PORC meeting including expedited meetings that, as a minimum, document the result of all PORC activities performed under the responsibility and authority provisions of these technical specifications.

Copies shall be provided to the Site Director and the Nuclear Safety Review Board.

6.5r2 NUCCEAR SAFETY REVIEW BOARD FUNCTION 6.5,2.1 The NSRB shall function to provide independent review and audit cognizance of designated activities in the areas of:

a, Nuclear power'lant operations b.

Nuclear engineer ing c.

Chemis t;r y and radiochemistry d.

Metellurgy e.

Instvumentation and control E.

Radiological safety g.

Mechanical and electrical engineering, and h.

Quality assurance pvact.ices CQMPOSlTION 6.5:2.2 The NSRB shall be composed of at least five members, including the Chairman.

Membevs of the NSRB may be fvom the Office of

'uclear Power oi other TVA organizations, or exLernal to TVA.

0 6-9

QUALXFICATIONS 6.5.2.3 The Chairman,

members, alternate members of the NSRB shall be appointed in writing by the Manager of Nuclear Power and shall have an academic degree in engineering or a physical science

field, or. the equivalent; and in addition, shall have a minimum of 5 years technical exper.ience in one or more ar.eas given in 6.5.2.1.

No more than two alternates shall participate as voting members in NSRB activities at any one time.

CONSULTANTS 6.5.2.4 Consultants shall be ut.ilized to provide expert advice as determined by the NSRB.

MEETING FREQUENCY 6.5.2.5 The NSRB shall meet at least once per six months, QUORUM 6.5.2.6 The minimum quorum of the NSRB necessary for the performance of the NSRB review and audit functions of these technical specifications shall consist of mor:e than half of the HSRB membership or at least Cive members, whichever is greater.

The quorum shall include the Chairman or his appointed alternate and the NSRB members including appointed alternate members meeting the requirements of 6.5.2.3.

No more than a minority of the quorum shall have line responsibility for operation of the unit.

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6.5.2.7 The NSRB shall review:

a.

The safety evaluations for:

(1) changes to procedures, equipment or systems, and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.

b.

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CYR.

c.

Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

d.

Proposed changes to Technical Specifications or this Operating l.icense.

e.

Violations of Codes, regulations,

orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

f.

Significant operating, abnormalities or deviations from normal and, expected performance of plant equipment that affect nuclear safety.

g.

All Reportable Events h.

All recognized indications oE an unanticipated deficiency in some aspect of design or operation of structures,

systems, or components that could affect nuclear safety; and i.

Reports and meeting minutes of the PORC.

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AUDITS

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".,'",5.5.2.8 Audits of unit activities shall be performed under.

the cognizance of th'e HSRB.

These audits shall encompass:

a.

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

b.

The performance, tvaining and qualifications of the entire plant, staff at least once per 12 months, c.

The results of actions taken to correct deficiencies occurring in site equipment, structures, systems ov method of operation that affect nuclear safety at least once pev 6

months.

d.

The pevformance of activities required by the Operational Quality Assurance Program to meet the critevia of Appendix B,

10 CFR Part 50, at least once pev 24 months.

e.'he Site Radiological Emergency Plan and implementing procedures at least once every 12 months.

f.

The Plant Physical Security Plan and implementing procedures at least once every 12 months.

g, Any other area of site operation considered appropriate by the NSRB or the Manager of Nuclear Power.

h.

The fire protection programmatic controls including the implementing proceduves at least once per 24 months.

6-12

4 i.

An independent fire protection and loss prevention program inspection and audit shall be performed annually utilizing either qualified offsite license personnel or an outside fire protection firm.

j.

An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at intervals no greater than 3

I years.

k.

The Radiological Environmental Monitoring program and the results thereof at least once per 12 months.

1.

The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 4.15, December 1977 at least once every 12 months.

m.

The performance of activities required by the Safeguards Contingency Plan to meet the criteria of 10 CYH 73.40(d) at least once every 12 mo>xths.

AUTIIORITY 6.5.2.9 The NSRB shall report to and advise the Manager of Nuclear Power on those areas of responsibility specified in Specifications 6,.5.2.7 and 6.5.2.8.

RECORDS 6.5.2.10 Reports of activities shalL be prepared,

approved, and distributed as indicated below:

6-13

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a.

Minutes of each NSRB meeting shall be prepared, approved and forwarded to the 'Manager of Nuclear Power within 14 days following each meeting.

b.

Reports of reviews encompassed by Section 6.5.2.7

above, shall be pxopared, approved and forwarded to the Manager of Nuclear Power within 14 days following completion of the xeview.

c.

Audit reports encompassed by Specification 6.5.2.8

above, shall be foxwarded to the Manager. of Nuclear Power and to the management.

positions responsible for the areas audited within 30 days aftex'ompletion of the audit.

6.5,3 TECHNICAL REVIEW AND APPROVAL OF PROCEDURES ACTIVITIES 6.5.3.1 Procedures required by Technical Specification 6.8.1.1 and other procedures which affect plant nuclear safety, and changes (other than editorial or typogx'aphical changes)

thereto, shall be prepaxed, reviewed and approved.

Each pxocedure ox procedure change shall be reviewed by an individual other than the preparer.

The reviewer may be from the same organization ox from a different organization.

Procedures other than Site Director Standard Pxactices will be approved by the responsible group/section supervisor, ox'pplicable plant superintendent..

6-14

6.5.3.2 Proposed changes or modifications to plant nuclear safety-related structures, systems and components shall be

'eviewed as designated by the Plant Manager.

Each such modification shall be reviewed by an individual/group other than the individual/group which designed the modification, but who may be from the same organization as the individual/group which designed the modification.

Proposed modifications to plant nuclear safety-related structures, systems and components shall be approved by the Plant Manager, prior to implementation.

6.5.3.3 Individuals responsible for reviews performed in accordance with 6.5.3.1 shall be members of the site supervisory staff previously designated by the Plant Manager.

Each such

'eview shall include a determination of whether or not additional, cross-disciplinary, review is necessary.

If deemed necessary, such review shall be performed by review personnel of the appropriate discipline.

6.5.3.4 The Plant Manager shall approve all administrative procedures requiring PORC review prior to implementation.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS; a.

The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each REPORTABLE EVENT shall be reviewed by the PORC and the results of this review shall be submitted to the NSRB and the Site Director.

6-15

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

a.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The Manager of Nuclear Power and the NSRB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

A Safety Limit Violation kepoaL shall be prepared.

The report shall be reviewed by the PORC.

This repoc.t 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.

c.

The Safety Limit Violation Report shall be submitted to the Commission, the NSRB and the Manager of Nuclear Power within 14 days of the violation.

d.

Critical operation of the unit shall not be resumed until author ized by the Commission.

6-16

6.8 PROCEDURES/INSTRUCTIONS AND PROGRAMS 6.8.1 PROCEDURES 6.8.1.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a.

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

b.

Limitations on the amount of overtime worked by individuals performing safety related functions in accordance with NRC Policy statement on working hours (Generic Letter No. 82-12),

c.

Surveillance and test activities of safety related equipment.

d.

Security plan implementation.

e.

Emergency plan implementation.

f.

Fire Protection Program implementation.

6.8.1.2 Each administrative procedure reguired by Section 6.8.l.l.a.

shall be reviewed by PORC and all other procedures reguired by Section 6.8.l.l.a. shall be reviewed in accordance with Section 6.5.3.

6-17

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6.8.1.3 Temporary changes to procedures of Specification 6.8.1.1 may be made provided:

a.

The intent of the original procedure is not altered; b.

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

The change is documented, reviewed by the PORC and approved by the Plant Manager within 14 days of implementation, for changes in administrative procedures requiring PORC review.

d.

The change is documented, reviewed per specification 6.5.3, and approved by the responsible group section supervisor within 14 days of implementation, for changes to procedures other than administrative procedures.

DRILLS 6.8.2 Drills on actions to be taken under emergency conditions involving release of radioactivity are specified in the Radiological Emergency Plan and shall be conducted annually.

Annual drills shall also be c'onducted on the

'ctions to be taken following failures of safety related systems or components.

RADIATION CONTROL PROCEDURES 6.8.3 Radiation Control Procedures shall be maintained and made available to all station personnel.

These, procedures shall show permissible radiation exposure and shall be consistent 6-18

with the requirements of 10 CFR 20.

This radiation protection program shall be organized to meet the requirements of 10 CFR 20 except in lieu of the "control device" or "alarm signal" required by paragraph 20.203 (c) of 10 CFR 20.

6.8.3.1 Each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a C

Radiological Work Permit.~

Any 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 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 after the dose rate lev'el in the area has been 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 surveillance at the frequency specified by the facility Health Physicist in the Radiological Work Permit.

'.8.3.2 Each high radiation area in which the intensity of radiation is greater than 1,000 mrem/hr shall be subject to the provisions of (1) above; and, in addition, access to the source and/or area 6-19

shall be secured by lock(s).

The key(s) shall be under the administrative control of the shift engineer.

In the case of a high radiation area established for a period of 30 days or less, direct surveillance to prevent unauthorized entry may be substituted for permanent access control.

<<Health Physics personnel, or personnel escorted by Health Physics personnel, in accordance with approved emergency procedures, shall be exempt from the RWP issuance reguirement during the performance of their assigned radiation protection

duties, provided they comply with approved radiation protection procedures for.entry into high radiation areas.

6.9 REPORTING RE UIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting, requirements of'itle 10, Code of Federal Regulations, the following identified.

reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.

6.9.1.1 STARTUP REPORT a.

A summary report of plant star tup 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 6-20

performance of the plant.

The report shall address each of the tests identified in the FSAR and shall include a

description of the measured values of the operating 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.

b.

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

Xf the Startup Report

,does not cover all three events (i.e., initial criticality, completion of startup test program, and resumptio'n or, commencement of commercial power operation),

supplementary reports shall be submitted at least every three months until all three events have been completed.

6.9.1.2 ANNUAL OPERATING REPORT+

a.

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mremlyr and their associated man rem exposure according to work and job functions, '<<e.g.,

reactor operations and surveillance,

'nservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The dose assignment to various duty functions may be estimates based 6-21

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on pocket dosimeter, TLD, or film badge measurements.

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

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

b.

Any mainsteam relief valve that opens in response to reaching its setpoint oq due to operator action to control reactor pressure shall be reported.

6.9.1.3 MONTHLY OPERATING REPORT Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Office of Inspection and Enforcement, U.S. Nuclear Regulatory Commission, Vashington, D.C. 20555,, with a copy to the Regional Office, to be submitted no later than the fifteenth of each month following the calendar month covered by the report.

A narrative summary of operating experience shall be submitted in the above schedule.

6.9.1.4 REPORTABLE EVENTS Reportable

events, including corrective actions and measures to prevent reoccurrence, shall be reported to the NRC in accordance with Section 50.73 to 10 CFR 50.

<<A single submittal may be made for a multiple unit station.

<<This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.

6-22

0 0

6.9.1.5 RADIOACTIVE EFFLUENT RELEASE REPORT<

A report on the radioactive discharges released from the site during the previous 6 months of operation shall be submitted to the Director of the Regional Office of Inspection and Enforcement within 60 days after January 1 and July 1 of each year.

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents released and solid waste shipped from the plant as delineated in Regulatory Guide 1.21, Revision 1, "Heasuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants" with data summarized on a quarterly basis following the format of Appendix 8 thereof.

The report shall include a summary of the meteorological conditions concurrent with the release of gaseous effluents during each quarter as outlined in Regulatory Guide 1.21, Revision 1, with data summarized on a quarterly b'asis following the format of Appendix 8 thereof.

Calculated offsite dose to humans resulting from the release of effluents and their subsequent dispersion in the atmosphere shall be reported as recommended in Regulatory Guide 1.21, Revision l.

6.9.1.6 SOURCE TESTS Results of required leak tests performed on sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.

• A single submit.tul may bo mado for a multipl.e unit station.

The'ubmittal should combine those sections that are common to all units at the station.

6-23

6.9.2 S ecial Re orts Reports on the following areas shall be submitted in writing to the Director of Regional Office of Inspection and Enforcement:

1.

Fatigue Usage 6.10.l.q Annual Operating Report 2.

Relief Valve Tailpipe 3.2.F Within 30 days after inoper-ability of thermocouple

'and acoustic monitor on one valve.

3.

Seismic Instrumentation Inoperability 3.2.J.3 Within.10 days after 30 days of inoperability.

4.

Meteorological Monitoring 3.2.I.2 Instrumentation Inoperability Within 10 days after 7 days of inoperability.

5.

Primary Containment Integrated Leak Rate Testing 4.7.A.2 Within 90 days of completion of each test.

6.

Data shall be retrieved from all seismic instruments actuated during a seismic event and analyzed to determine the magnitude of the vibratory ground motion.

A Special 6-24

Report shall be submitted within l0 days after the event describing the magnitude, frequency spectrum, and resultant effect upon plant features important to safety.

7.

Secondary Containment Leak Rate Testing" 4.7.C.

Within 90 days of completion of each test.

~Each integrated leak rate test of the secondary containment shall. be the subject of a summary technical report.

This report should include data on the wind speed, wind direction, outside and inside temperatures during the test, concurrent reactor building pressure, and emergency ventilation flow rate.

The report shall also include analyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.

6-25

6.10 STATION OPERATING RECORDS AND RETENTION 6.10.1 Records and/or logs shall be kept in a manner convenient for

~ review as indicated below; a.

All normal plant operation including such items as power level, fuel exposure, and shutdowns b.

Principal maintenance activities c.

Reportable Events d.

Checks, inspections,

tests, and calibrations of components and systems, including such diverse items as source leakage e.

Reviews of changes made to the procedures or equipment or reviews of tests and experiments to comply with 10 CFR 50.59 f.

Radioactive shipments

.g.

Test results in units of microcuries for leak tests performed pursuant to Specification 3.8.D 6-26

h.

Record of annual physical inventory verifying, accountability of sources on record i.

Gaseous and liquid radioactive waste released to the environs Offsite environmental monitoring surveys k.

Fuel inventories and transfers 1.

Plant radiation and contamination surveys

'm.

Radiation exposures for all plant personnel n.

Updated, corrected, and as-built drawings of the plant o.

Reactor coolant system inservice inspection p.

Minutes of meetings of the NSRB.

q.

Design fatigue usage evaluation Monitoring and recording requirements below will be met for various portions of the reactor coolant pressure boundary (RCPB) for which detailed fatigue usage evaluation per the ASME Boiler and Pressure Vessel Code Section III was performed for the conditions defined in the design specification.

In this plant, the applicable codes require fatigue usage evaluation for the reactor pressure vessel only.

The locations to be monitored shall be:

1.

The feedwater nozzles 2;

The shell at or near the waterline 3.

" The flange studs 6-27

0

~,

Transients that occur during, plant operations will be reviewed and a cumulative fatigue usage factor determined.

For transients which are more severe than the transients evaluated in the stress

report, code fatigue usage calculations will be made and tabulated separately.

In the annual operating report, the fatigue usage factor determined for the transients defined above shall be added and a cumulative fatigue usage factor to date shall be reported.

When the cumulative usage factor reaches a value of 1.0, an inservice inspection shall be included for the specific location at the next scheduled inspection (3-1/3-year interval) period and 3-1/3-year intervals thereafter, and a subsequent evaluation performed in accordance with the rules of ASME Section XI Code if any flaw indications are detected.

The

.results of the evaluation shall be submitted in a Special Report for review by the Commission.

6.10.2 Except where covered by applicable regulations, items a through h above shall be retained for a period of at least 5 years and items i through q shall be retained for the life of the plant.

A complete inventory of radioactive materials in possession shall be maintained current at all times.

1.

See paragraph N-415.2, ASME Section XII, 1965 Edition.

6-28

C1 0

GENERAL ttM)AGER HANAGER OF NUCLEAR POWER DEPUTY HANAGER NUCLEAR SAFETY REVIEW BOARD HAtfAGERt t'tUCLEAR OP E POTIONS HP2fAGER QUALITY ASSURANCE SITE DIRECTOR BFrtP SITE DIRECTOR SQNP SITE DIRECTOR WBNP DIRECTOR OF NUCLEAR SERVICES FIGURE 6.~-1 OFFSITE ORGANIZATION FOR FACILI'TY.HANAGEttENT AND TECllNICAL SUPPORT I

Site Director KunctianalReuortfna fan ager quality Assurance Modifications Manager Site Services Manager Design Service Manager Quality Assurance Plant Manager Plant Compliance Operations Engineering Superintendent Maintenance Superintendent Operations Engineering Health Physics Assistant Operations

.roup Supervisor Shift Crews

- Shift Engineer ASE

- UO AUO FIGURE 6.2-2 FACILITY ORGANIZATION

PROPOSED CHANGES VNI'i'

Section B.

Core Monitoring

~Pa e No.

305 C;

Spent Fuel Pool Water 306 D.

Reactor Building Crane 307 E.

Spent Fuel Cask 307 F.

Spent Fuel Cask Handling-Refueling Floor 308 3.11/4.11 Fire Protection Systems 315 A.

High Pressure Fire Protection System B.

CO Fire Protection System 315

,319 C.

Fire Detectors D.

(Deleted)

E.

Fire Protection Systems Inspection F.

(Deleted) 320

'3n 322 322 G.

Air Masks and Cylinders 323 H.

Continuous Fire Watch 323 5.0 I.

Open Flames, Welding and Burning in Cable Spreading Room I

Major Design Features the 323 330

~ ~

5.1 Site Features 5.2 Reactor 330 330 5.3 Reactor Uessel

~

~

~

~

~

~

~

~

~

~

~

~

~

330 5.4 Containment

~

~

~

~

330 5.5 Fuel Storage

~

~

~

~

~

~

~

~

~

~

330 5.6 Seismic Design 331

0

LIST OF TABLES Cont'd Table 4.2.F Title Minimum Test and Calibration Frequency for Surveillance Instrumentation

~Pa e No.

105 4.2.G'urveillance Requirements for Control Room Isolation Instrumentation 106

4. 2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation 107

4. 2.J 3.5-1 3.5.I Seismic Monitoring Instrument Surveillance Minimum RHRSW and EECW Pump Assigpment MAPLHGR Versus Average Planar Exposure 108 152a 171, 172,

172a, 3.7.A 3.7.B 3.7.C 3.7.D 3.7.E Primary Containment Isolation Valves 250 Testable Penetrations with Double 0-Ring Seals 256 Testable Penetrations with Testable Bellows 257 Air Tested Isolation Va'ves 258 Primary Containment Isolation. Valves which Terminate below the Suppression Pool Water evel L

262

, 3.7.F 3.7.G 3.7.H Primary Containment Isolation Valves Located Pzi Water Sealed Seismic Class 1 Lines Deleted Testable Electrical Penetrations 263 264 265 4.8.A 4.8 '

Radioactive Liquid Waste Sampling and Analysis 287 Radioactive Gaseous Waste Sampling and Analysis 288 4.9.A.4.C Voltage Relay Setpoints/Diesel Generator Start 298a 3.11.A

6. 2.A Fire Protection System Hydraulic Requirements 324 Minimum Shift Crew Requirements

~

~

6 3 vii

LIST OF ILLUSTRATIONS Figure 2,1.1

?.1-2 4.1-1 Title

~Pa e lio.

Graphic Aid in the Selection of an Adequate Interval Between Tests 49 APRN Flow Reference Scram and APRH Rod Block Settings 13 APRH Flow Bias Scram Ys. Reactor Core Flow 26 4.2-1 3.4-1 3.4-2 System Unavailability Sodium Pentaborate Solution Requirements Sodium Penta bora te So 1 u tion Requirements 119 Vo 1 ume Content ra tion

~

~

~

~

~

~

~

~

o 130 Temperature 139 3.5.K-1 3.5.2 3.6-1 3.6-2 MCPR Limits.

Kf Factor

~

Hinimum Temperature

'F Above Change in Transient

. Tempera ture Change in Charpy Y Transition Temperature Vs.

Neutron Exposure 172b 173 l

194 195 l' 6.2. 1 6.2

~ 2 Offsite Organization for Facility Management and Technical Support Facility Organization.

6-30

LL'BITING CQ'A)ITIORS FOR OPERATION 3iZiH Flood Protection SURVEILLANCE RZOtJIR~S 4i2.H FIood Protection The.unit shall be shutdown and placed in the cold condition vbea Mheeler Reservoir lake stage rises to a level such thee vater fzon the reoervoir bagiao to run across the punpiag atatioa deck at elevatioa 565.

Surveillance shall be pcrforaed on the instrumentation that monitors the reservoir level as indicated ia Table 4.Z.H.

Requircaents for instrunentation thee monitors the reservoir level io given ia Table 3.2 H.

3.2 I Meteoroloaical Monitorin Instn ~encation 4.2.I Meteo olo ical, Monitorin Instrumentation The meteorological monitoring instru-mentation listed in table 3.2.I sha11 be operable at a11 imes.

l.

Mith the number of operable meteorological monitoring channels less than required by table 3.2.I, restore the inoperable channel(s) to operable status within 7 days.

Each meteorologicaJ.

moni oring instn-en'hannel sha11 be demonstrated operable by the performance of the CvMCtEL Ch:-CK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the CKVQfr CALIBRATIONat least once each 6 months.

2.,

Mith one o. more of the meteoro-logical monitoring channels inoperablc for more t~mn 7 days, prepare and subcLit a Special Rcport to the Commission, pursuant to specification 6.9 '

within the next 10 days outl'ing the cause of the ma1function "nd the plans for restoring the system to operable status.

53

5

~

If'.ITIfi0 CO!lDITIOiiS FOR OKRATIOiI 3.2.J Seismic Honitorins Instrumentation SURVEILLAhCE REQUIRE!KfPl'S 4.2.J cismic Honitorina Instrumcntatfon l.

The seismic monitoring instruments listed ir. table 3.2.J shall be operablc at all times.

2., 4ith the number of seismic monitoring instruments less than the number listed in t"ble 3.2

~ J, rrstorc thc inoperable instr>>mc>>t(::) tu oped ~Lhlr status viL,hi>> 30 day>>.

3.

>(ith one or more of the instruments listed in table 3.2.J inoperable f'r more than 30 days, submit a Special Report to the Co@mission pursuant to specification 6.9.2 within thc next 10 days describiny, the cause of thc malfunction and plans for restoring the instruments to operable statu l.

Each of the seismic munit,orinq in "ru-mcnts shall be demonstrated operab'e by performancr of tc ts nt the frequencies listed in table 4.2.J.

2.

Data shall bc retrieved from all sci"mir instr>>ments art>>ntcd cl>>r in~. >>>>a l "..mi<<<<vc>>l >>>>d>>>>a fy::cu to dcLc~minc thc rnaf;>>it>>dc nf thc vibrat.o.y g~u in l motion.

A Special Report shall be submitted to the Commissio>> p>>rsuant to specif'cation 6.9 2

uithin 10 day" describing the magnitude, frcquen~y spectr>>m, and ren>>lt.n>>L effect

>>pnn pl>>>>t.

feature

'mportant t,o safct.y.

PROPOSED CHANGES UNlT 2

0

Section

~PS e No.

B.

Core Monitor ing............

305 C.

Spent Fuel Pool Mater 305 D.

Reactor Building Crane 307 E.

Spent Fuel Cask 307 V

. F. Spent Fuel Cask Handling-Refueling Floor

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

308

~ e t3. 11/4. 11 Fire Proteotion Systems 315 A.

High Pressur e Fire Protection System 315 B.

C02 Fire Protection System,.

C.

Fire Detectors

~

D.

(Deleted)

~

~

~

~

~

~

~

~

~

~

~

E.

Fire Protection Systems Inspection 319 320 Sn 322 F.

(Deleted)

~

~

~

~

~

SSS

~:

G.

Air Masks and Cylinders 323 H.

Continuous Fire Watch 323 I.

Open Flames, Welding, and Burning in the Cable Spreading Room 323

.5.0 Ma)or Design Features

~

~

~

~

~

~

~

~

~

~

~

330 5.1 Site Features........

~

~

~

~

~

330 5.2 Reaotor

~

330 5.3 Reactor Vessel 330 5.4 Containment 330 5.5 Fuel Storage

~

~

~

~

~

~

~

~

~

~

~

~

~

5e6 Seismic Design

~

~

~

~

~

~

~

~

~

~

~

~

330 331 iv

LIST OF TABLES Cont'd Table

4. 2.F Title Minimum Test and Calibration'requency for Surveillance Instrumentation

~Pa e Ne.

105

4. 2.G Surveillance Requirements for Control Room Isolation Instrumentation 106

4. 2.H Minimum Test and Calibration Frequency for Flood Protection Instrumentation l(7

4. 2.J 3.5-1 Seismic Monitoring Instrument Surveillance Minimum RHRSW and EECW Pump Assignment 108 152a

'3.5.I MAPLHGR Versus Average Planar Exposure

171, 172,

172a, 3.

7.A'.7.B 3.7.C 3.7.D 3'.E Primary Containment Isolation. Valves which Terminate below the Suppression Pool Water Level 262 Primary Containment Isolation Valves........

250 Testable Penetrations with Double 0-Ring Seals 256 Testable Penetrations with Testable Bellows 257 Air Tested Isolation Va'ves 258 3.7.F 3.7.G 3.7.H Primary Containment Isolation Valves Located An Water Sealed Seismic Class 1 Lines

'Delet'ed Testable Electrical Penetrations 263

%64

='65 4.8.A 4 '.B Radioactive Gaseous Waste Sampling and Analysis 288 Radioactive Liquid Waste Sampling and Analysis 287 4.9.A.4.C Voltage Relay Setpoints/Diesel Generator Start 298a 3.11.A Fire Protection System Hydraulic Requirements 324

6. '2. A Minimum Shift Crew Requirements 6-3 Uii

LIST OF I USTRATIONS F~iure 2.1,1 Title APRN Flow Reference Scram and APRH Rod Block Se'ttin9$

~

~

~

~

~

~

~

~

~ka

~ ka.

2.1-2 4.1 1

Graphic Aid in the Selection of an Adequate Interval Between Tests 49 APRH Flow Bias Scram Vs k Reactor Core F low....

26 4.?-1 3.4 1

2 System Unavailability Sodium Pcntaborate Solution Volume Concentration Requirements Sodium Pentaborate Solution Temperature Requirements 119 136 139 3.5.K-1 MCPR Limits 172a 3.5.2 Kf I'actor

~

~

~

~

~

~

~

~

~

173

3. (k-I 3.6-2 Hinimum Temperature

'F Above Change in Transient Temperature k

~

~

~

~ ~

~

~

~

~

~

~

~

~

~

~

~

~

~

a

]94 Change in Charpy V Transition Temperature Vs.

Neutron Exposure 195 6.2. I Offsite Organization for Facility Management and Technica'.

Support 6-29 6.2.2 Facility Organization 6-30

iL~~~~Z~" q~ ~gZ",Zg!tS FOq OPvP~<TZC'.I iUnVEX I KHCZ 'REOUI3..!':-.i Z3 FIead protection 4.2.!! Flood Protection shale bo shutdown and placed in tha cold'ondi iou uhan b'heeler Reaarvoir labe ata"a riaea to a level ouch that cater

."rou tho :eacrvoir begins to run acroaa tha puupiatI csatan datk at elevation 5G5.

4 Requirauanta for inatruuantation that eonitora tha r aarvoir lavol

" io Given in Tabla 3.2 ~II.

"u~eillanee oh:.11 ba pe::domed on tha inatruuantation th"t

~nitora tha raaarvoir lavel na indicated iu Yabla 4.2.)!.

3.2.I, Meteorological Vonitorina In t~-cn-~t on 4.2.I i I!ctcorolo~~ica3. Honitnrln.-

Instru~entation

~i..e;"et=oroiogical ~nitoring instru-nenta ion 1'sted in table 3.2.I shall bc operabl'e at a3'~ s.

., 'Zith 'hc.".-." " of oper blc aeteorolouical ~nitoring channel

='ess than rcqui-cd by table 3.2.I, cstore

? e inoperable channel( -)

o o"erablc status within 7 day.

Each ncteorological moni or'r~~ 'instru=eat channel shall be demonstrated opcrab'c by thc pcrfomncc of thc Q~VPiFL CHECK at least once per 24 hou s and th" Q.:QL'~r CALXPBATIONat least once each 6 r.'onths.

2..-Mith o.".c o" narc of he nctcoro-lo 'c"1 =onitoring c"annals inoperablc or ="re t~~s 7 'days, pre"are ard sub&t a Special Rcport o the Co~anion, purst't o

pcuiQ.cation 6. 9 within the

,:...n~ '0 days outL'n'~ the cau"c

o. thc m urct'cn and the plan" or ". coring the system to operablc r,~a~us

• 53

~ il Set

e. '..or It rial Instr'i rnntation tr.<?.J

..cxsntc rt.inr tortrr'r ln.t-"rncr.t'<-.:."n

!re seis=.ic =.onitorinc ins rurents

-'.'ste'n

" 3'c 3.2.J

- hall bc oper ble at all t'.-,.cs.

2.

'~'ith the n:=,ber of seismic r.".onitorin!;

stcd in table 3.2.J, restore t.hc inoperab',e in.".'<r~crrt(r:) tv oper nlilc'=.".

"n= or."..ore of the in" rc:ents

s.

d in table 3.2.J inoperable t'or t:orc han 30 days, ub~it a Special Peport to -...e Cc:".~ission pursuant to specification 6.9.2 vithin the next 10 cays dc cribiny t.he cause of the ralfunction a.".a plans for restorint; nst"u.".ents "o opcrab)e statu B>ch of tlrc sci rrric morrisorin: 'n".ru-ncnts shall be d<<r"onstr'a ed ope.ab:e by perfor".<rane<<of tc"t.s at t!re fs couencics li"ted in 'abl<< '<.P.J.

<<ta shal be

. <<, ieve<

scis.".ri<

in::- r<,i< nt" n<

ante<!

<r>>ri>>rt.i >>ci"..<<ii<'vcrrt.:<<<<1 <illa t y

cur t.o <(etc< r,:i r<<r t.!rc rrattnitr<<i<: <if t!<c vibr"tory t;r v>>>>r rot.ion.

>t:lpcc'a; Report shat\\

te sub~itted t.<.. the Corrnissinn ptrrsuant.

to specification 6.9 2

vithin '0 days dcscribinf the nat,nitude, frcqucn< y spcctrrn, and result:rr<L effect.

rrpon pln>>t feat;ur c'npor tarrt.

t.o sni'cty.

PROPOSED CHANGES UNIT 3

0

Sect~@

3. 10r4. 10 C..Operation in Cold Shutdcnm Core Alterations

~ae Ho.

326 331 Bo C

D ~

Refueling Znterlvcks Core Monitoring Spent Fuel Pool Mater Reactor Building Crerne Spent Fuel Cask Spent Fuel Cask Handling-Refueling Floor 331 336 337 338 339 339 3 ~ 11r4 ~ 11 Fire Protection Systems High Pressure Fire Protection System 347 347 Bi Ci CO ~ Fire Protection System Eire Detectors 351 352 D.

Deleted E.

Fire Protection Systems Inspections Deleted G.

Air Masks and Cylinders H.

Continuous Fire Watch 353 354 355 355 5

0 I.

Open Flames, Welding, and Burning in the Cable Spreading Room Major Design Features

5. 1 Site Features 5.2 Reactor 5.3 Reactor Vessel 5.4 Containment 5.5 Fuel storage 5.6 Seismic Design 355 360 360 360 360 360 360 361

4.2.F.,

4.2.F Hinimum Test and Calibration Frequency for Dryvcll Leak Detection Inscrumentation Hinimum Test and Calibration Frequency for Survcillan Instrumencacion or urvcillancc 101 102

4. 2.C Surveillance Requiremencs for Concroi R

I olation instrumentation Room Isolation 103 4,2.H C,2,J 3.3.-1 3.3.1 Hinimum Test and Calibration Frequenc for F Instrumentation n

requency for Flood Protection Seismic Honitoring Instrument Surv ill urve ance Requiremencs Hinimum RHRSM and EECM Pump Assignment HAPLHCR vs. Average Planar Exposure 104 105 156a 18 1 s 182 ~

182ay 182b

~ 3. 7.A Primary Concainment Isolation Valves 262 3.2mB 3.7.C

3. >.0 Tescable Penecrations vich Double 0-Ring Seals Testable Penecrations vich Testable Bellovs Air Tested Isolacion Valves 268 269 270 3.7 ~ 8 Primary Concainmenc Isolation Valves vhich Term Suppression Pool Macer Level c

crminacc Belov chc 279 3oj.y Primary Concainmenc Isolacion Valves Locaced 1

M Seismic Class 1 Lines a

e n

accr Sealed 280 3.7.0 3 ~ 2.H

'eleced Testable Electrical Penecracions 283 4.8.A C.B.B

,4.9.A;4.c 3.11.A Radioactive Liquid Masce Sampling and Analysis Radioactive Caseous Masts Sampling and Analysis Volcage Relay Setpoincs/Diesel Ceneracor Start Fic'e Protection System Hydraulic Rcquircmencs Hinimum Shift Crcv Requirements 310 311 327 355a 6-3 vii

LIST OF Il LUSTRATIONS Figure 2.1-1 2 ~ 1 2

4. 1-1 4 2-1

3. 4-1

3. 4-2 Title APRM Flow Reference Scram and APRM Rod Block Settings APRM Flow Bias Scram Relationship to Normal Operating Conditions Graphic Aid in the Selection of an Adequate Interval Between Tests Sys tem Una vai 1ability Sodium Pentaborate Solution Volume Concentration Requirements Sodium Pentaborate Solution Temperature Requirements

~ae 25 48 117 141 142 3.5.K-l

3. 5. 2 3.6-1 3i 6-2 MCPR Limits Kg Factor vs. Percent Core Flow Temperature-Pressure Limitations Change in Charpy V Temperature vs, Neutron Exposure 182c 183 207 208 6.2.1 Offsite Organization for Facility Management and Technical Support 6-29 6.2.2 Facility Organization 6-30 viii

LZHZTZNG CONDITIONS FOR OPERATION SURVEILLANCE REQUZREl'KNTS 3.2 PROTECTXVF. XN6TBUNENTA'~ON 4

2 PROTRCTXVE XNSTRUNENTATXON With one or nore of the meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission, pursuant to specification 6 ~9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the system to operable status 55

I

LZNITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3,2 eAol'EcTIVE Ns

~ RUMP2iTATIQN 4

2 PROTECT XVE INSTRUMENTATION Seismic !!onitorin inatrumentatio Seismic Monitorin Instrumentation 20 3 ~

The seismic monitoring instruments listed in Table 3.2.J shall be operable at all times+

With the number of seismic monitoring instruments less than the number listed in, Table 3.2 J, restore the inoperable instrument(s) to operable status within 30 days.

with one or more of the instruments listed in Tabla 3.2.J inoperable for more than 30 days, subbed.t a Special Report to the Commission pursuant to specification 6.9.2 within the next 10 days describing the cause of the malfunction and plans for restoring the instruments to operable status.

20 Data shall be retrieved from all seismic instruments an uatcd during a seismic event, and analyzed to determine the magnitude of the vibratory ground motion.

A Special Report. shall be submitted to the Commission pursuant to specification

6. 9.)

within

$ 0 days describing the magnitude, frequency spectrums and resultant, effect upon plant features important to safety.

1.

Each of t,he seismic monitoring instruments shall be demonstrated operable by performance of tests at the frequenciea listed in Table 4.2.J.

ENCLOSURE-2 DESCRIPTION AND JUSTIFICATION TVA BFNP TS 201 SUPPLEMENT 1

Description of Changes

(.1),

The amendment will adopt the Standard Technical Specifications (STS)

'method for page numbering and format in the Administrative Controls section.

Pages 332 thru 364 (unit 1),

333 thru 364 (unit 2) and 362 thru 394 (unit 3) are renumbered 6-1 thru 6-30.

i

'I (2) 'he Table of Contents

pages, (iv, v, vii, and viii) for each unit will be ievised to reflect the other changes in this amendment.

References,=

to section 6 are also revised on pages 53 and 54 of units 1 and 2 and on pages 55 and 56 of unit 3.

.(3)

Pages 6-2 and 6-3:

(a)

The note requiring the Shift Technical Advisor to be present at all times has been deleted since this requirement is sufficiently covered in Table 6.2.A.

(b)

The note (86) which required either the plant manager or a plant superintendent to have SRO training and the operations supervisor or the assistant operations supervisor to have an SRO license has also been deleted since the amended staffing requirements meet the guidance provided by NRC in STS and 10 CFR 50.54(m).

(c):

Por NRC comments a statement has been included that says, "In

addition, a person holding a senior operator license shall be in the control room for that unit whenever it is in an operational mode other than cold shutdown or refueling."

(d)

Also per NRC comment the requirement for a senior licensed operator in respect to fuel handling has been clarified.

(e)

A third comment by NRC is addressed by the addition of the, STS requirement to maintain a site fire brigade.

This change also includes the addition of tho STS note regarding the minimum requirements for the site fire brigade and health physics technician.

Finally the Minimum Shift Crew Requirements (Table 6.2.A) have been revised to meet that described in 10 CFR 50.54(m).

(4)

Pages 6-4'thru 6-9 describing the Plant Operations Review Committee and pages 6-14 thru 6-18 describing technical review and approval of procedures have been extensively revised.

The revisions to the current TS section 6.2.B (new section 6.5.1) changes the format of thissection'o allow PORC to function as a body or as individuals collectively when'he vote is unanimous, and-thru the use of subcommittees.

It also changes PORC review responsibilities to require it to focus on the safety related

issues, the emergency operating, instructions and the administrative procedures that control the technical and cross-disciplinary review of those written procedures covering safety related activities.

ENCLOSURE 2 (Continued)

The revision to the current TS section 6.3.A (new section 6.8.1) more clearly delineates, by reference to Appendix A of Regulatory Guide 1.33, thyrse procedures which are required to be established, implemented, and maintained for operation of the plant and deletes from PORC the review responsibility for those corporate site-level procedures issued by the site director.

The revision to the current TS section 6.3.B (new section 6.5.3) establishes the required technical and cross-disciplinary review and approval to support the changes in PORC review responsibilities.

It establishes plant manager control of the required revisions of proposed changes or modifications to nuclear safety related instructions l systems or components at BFN. It assigns the required technical and cross-disciplinary review responsibilities to members of the site supervisory staff designated by the plant manager and requires the responsible individuals to consider whether or not cross-disciplinary review is necessary.

In addition, it specifies those procedures requiring plant manager approval prior to implement.aLion.

Also, the NRC comment that PORC distribute written minutes of each PORC meeting to the Site Director and the Nuclear. Safety Review Board (NSRB) has been incorporated.

Pages 6-9 thru 6-14 describe the NSRB.

The change in TVA BFN TS 201 to replace the NSRB with the Nuclear Safety Staff (NSS) has been withdrawn.

Pages 6-15 and 6-16 describing actions to be taken in the event of a Reportable Event or a Safety Limit Violation has been revised to utilize the guidance provided by NRC in STS.

This change also satisfies an NRC comment on TS 201

~

(7)

Page 6-19 contains a minor change to show that BFN has changed the title of the Special Mork Permit to the Radiological Mork Permit.

(8)

Pages 6-20 thru 6-25 discuss reporting requirements.

(a) Footnotes 2 and 3 of the current TS section 6.7, which defined the

terms, "forced reduction in power" and "forced outage" are deleted by this amendment since they are not referenced anywhere in TS.

(b) The notes 1 and 2 are moved and properly referenced by the Annual Operating Report requirement and the Radioactive Effluent Release Report requirement as shown by NRC guidance provided in the STS.

ENCLOSURE (Continued) h I

(c) The reference to safet,y relief valves in the current TS on the Annual Operating Report has been changed Lo propevly reference only relief valves since "safety vaLves" were deleted from units 1, 2, and 3 TS by amendments 92, 85, and 51 r'espectively.

(d) The requirement for the Monthly Operating Report has been changed to.

allow 15 days for submission to NRC instead of the current requirement of 10 days.

This extension of 5 days is consistent with t,he guidance provided in STS.

(9)

Pages 6-27 and 6-28 have been revised to slightly modify the wovds of the current TS section 6.6.A.17 (new section 6.10.1.0) to reflect tho r'evised format and the curvent TS section 6,6.A.18 has been deleted since it was solely in reference to a paragraph 6.10 which was pveviously deleted by amendments 79, 75, and 48 for units 1, 2, and 3

respectively.

(10)

Pages 6-29 and 6-30,are the revised figuves for Offsite Organization and

Facility Organization.

The "Manager of Nuclear Power" shown on figure 6,2-1 replaced the position of "Manager of Powev and Engineering" shown'n this figure submit.ted by TS 201, dated September 27, 1984.

TVA request for technical specification change TS 174 and TS 174 Supplement

'l, which was superseded by TS 201, this posit.ion was shown as the Director, Division of Nuclear Power.

This figuve also shows Lhe Nuclear Saf'ety Review Board instead.of the Nuclear Safety Staff included in TS 201.

The figuve shows a change in the st,vucture of Quality Assurance.

Xt also shows the position of

Manager, Nuclear Opevations..

Reason for Change (1).and'(2)

Changing the TS format should make the Administrative Controls

.section easier to follow due to the more logical order in which each section appears and the better indexing and improved sect.ion headings should make finding needed information quicker and easier.

The standard page numbering syst.em will allow all units t.o have identical pages, Lhereby reducing administrative buvdon.

(3)

(a). Curvent TS are. redundant and overlyvestrictive by vequiring, the STA Lo be onsite at all times.

(b) Note 6 in current TS does not correspond t,o any vequirement in STS and since this section is being revised to comply wit.h standards, this note is being deleted.

(c thru f) 1hese new v'equirements ave in response to NRC conunents and to comply with STS and 10 CFR 50.54(m).

(5)

A.very large administrative burden to PORC has resulted from the current interpretation of review requirements for administrative procedures and procedural changes.

The proposed changes to technical specifications will allow the use of individual reviewers to reduce this burden.

The current requirements have reduced the time available for PORC to review and focus on significant safety issues, and limits supervisory time available for verifying that operating activities are conducted safely aqd in accordance with administrative controls.

Adding the distribution, of PORC meeting minutes is consistent with STS and in response to an NRC comment on TS 201.

The proposed reorganization of the NSRB to form the NSS has been cancelled.

(6)

Adopting STS language for actions to be taken in the event of a safety limit violation or reportable event will clarify these sections and satisfy an NRC comment on TS 201.

(7) 'Changing to a "Radiological Work Permit" will make BFN consistent with other TVA nuclear plants.

0 (8)

(a) Deleting notes that are not referenced in the current TS and are not in'STS will clarify the actual requirements by removing unused words.

'b)

Moving notes to the area where they are to be referenced and pt;operly noting the reference will clarify these requirements.

(c) Removing this reference to safety valves is correcting an error and does not actually change the requirement.

(d) Allowing 15 days to submit the monthly report will provide needed additional time for its preparation.

(9)

These changes are to improve the format and clarity of the

=Administrative Controls section.

(10)

These organizational charts are to reflect the TVA organization currently in effect.

Comment No.

1 from D.

B. Vassallo's June 2'I, 1985 letter to H. G. Parris was a recommendation that the offsite organization chart be expanded to show the function block of Manager-Radiological Health reporting to the-Director of Nuclear Services.

We do not believe that it would be beneficial to expand the offsite organization chart as recommended.

Our position is based partly on the fact that Sequoyah's latest administrative technical specification request did not include this

-block 'on the offsite organization chart.

Furthermore, expanding the chart would produce additional detail which would be subject to change.

Simply adding this block to the chalet will not improve nuclear safety.

Justification for (:hanges TVA is a corporate agency of the Federal Government whose major policies,

programs, and organizations are determined by a full-time, three member Board of Directors.

Members of the Board are appointed by the President and

0

confirmed by the Senate for nine-year terms.

The staffs of Browns Ferry Nuclear Plant within the Office of Nuclear Power is responsible for operating and maintaining TVA's Browns Ferry Nuclear Plant'.

The changes contained in this amendment request deal specifically and solely with the administrative controls required by TS.

Changing the Administrative Controls section to standard format with the standard page numbering system, deleting unnecessary

notes, deleting references to previously deleted requirements,

adopting, STS terminology, adding n'w minimum shift crew requirements, and updating the description of the TVA organization are ail within the guidance provided by NRC in STS and should result in increased clarity and usefulness and therefore, increased safety.

The changes to the description of PORC and technical review are consistent with the regulatory position of Regulatory Guj.de 1.33, revision 2, February

1978, and requires decisions affecting safety'o be made at the proper level of responsibility and with the necessary technical advice and review.

They do not degrade the effectiveness of the independent technical review.

Rather, these changes reinforce the technical and cross-disciplinary review and approval of the written procedures and permit PORC to focus its attention on the significant safety issues and the administrative controls important to the safe operation of the plant consistent with Regulatory Guide 1.33, revision 2.

The changes that reduce PORC review of procedures and modifications are consistent with recent NRC approved TS such as NUHEG 1042.

0

0 ENCLOSURE 3

Determination of No SigniEicant Hazards Consideration Description of Amendment Request The proposod amendment would revise the technical specifications (TS) of Browns Peery Nucleae Plant units 1, 2, and 3 to provide various improvements and clarificat,ions to the Administrative Controls seCtion.

The changes include:.

(l)

Revise section 6, Administrative Contvols, to incorporate the format of Standard Technical Specifications (STS).

This include" the use of tho STS page numbering system, revised index/table of contents and relocation of several requirements, revising, all t:he references within TS to sections in section 6, incorporating STS language for actions to be taken in the event of a saEety limit.'violation ov reportablo event, changing the title of the Special Work tIermit (SMP) to the Radiological Work Permit (RMP), and allowing 15 days fto submit t.he monthly report. to NRC as is allowed in STS.

(2)

Revise the minimum plant staffing requivemonts to cef lect the guidance.

of STS and 10 CFR 50.54(m).

This includes deletion of the note requieing that the Shift Technical Advisoc be onsite at all times, deletion of the note requiring sonioc operator license (SRO) training, foe managers, adding a note that requires an SRO to be in tho cont.eol room of a unit whenever it. is in an operat.ional mode other than cold shutdown oe refueling, clarifying the requirement Eoc an SRO to directly supervise fuel handling, addition of requirements Eoc the sit.e fire bcigade and the STS note regarding the minimum requivements for the fire brigade and health physics t.echnician, and finally to revise the table for minimum shift cc'ew requirements to reflect STS and 10 CER 50 '4(m).

(3)

Organizational changes within TVA have eesulted in revisions to the offsite and facility organization chart.s.

(4)

Deletion of unnecessary and/or obsolete notes and requirements.

Two footnotes at; the end of the Reporting Requirements soct.ion which defined the tevms, "focced outage" and "forced eeduction in power"'re deleted by this amendment.

since they are not referenced anywhere in these TS oe in STS,.

Two ot.hee notes which wore unreEevenced at the ond of this section were moved to t.he proper locations and referenced similar to STS.

The requirement for the Annual Operating Report specifies that opevation of the "safety/relief valves" should be included in the ropovt.

However, "safety valves" were deleted fvom the TS by previous amendments, thevefore, this refevence is being changed to only "relief valves".

Similarly the amendment. will delete a vequirement to maintain records "which ace covered under the provisions of parageaph 6.10,"

since paragraph 6.10 was deleted by previous amendments.

ENCLOSURE 3 (Continued)

(5)

Tfm section that. describes the Plant Operations Review Committee (PORC) has been revised to reflect a new emphasis on matters that relate to nuclear safety.

The PORC membership and meeting requirements have been more clearly defined and the responsibilities revised to reflect the new emphasis and the proposed change in the technical review of modiEi.cations and procedures.

The PORC section is also revised to reflect the new TVA organization and requirements of STS.

(6)

The technical review and approval of procedures and proposed modiEications has been changed to incorporate the use of members of the site supervisory staff designated by the plant manager to perform Lhe required technical and cross-disciplinary review responsibilities.

The procedures section has been revised to more clearly delineate those procedures which are required to be established, implemented, and maintained by referencing Appendix A of Regulatory Guide 1.33.

Basis for Proposed No Sigr>ificant Hazards Consideration Determination The Commission has provided guidance for the application of criLeria for no significanL hazards consideration determination by providing examples of amendmonts that are considered not likely Lo involve significant hazards considerations (48 YR 14870).

These examples include: "(i) - A purely administrative change to technical specifications:

for example a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.

(vi) -

A change which either may result in some increase to the probability or consequences of a previously analyzed accident or reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect Lo the system or component. specifi;ed in Lhe Standard Review Plan (SRP):

For example, a

change resulting from the application of a small refinement of a previously-used calculational model or design method.

(vii) A change to make a license conform to changes in the regulations where Lhe license change resulLs in very. minor changes Lo facility operations clearly in keeping with the regulations."

The proposed amendments concerning the adoption of STS format, organizational

changes, and deletion of unnecessary and/or obsolete notes and requirements,,

(items 1, 3, and 4) are administrative in nature and are therefore encompassed by example (i).

Tho changes proposed Eor the Minimum Plant Staffing, PORC responsibilities, and the Lechnical review and approval of procedures and modificaLions, (iLem 2, 5, and 6) are encompassed by example (vi) in that the revisions reflect the requirements established in the STS (NUREG-0123, Revision 3) as endorsed by chapter 16 of the Standard Review Plan.and other more recent TS issued by the

NRC, Eor example NUREG 1042.

They are also encompassed by examples (i) and (vii) in that they are administrative!programs and are partially the result of new regulations.

P

FNCLOSURE 3 (Continued)

The Commission has also provided standards for determining whether a

significant hazards consideration exists as stated in '10 CFR 50.92. (c).

A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the'roposed amendment would not:

(l) Involve a significant increase in the probability or consequences oE an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident pr.eviously evaluated; or (3) involve a significant reduction in a margin of safety.

The proposed changes will not significantly increase the probability or consequences of an accident previously evaluated or create the possibility of a new or different kind of accident from any jccident previously evaluated because no operability or surveillance requirqments for systems, structures or components used to terminate or mitigate accidents would be reduced and no equipment changes are involvod.

I The proposed changes will not involve a significant reduction in a margin of safety since the changes are administrative in nature and conform to HRC "guidance in STS and/or recently issued TS and for the same reasons as stated above.

Since the application for amendment involves proposed changes that are l

encompassed by the criteria for which no significant hazards consideration exists and are encompassed by the above

examples, TVA proposes to determine that the proposed amendments do not involve a significant hazards consideration.

V