ML18033A544

Proposed Tech Specs,Revising Section 6, Administrative Control to More Standardized Format as Detailed at Tva/Nrc Mgt Meeting on 890109

ML18033A544
Person / Time
Site:	Browns Ferry
Issue date:	02/01/1989
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML18033A543	List: ML18033A542 ML18033A544
References
NUDOCS 8902090203
Download: ML18033A544 (66)

Text

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6.

1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Operations Supervisor (or, during his absence from the control

room, a designated individual) shall be responsible for the control room command function.

A management directive,to this

effect, signed by the Site Director shall be reissued to all station personnel on an annual basis.

6. 2 ORGANIZATION 6.<. 1 OFFSITE AND ONSITE ORGANIZATIONS Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.'he onsi'te and offsite

'rganizations shall include the positions for activities affecting the safety of the nuclear power plant.

'a

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b.

C.

d.

Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions.

These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.

These requirements shall be documented in the FSAR'nd will be updated in accordance with 10 CFR 50.71(e).

The Senior Vice President, Nuclear Power, shall have corporate responsibility for overall plant nuclear safety.

This individual shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

The Plant Manager shall be responsible for overall unit safe operation, and shall have control over those onsite resources necessary for safe operation and maintenance of the plant.

The individuals who train the operating staff and those who carry out radiological control and quality assurance functions may report to the appropriate onsite manager;

however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

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6:2. 2 FACILITY STAFF l'.

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

b.

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

In addition, while the unit is not in a COLD SHUTDOWN CONDITION at least one licensed Senior Operator shall be in the control room; c.

A Radiological Control Techni'cian" shall be onsite when fuel is in the reactor; d.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling.

The assigned individual will have no other concurrent responsibilities during this operation; e.

A site Fire Brigade of at least five members" shall be maintained on site at all times.

The Fire Brigade shall'ot include the Shift Operations Supervisor and the other members of the minimum shift crew necessary for the safe shutdown of the unit nor any personnel required for other essential functions during a fire emergency.

"The Radiological Control 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 fill the required positions.

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Table 6.2.A Minimum Shift Crew Re uirements'osition Units in 0 eration T

e of License 0

1 24 3

Senior Operator'enior Operator Licensed Operators Additional Licensed Operators'ssistant Unit Operators (AUO)

Shift Technical Advisor (STA) 1 1

1 1

0 1

3 3

2 2

3 3

0 1

2 2

4 4

5 5

0 1

1 1

SRO SRO RO or SRO RO or SRO None None Radiological Control Technician 1

1 1

1 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.

b.

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.

c.

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

d.

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

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t 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics, NRC issuances, industry advisors, Licensee Event

Reports, and other sources which may indicate areas for improving plant safety.

COMPOSITION 6.2.3.2 The ISEG shall be composed of at least 3 dedicated full-time engineers located onsite.

These engineers will be supplemented by 2 full-time engineers shared among all TVA nuclear sites.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of plant activities to provide independent verification* that these activities are performed correctly and that human errors are reduced as much as practical.

AUTHORITY 6.2.3.4 The ISEG shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving plant safety to the Manager of Licensing and Regulatory Affairs.

6.2. 4 SHIFT TECHNICAL ADVISOR (STA) l 6.2.4.1 The STA shall serve in an advisory capacity to the shift supervisor on matters pertaining to the engineering aspects of assuring safe operation of the unit.

The STA 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 FACILITY STAFF VALIF ICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions and the supplemental requirements specified in Section A and C of Enclosure 1

of the March 28, 1980 NRC letter to all licensees, except for the Site Radiological Control Superintendent who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

An additional exception is that the SRO license requirement for the Operations Superintendent shall be met if ei'ther the Operations Superintendent or a manager reporting directly to the Operations Superintendent and responsible for the direction of Shift Operations holds a valid SRO license.

The holder of the SRO license shall direct the licensed activities of the licensed operators.

• Not responsible for sign-off function.'FN Units 1, 2 and 3

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6'. 4 TRAINING 6.4.1 A retraining and replacement training program for the facility operations staff shall be maintained under the direction of the Operations Superintendent and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Section A

and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience 6.5 REVIEN AND AUDIT 6.5.1 PLANT OPERATIONS REVIEH COMMITTEE (PORC)

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

COMPOSITION 6.5.1.2 The PORC shall be composed of the following individuals and designated managers (representative):

Chairman Member:

Member:

Member:

Member:

Member:

Member:

Member:

Plant Manager Operations Superintendent Radiological Control Superintendent Maintenance Superintendent Technical Support Services Superintendent Site Quality Representative Engineering Representative Plant Support Representative The Chairman shall designate the representatives in writing.

The Chairman and members shall meet the qualification requirements of ANSI Standard N18.1 1971.

ALTERNATES 6.5.1.3 Alternate chairmen and alternate members shall be appointed in writing by the Plant Manager;

however, no more than two alternates shall participate as voting members in PORC activities at any one time.

All alternate chairmen and alternate members shall meet the qualification requirements of ANSI Standard N18.1 1971.

MEETING FRE UENCY 6.5.1.4 The PORC shall meet at least once per calendar month and as convened by the PORC Chairman or his designated alternate.

QUORUM 6.5.1.5 The minimum quorum of the PORC necessary for the performance of the PORC responsibility and authority provisions of these technical specifications shall consist of the Chairman or his designated alternate and four members including alternates.

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

The PORC may delegate the performance of reviews to subcommittees or qualified reviewers but will maintain responsibility for these revi ews.

a.

C.

d.

e.

Review safety evaluations for:

(1) procedures and (2) change to procedures, equipment, systems or facilities to verify that such actions do not constitute an unreviewed safety question.

Review of safety evaluations for proposed tests and experiments that affect nuclear safety.

Review of all proposed changes to Technical Specifications or this Operating License.

Review of reports covering evaluation and recommendations to prevent recurrence of all NRC violations.

Review of all reportable events and violations of Technical Specifications which impact nuclear safety.

Review of unit operations to detect potential nuclear safety hazards.

g.

Performance of special

reviews, investigations or analyses and reports thereon as requested by the Plant Manager.

l Review of each unplanned onsite effluent release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence to the Site Director and to the Nuclear Safety Review Board.

Review proposed changes to the Radiological Effluent Manual (REM)

AUTHORITY 6.5.1.7 The PORC shall:

a.

b.

C.

Recommend in writing to the Plant Manager approval or disapproval of items considered under 6.5.1.6 (c) above.

Provide for a determination in writing with regard to whether or not each item considered under 6.5.1.6(a),

and (b) above constitutes an unreviewed safety question.

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 disagreement between the PORC and the Plant Manager;

however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

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RECORDS 6.5.1.8 The PORC shall maintain written minutes of each PORC meeting that, at a minimum, document the results 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.

6.5.1.9 TECHNICAL REVIEH AND CONTROL Activities which affect nuclear safety shall be conducted as follows:

Procedures required by Specification 6.8.1 and other procedures which affect plant nuclear safety, or personnel health and safety and changes

thereto, shall be prepared, reviewed and approved.

Each such procedure or procedure change shall be reviewed by a qualified individual other than the individual who prepared the procedure or procedure

change, but who may be from the same organization as the individual who prepared the procedure or-procedure change.

Procedures shall be approved by the appropriate responsible manager as designated in writing by the Plant Manager.

b.

'roposed changes or modifications to plant nuclear safety-related structures,

systems, and components shall be reviewed by a qualified 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 modifications.

Proposed modifications to plant nuclear safety-related structures,

systems, and components and the implementing workplans shall be approved prior to implementation by the responsible manager as designated by the Plant Manager.

C.

Individuals responsible for reviews performed in accordance with Specifications 6.5.1.9a and -b, shall be designated by approved written procedures.

Each such review shall be performed by qualified personnel of the appropriate discipline and shall include a determination of whether or not additional, cross-disciplinary review is necessary.

Each such review shall also include determination of whether or not an unreviewed safety question is involved pursuant to Section 10 CFR 50.59.

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6.5.2 NUCLEAR SAFETY REVIEH BOARD (NSRB)

FUNCTION 6.5.2.1 The NSRB shall provide for'independent review and audit cognizance to assure adequacy of designated activities in the areas of:

a.

nuclear power plant operations b.

nuclear engineering c.

chemistry and radiochemistry d.

metallurgy I

e.

instrumentation and control f.

radiological safety g.

mechanical and electrical engineering h.

quality assurance practices

i. 'on-destructive testing COMPOSITION 6.5.2.2 The NSRB shall be composed of at least five members, including the Chairman.

Members of the NSRB may be from Nuclear Power, or other TVA organization or external to TVA.

VALIFICATIONS 6.5.2.3 The Chairman,

members, and alternate members of the NSRB shall be appointed in writing by the Senior Vice President, 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 five years technical experience in one or more areas 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 utilized as determined by the Chairman, NSRB to provide expert advice to the NSRB.

MEETING FRE UENCY 6.5.2.5 The NSRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per six months thereafter.

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QUORUM 6.5.2.6 The minimum quorum of the NSRB necessary for the performance of the NSRB review and audit cognizance functions of these technical specifications shall consist of no fewer than a

majority of the members.

This quorum shall include the Chairman or his appointed al.ternate and the NSRB members, including appointed alternate

members, meeting the requirements of Specification 6.5.2.3.

No more than a

minority of the quorum shall have line responsibility for operation of the unit.

REVIEH 6.5.2.7 a-b.

The NSRB shall be responsible for the review of documents listed below.

The NSRB may exercise this responsibility by oversight and sampling of reviews by others; the Chairman, NSRB may delegate reviews to subcommittees, task groups, or qualified reviewers:

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.

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

C.

d.

e.

Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50'.59, 10 CFR.

Proposed changes to Technical Specifications or this Operating License.

Violations of codes, regulations,

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

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

g.

All Reportable Events.

h.

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

systems, or components that could significantly affect nuclear safety.

i.

Activities of the PORC relevant to nuclear safety.

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AUDITS II, 6.5.2.8 Audits of unit activities shall be performed by the appropriate audit group under the cognizance of the NSRB.

Audits of these activities shall be conducted on a frequency commensurate with their safety significance and the results of their performance history.

The audits will be conducted at the frequencies listed below unless a documented assessment based on past performance (e.g.,

performance based verification techniques,

trends, and other performance indicators) is performed to justify extension of their frequency.

In no instance will the frequency be extended by more than 12 months beyond the listed frequency.

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

b.

The performance, training and qualifications of the entire facility operations staff at least one per 12 months.

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

months.

d.

e.

g.

h.

The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B," 10 CFR

.50, at least once per 24 months'he Site Radiological Emergency Plan and implementing procedures at least once per 12 months.

The Plant Physical Security Plan, the Safeguards Contingency

Plan, and implementing procedures at least once per 12 months.

The Facility Fire Protection Program and implementing procedures at least once per 24 months.

An independent fire protection and loss prevention program inspection and audit shall be performed once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm.

k.

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 years.

The radiological environmental monitoring program and the results thereof at least once per 24 months.

The program for handling and controlling radiological waste and implementing procedures at least once per 24 months.

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

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Any other area of unit operation considered appropriate by the NSRB or the Senior Vice President, Nuclear Power.

AUTHORITY 6.5.2.9 The NSRB shall report to and advise the Senior Vice President, Nuclear Power, on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.

RECORDS 6.5.2.10 Records of NSRB and audit activities shall be prepared.

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b.

Minutes of each NSRB meeting and reports of reviews encompassed by Section 6.5.2.7 above shall be prepared,

'pproved and forwarded to the Senior Vice President, Nuclear Power within 14 days following each meeting.

Audit reports encompassed by Section 6.5.2.8

above, shall be forwarded to the Senior Vice President, Nuclear Power Group and to the management positions responsible for the areas audited within 30 days after completion of the audit.

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 subsection 50.73 to 10 CFR, 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.

6.7 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 Plant Manager, Site Director 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 Report shall be prepared.

The report shall be reviewed by the PORC.

This report shall describe:

(1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components,

systems, or structures, and (3) corrective action taken to prevent recurrence; c.

The Safety Limit Violation Report shall be submitted to the Commission, the NSRB, the Plant Manager and the Site Director within 14 days of the violation; and

d. Critical operation of the unit shall not be resumed until authorized by the Commission.

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6.8 PROCEDURES 8(

PROGRAMS 6.8.1

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d.

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

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

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).

Refueling operations.

Surveillance and test activities of safety-related equipment.

e. Plant Physical Security Plan implementation.

Site Radiological Emergency Plan implementation.

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6.8.2 a.

b.

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Fire Protection Program implementation.

OFFSITE DOSE CALCULATION MANUAL implementation.

PROCESS CONTROL PROGRAM implementation.

The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33; PORC administrative process implementation.

Radiological Effluent Mannual implementation.

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

The intent of the original procedure is not altered.

For site controlled procedures, the change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

The change is approved in accordance with Specification 6.5.1.9 above within 14 days of implementation.

6. 9 REPORTING RE UIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the Regional Office of NRC, unless otherwise noted.

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6.9.1.1 STARTUP REPORT A summary report of plant startup and power ascension 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.

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 startup test program, (2) 90 days following "

resumption or commencement of commercial power operation, or (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'ommercial 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*

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A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and job functions, **e.g., reactor operations and surveillance, inservi ce inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The dose assignment to various duty functions may be estimates based 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 or due to oper'ator action to control reactor pressure shall be reported.

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

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

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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, Washington, 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 THIS SECTION INTENTIALLY LEFT BLANK 6.9.1.5 THIS SECTION INTENTIALLY LEFT BLANK 6.9.1.6 SOURCE TESTS Results of required leak tests performed on sealed sources if the tests reveal the presence of 0.005 microcurie or more of removable contamination.

6.9.2 SPECIAL REPORTS 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 4.

Meteorological Monitoring Instrumentation Inoperability 5.

Primary Containment Integrated Leak Rate Testing 3.2.J.3 3.2.I.2 4.7.A.2 Within 10 days after 30 days of inoperability.

Within 10 days after 7 days of inoperability.

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 Report shall be submitted within 10 days after the event describing the magnitude, frequency

spectrum, and resultant effect upon plant features important to safety.

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7.

Diesel Generator Reliability Improvement Program Report shall be submitted withi n 30 days of meeting failure criteria in Table 4.9.A.

As a minimum, the reliability Improvement Program report, for NRC audit shall include:

a.

A summary of all tests (valid and invalid) that occurred within the time period over which the last 20/100 valid tests were performed.

b. Analysis of failures and determination of root causes of failures.

c. Evaluation of each of the recommendations of NUREG/CR-0660, "Enhancement of Onsite Emergency Diesel Generator Reliability in Operating Reactors,"

with respect to their application to the plant.

d. Identification of all actions taken or to be taken to (1)

Correct the root causes of failures defined in b above and (2) Achieve a general improvement of diesel generator reliability.

e.

A supplemental report shall be prepared for an NRC audit within 30 days after each subsequent failure during a valid demand, for so long as the affected diesel generator unit continues to violate the criteria (3/20 or 6/100) for the reliability improvement program remedial action.

The supplemental report need only update the failure/demand history for the affected diesel generator unit since the last report for that diesel generator.

The supplemental report shall also present an analysis of the failure(s) with a root cause determination, if possible, and shall delineate any further procedural, hardware or operational changes to be incorporated into the site diesel generator improvement program and the schedule for implementation of those changes.

8.

Secondary Containment 4.7.C.

Within 90 days Leak Rate Testing" of completion of each test.

9.

High-Range 'Primary Containment Radiation Monitors 3.2.F Within 7 days after 7 days of inoperability.

10.

High-Range Gaseous Effluent 3.2.F Radiation Monitors Within 7 days after 7 days of inoperabi 1 i ty.

• 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.

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j 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.

e.

g.

Checks, inspections,

tests, and calibrations of components and

systems, including such diverse items as source leakage Reviews of changes made to the procedures or equipment or reviews of tests and experiments to comply with 10 CFR 50.59 Radioactive shipments Test results in: units of microcuries for leak tests performed pursuant to Specification 3.8.E Record of annual physical inventory verifying accountability of nonexempt sealed sources required to be leak tested per TS 3.8.E i.

Gaseous and liquid radioactive waste released to the environs j.

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.

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Transients that occur during plant operations wil-1 be reviewed and,a cumulative fatigue usage factor determined.

For transients which are more severe than the transi ents 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 item i through q shall be retained for the life of the plant.

An inventory of nonexempt sealed sources in possession shall be maintained.

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) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than or equal to 1000 mrem/hr shall be barricaded and conspicuously posted as a high radiation arha and entrance thereto shall be controlled by requiring issuance of a 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 radiation monitoring device which continuously indicates the radiation dose rate in the area.

1.

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

"Radiological control personnel or personnel escorted by Radiological control personnel in accordance with approved emergency procedures, shall be exempt from the RWP issuance requirement during the performance of their assigned

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

BFN Units 1, 2 and 3

6.0-17

C.

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 level in the area has been established and personnel have been made knowledgeable of them.

An individual qualified in radiation protection procedures who'is equipped wi th 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 Site Radiological Control Superintendent in the Radiological Hork Permit.

6.12.2 The requirements of 6.12.1,

above, shall also apply at 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 Shift Operations Supervisor on duty and/or the Radiological Control Shift Supervisor on duty.

In 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, 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 6.13.2

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The PCP shall be approved by the Commission prior to implementation.

Licensee-initiated changes to the PCP:

Shall be submitted. to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was madel'his submittal shall contain:

b.

1)

Sufficiently detailed information to support the rationale for the change without benefit of additional or supplemental information; 2)

A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and Shall become effective upon review and approval in accordance with Section 6.5.1.9.

BFN Units 1, 2 and 3

6.0-18

6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.'1 The ODCM shall be approved by the Commission prior to implementation.

6.14.2 Licensee-initiated changes to the ODCM:

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

This submittal sha)l contain:

1)

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

Information submitted should consist of a package of those pages of the ODCM to be changed with each page

numbered, dated and containing the revision

number, 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

b. Shall become effective upon review and approval in accordance with Section 6.5.1.9.

6.15 RADIOLOGICAL EFFLUENT MANUAL (REM) l.

The REM shall be approved by the Commission prior to implementation.

I 2.

Changes to the REM shall be reviewed by PORC prior to implementation.

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Changes to the REM shall be approved by the Commission prior to implementation.

BFN Units 1,

2 and 3

6.0-19

1 I

ENCLOSURE 2

DESCRIPTION AND JUSTIFICATION BRONNS FERRY NUCLEAR PLANT (BFN)

REASON FOR CHANGE BFN units 1, 2, and 3 Technical Specifications Section 6.0 "Administrative Controls" are being amended to provide more clarification for areas such as how plant procedure reviews will be conducted, better define some of the PORC responsibilities and provide an overall standardized format as detailed during a meeting with NRC management on January 9,

1989.

Enclosure 2

Page 2 of 17 DESCRIPTION AND JUSTIFICATION FOR CHANGE Existing TS 6.2.2.b reads:

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

NEH 6.2.2.b reads:

. "In addition, while the unit is not in a COLD SHUTDONN CONDITION at least one licensed Senior Operator shall be in the control room."

JUST I F ICATION:

This change provides clarification as to the requirements for a Senior Reactor Operator (SRO) to be in the control room luring plant conditions other than COLD SHUTDOHN.

The current TS only requires a

SRO to be

onsite, which is less restrictive than the proposed TS.

During refueling operation (core alterations),

TS 6.2.2.d is applicable and requires direct supervision by either a licensed SRO or a SRO limited to Fuel Handling.

For other conditions BFN will comply with shift manning requirements in Table 6.2.A of these TS.

Not requiring a

SRO to be in the control room while the reactor is in a COLD SHUTDONN CONDITION is consistent with current BFN practices.

ADD TS 6.2.3 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics, NRC issuances, industry advisors, Licensee Event Reports, and other sources which may indicate areas for improving paint safety.

COMPOSITION 6.2.3.2 The ISEG shall be composed of at least 3 dedicated full-time engineers located onsite.

These engineers wi 11 be supplemented by 2 full-time engineers shared among all TVA nuclear sites.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be responsible for maintaining surveillance of plant activities to provide independent verification'hat these activities are performed correctly and that human errors are reduced as much as practical.

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Enclosure 2

Page 3 of 17 AUTHORITY 6.2.3.4 The ISEG shall make detailed recommendations for revised procedures, equipment modifications, or other means of improving plant safety to the Manager of Licensing and Regulatory Affairs.

JUSTIFICATION BFN has an ISEG organization performing the subject functions.

By adding this section to the TS ensures that specific requirements imposed by

. NRC-NUREG 0737, current industry practices and GE standard TS are maintained.

The addition of this section provides an independent review of various plant activities to enhance overall plant safety.

ADD TS 6.2.4 6.2.4.1 JUSTIFICATION The STA shall serve in an advisory capacity to the shift supervisor on matters pertaining to the engineering aspects of assuring safe operation of the unit.

The STA 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.

The first sentence to this section is being added to clarify the capacity of the Shift Technical Advisor (STA) with regard's to assuring safe plant operation.

The second part of this TS exists in the current BFN section 6.3.

This change does not alter the importance of or role of the STA.

This change provides clarification and places the requirements for the STA in it own independent section.

EXISTING TS 6.3 reads:

Qualifications.of the Browns Ferry Nuclear plant management and operating staff shall meet the minimum acceptable levels as described in ANSI-NI8.1, 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.

The Shift Technical Advisor shall have a bachelor's degree or equivalent in

'cientific or engineering discipline with specific training in plant design and transient and accident response and analysis.

Enclosure 2

Page 4 of 17 CHANGE TS 6.3 TO READ:

6.3.1 Each member of the facility shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, except for the Site Radiolqgical Control Superintendent who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

An additional exception is that the SRO license requirement for the Operations Superintendent shall be met if either the Operations Superintendent or a manager reporting directly to the Operations Superintendent and responsible for the direction of Shift Operations holds a valid SRO license.

The holder of the SRO license shall direct the licensed activities of the licensed operators.

JUSTIFICATION:

This change provides clarification and identifies additional qualification requirements other than ANSI N18.1-1971.

In addition, this change requires that the Operations Superintendent or a manager reporting directly to the Operation Superintendent hold a valid SRO license.

This ensures that an individual holding-one of these jobs has an.acceptable working knowledge and training in order to better understand the overall operation and response of the plant.

EXISTING TS 6.4 reads:

~Trai n i n A retraining and replacement training program for station personnel shall be in accordance with ANSI-N 18.1, Selection and Training of Nuclear Power Plant Personnel, dated March 8, 1971.

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

NEW TS 6.4 WILL READ:

6. 4 TRAINING 6.4.1 A retraining and replacement training program for the facility operations staff shall be maintained under the direction of the Operations Superintendent and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.

JUSTI F ICATION

'he proposed TS better defines and clarifies the training requirements for the operations staff.

This change identifies the appropriate 10 CFR and NRC "March 1980" training requirements and familiarization with industry operational experience.

Enclosure 2

Page 5 of 17 1.

Plant Operation Review Committee (PORC) Composition and Alternates A

Existin TS

(

6.0-6) reads:

6.5.1.2 The PORC shall be composed of the:

a.

Chairman:

Alternate Chairman:

= Plant Manager Assistant to Plant Manager Alternate Chairman or Member:

Technical Services Superintendent Member:

Member; Member:

Member:

Unit Superintendents (3)

Maintenance Superintendent Quality Assurance Staff Supervisor Health Physics Supervisor b.

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

CHANGE TS TO READ'.5.1.2 The PORC shall be composed of the following individuals and designated managers (representatives):

Chairman:

Member:

Member:

Member:,

Member:

Member:

Member:

Member:

Plant Manager Operations Superintendent Radiological Control Superintendent Maintenance Superintendent Technical Support Services Superintendent Site Quality Representative Engineering Representative Plant Support Representative The Chairman shall designate the Representatives in writing.

The Chairman and members shall meet the qualification requirements of ANSI Standard N18'.1 1971.

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Enclosure 2

Page 6 of 17 ALTERNATES 6.5.1.3 Alternate chairmen and alternate members shall be appointed in writing by the Plant Manager;

however, no more than two alternates shall participate as voting members in PORC activities at any one time.

All alternate chairmen and alternate members shall meet the qualification.

requirements of ANSI Standard N18.1 1971.

'I JUSTIFICATION The proposed technical specification change establishes a maximum of 2 alternates as voting members.

establish qualification requirements for members and alternates adds Engineering to the composition In establishing a maximum of 2 alternates as voting members the Plant Manager exercises direct control over the membership of PORC including designation of a Chairman.

Engineering is added as a specific organization to round out the PORC composition in this important functional area.

PORC exists to advise the Plant Manager on all matters related to nuclear safety.

PORC performs its mission of independent review and audit of unit operations utilizing a standing committee of individuals from all functional areas.

This change will allow the flexibility necessary to perform that function while allowing the Plant Manager to designate the individuals to advise him on matters related to nuclear safety.

As much as possible the same group of individuals is utilized by the Plant Manager to provide this advice.

In designating managers in some cases from applicable organizations instead of a specific job title safety is not compromised in that the managers should have sufficient comprehensive knowledge or the ability to get further advise from their organization to give the Plant Manager their experience and advise him on matters related to nuclear safety.

2.

PORC Quorum EXISTING TS

(

a e 6.0-7)

READS:

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.

0 Enclosure 2

Page 7 of 17 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.

CHANGE TS TO READ:

The minimum quorum of the PORC necessary for the performance of the PORC responsibility and authority provisions of these technical specifications shall consist of the Chairman or his designated alternate and four members including alternates.

JUSTIFICATION:

This change deletes the ability to convene the PORC individually (especially by telephone) and specifies a smaller minimum quorum.

BFN has discontinued the use of telephone PORC and has been doing business by formal meeting only.

The establishment of a minimum quorum of the chairman and 4 members is in keeping with industry standards for a review committee of this sort and does not compromise their ability to advise the Plant Manager.

The TSs do not preclude a larger committee but sets a

minimum voting quorum.

The maximum size of the PORC will be administratively controlled.

3.

PORC Responsibilities EXISTING TS (

s 6.0-7 8 and 9)

READS:

The PORC shall be responsible for the activities listed below.

The PORC may delegate the performance of reviews, but will maintain cognizance over and responsibi 1 i ty for them, e.g.,

subcommi ttees.

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 Appendix A of Regulatory Guide 1.33, Revision 2, February 1978 and changes thereto.

c.

Review of emergency operating procedures and changes thereto.

d.

Review implementing procedures of the Radiological Emergency Plan and the Industrial Security Program.

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

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Enclosure 2

Page 8 of 17 g.

Review proposed changes to the Radiological Effluent Manual.

h.

Review adequacy of the Process Control Program and Offsi te Dose Calculation Manual at least once every 24 months.

i.

Review changes to the radwaste treatment systems.

j.

Review of every unplanned onsite release of radioactive

, material to the environs including the preparation and forwarding of reports covering evaluation, recommendation, and disposition of the corrective action to prevent recurrence to the Senior Vice President, Nuclear Power, and to the Nuclear Safety Review Board.

k.

Review of all safety evaluations for modifications to structures, systems or components that affect nuclear safety 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; 1.

Review of reportable

events, unusual

events, operating anomalies, and abnormal performance of plant equi pment.

m.

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

n.

Review of unit 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).

o.

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

CHANGE TS TO READ:

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

The PORC may delegate the performance of reviews to subcommittees or qualified reviewers but will maintain responsibility for these reviews.

a.

Review safety evaluations for:

(1) procedures and (2) change to procedures, equipment, systems or facilities to verify that such actions do not constitute an unreviewed safety question.

be Review of safety evaluations for proposed tests and experiments that affect nuclear safety.

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Enclosure 2

Page 9 of 17 c.

Review of all proposed changes to Technical Specifications or this operating license.

d. Review of reports covering evaluation and recommendations to prevent recurrence of all NRC violations.

e.

Review of all reportable events and violations of Technical Specifications which impact nuclear safety.

f. Review of unit operations to detect potential nuclear safety hazards.

g. Performance of special

reviews, investigations or analyses and reports thereon as requested by the Plant Manager

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h. Review of each unplanned onsite effluent release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence to the Site Director and to the Nuclear Safety Review Board.

i. Review proposed changes to the Radiological Effluent Manual (REM).

JUSTIFICATION:

This change reformats the PORC responsibilities and specifically deletes reviews of procedures that are better done under the; qualified review process (current tech spec items a, b, c, d).

PORC will review the safety evaluations for procedures and changes to procedures under new tech spec item (a).

In addition, PORC supervises reviews of procedures under the qualified reviewer program.

In this way they will not be bogged down in reviewing every change to these procedures but through a

review of the associated safety evaluation will be aware of any aspect that could potentially impact nuclear safety.

"0 Enclosure 2

Page 10 of 17 The following responsibilities are one for one from the old to the new tech specs:

Old TS item New TS item Existing technical specification (i), radwaste treatment system changes is provided for by PORC' review of safety eva'ovations for changes to equipment, system or facilities to verify such actions do not constitute an unreviewed safety question.

Existing technical specification (j) is modified by the addition of the word "effluent" to become new technical specification (h).

This change will allow PORC to address the details significant to nuclear safety.

Existing technical specifications items (1) and (m) are encompassed by new tech spec item (e)

Existing technical specification item (h) is deleted.

Normal reviews for adequacy are performed by the appropriate line manager and any changes will be handled under the provisions of technical specification sections 6.13 and 6.14.

By this change the PORC responsibilities are scaled down to address those specific items that are related to nuclear safety.

Specific procedure reviews are deleted from PORC review but are reviewed by a qualified reviewer.

PORC maintains supervision of the qualified reviewer process.

In this manner PORC can better focus on the essence of nuclear safety and not the particulars of a given procedural change..

The Plant Manager will exersise administrative control over what materials will be reviewed by PORC through Implementing procedures.

None of this precludes a special review of a procedure or major plant evolution by PORC as requested by the Plant Manager or the Nuclear Safety Review Board.

By deleting these procedure reviews PORC can better perform its function as an oversight committee for the Plant Manager on matters related to nuclear safety.

Unusual

events, operating abnormalities and abnormal performance of plant equipment are brought to the attention of the Plant Manager through the responsible managers that report to him.

Enclosure 2

Page 11 of 17 PORC Authority EXISTING TS

(

a e 6.0-10)

READS:

The PORC shall:

a.

Recommend to the Plant Manager in writing, approval, or disapproval of items considered under 6.F 1.6.a through i 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'.1.6.f above constitutes an unreviewed safety question.

c:

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

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.

CHANGE TS TO READ'he PORC shall:

a, Recommend in writing to the Plant Manager approval or disapproval of items considered under 6.5.1.6 (c) above.

b. Provide for a determination in writing with regard to whether or not each item considered under 6.5.1.6(a),

and (b) above constitutes an unreviewed safety question.

c. 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 disagreement between the PORC and the Plant Manager;

however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

JUSTIFICATION

'he authority section of the TS is reformatted to be consistent with the responsibilities above and delete the ability for PORC to act individually.

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Page 12 of 17 ADD TS 6.5.1.9 TO READ:

TECHNICAL REVIEH AND CONTROL Activities which affect nuclear safety shall be conducted as follows:

a.

Procedures required by Specification 6.8.1 and other procedures which affect plant nuclear safety, or personnel health and safety and changes

thereto, shall be prepared, reviewed and approved.

Each such procedure or procedure change shall be reviewed by a qualified individual other than the individual who prepared the procedure or procedure

change, but who may be from the same organization as the individual who prepared the procedure or procedure change.

Procedures shall be approved by the appropriate responsible manager as designated in writing by the Plant Manager.

b. Proposed changes or modifications to plant nuclear safety-related structures,

systems, and components shall be reviewed by a

'qualified 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 modifications.

Proposed modifications to plant nuclear safety-related structures,

systems, and components and the implementing workplans shall be approved prior to implementation by the responsible manager as designated by the Plant Manager.

c. Individuals responsible for reviews performed in accordance with Specifications 6.5.1.9a and-b, shall be designated by approved written procedures.

Each such review shall be performed by qualified personnel of the appropriate discipline and shall include a determination of whether or not additional, cross-disciplinary review is necessary.

Each such review shall also include determination of whether or not an unreviewed safety question is

- involved pursuant to Section 10 CFR 50.59

'USTIFICATION

'hese changes enhance the existing TS by requiring a qualified individual to perform the procedure review.

This will ensure procedural quality by allowing the appropriate responsible manager to approve the appropriate procedures.

This will allow the Plant Manager to attend to other major responsibilities and plant operations while ensuring the procedures are adequately reviewed/approved by those managers responsible for their implementation.

This also relieves the burden for specific reviews from PORC.

PORC is still responsible for the review/approval of all 10 CFR 50.59.

Changes made to the subject procedures will go through the 10 CFR 50.59 process before they can be signed by the responsible manager.

This will ensure that PORC will maintain sufficient knowledge of the subject procedures as they relate to nuclear safety.

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Enclosure 2

Page 13 of 17 ADD TS 6.5.2.1..i....

Non-destructive testing JUSTIFICATION'dding this to the list of activities under NSRB cognizance increases the scope of the NSRB independent review and makes it conform to ANSI N18.7.

In addition TS 6.5.2.1 is revised to more clearly state the NSRB role in review and audit.

This change does not alter the function of NSRB as a

safety oversight board.

EXISTING TS 6.5.2. 4 Consultants shall be utilized to provide expert advise as determined by the NSRB.

CHANGE TS 6.5.2.4 TO READ:

Consultants shall be utilized as determined by the Chariman, NSRB to provide expert advice to the NSRB.

JUSTI F ICATION

'urrenttechnical specifications allow the use of members external to TVA but provide no details on their use.

This change will explicitly allow the Chairman of the NSRB to use industry consultants to provide their expert advise to the NSRB.

This change is consistent with industry and current TVA practice.

EXISTING TS 6.5.2. 5 The NSRB shall meet at least once per six months CHANGE TS 6.5 '.5 TO READ:

The NSRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per six months thereafter.

JUST I F ICATION:

-This change increases the meeting frequency subsequent to fuel load and operation to provide the necessary oversight of plant activities after resumption of power operation following the extended shutdown.

The minimum meeting frequency in the current technical specifications is unchanged.

EXISTING TS 6.5.2. 6 NSRB UORUM The minimum quorum of the NSRB necessary for the performance of NSRB review and audit functions of these technical specifications shall consist of more than half of the NSRB membership or at least five members, whichever is greater.

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Enclosure 2

Page 14 of 17 CHANGE TS 6.5.2.6 TO READ:

The minimum quorum of the NSRB necessary for the performance of the NSRB review and audit cognizance functions of these technical specifications shall consist of no fewer than a majority of the members.

JUSTIFICATION:

This is an administrative change that will allow some accommodation at NSRB meetings while not changing the requirement to provide adequate representation in all the areas of expertise required.

Procedures wi 1 1 be used to assure appropriate competence to address all issues before the NSRB.

EXISTING TS 6.5.2.7 The NSRB shall review:

CHANGE TS 6.5 '.7 TO READ:

6.5.2.7 The NSRB shall be responsible for the review of documents listed below.

The NSRB may exercise this responsibility by oversight and sampling of reviews by others; the Chairman, NSRB may delegate reviews to subcommittees, task groups, or qualified reviewers.

JUSTIFICATION:

This change allows better devotion of NSRB time resources to matters that

'impact nuclear safety by allowing the delegation of reviews to subcommittees, task groups or qualified reviewers.

It is consistent with their use in other places in these technical specifications.

This allows the NSRB more flexibilityto oversee plant operations while maintaining cognizance over the reviews that are important the safety considerations at the plant.

Procedural controls wi 11 ensure the appropriate involvement of the NSRB in changes to the technical specifications.

EXISTING TS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the NSRB.

These audits shall encompass:

CHANGE TS 6.5.2.8 TO READ:

Audits of unit activities shall be performed by the appropriate audit group under the cognizance of the NSRB.

Audits of these activities shall be conducted on frequency commensurate with their safety significance and the results of their performance history.

The audits will be conducted at the frequencies listed below unless a documented assessment based on past performance (e.g.,

performance based verification techniques,

trends, and other performance indicators) is performed to justify extension of their frequency.

In no instance will the frequency be extended by more than 12 months beyond the listed frequency.

Enclosure 2

Page 15 of 17 JUST I FICATION'-

This change will specifically allow the use of a performance based audit program that will identify problem areas requiring more frequent audits and lessen the cycle for areas wi th a good performance record.

Thi s i s in keeping with industry trends and allows a review board of this sort to more effectively identify problems and then verify corrective actions.

to justify the extension of audit'frequency of a particular area documented justification will be prepared and available for review.

REVISE TS 6.7 SAFETY LIMIT VIOLATION The current TS requires the Senior Vice President, Nuclear Power, to be involved in the reporting chain (6.7.l.a and c) if a safety limit is violated.

This proposed change will add the Plant Manager and Site Director and delete the Senior Vice President, Nuclear Power.

JUSTI F ICATION:

The Site Director is located on site and is involved with the daily activities that occur at the facility.

In addition, he does not have additional responsibilities at other nuclear plants which would detract from those at the site at which he is located.

The Site Director is familiar not only with the operations and problems associated at his site but is knowledgeable of the procedures, and controls the manpower to provide a fast response to resolve them.

This change is an enhancement

, dedicating a top level manager to the site which will also ensure that the time requirements specified in section 6.7 are complied with.

Adding the plant manager to this section is consistent with Section 6.1.1 in that he has the overall responsibility for the safe operation of the plant.

ADD TS 6.8.l.c TO READ AS FOLLONS:

"Refueling Operations" JUST I F ICATION

'his addition requires written procedures for refueling operations to be established, implemented',

and maintained.

This will not only provide clarification concerning refueling operations but it will provide consistency and assist in minimizing errors.

This addition supports good "operating" practices.

ADD TS 6.8.1.J TO READ AS FOLLOHS:

The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter 82-33 JUST I F ICATION

'his addition provides added assurance that NRC requirements identified in the subject requirements are established, implemented, and maintained. in the appropriate emergency operating procedures.

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Enclosure 2

Page 16 of 17 DELETE EXISTING TS 6

~ 8.2 DRILLS Drills on actions to be taken under emergency conditions involving release of radioactivity are specified in the Radiological Emergency Plan and sha'll be conducted annually.

Annual drills shall also be conducted on the actions to be taken following failures of safety-related systems or components.

JUSTIFICATION:

An annual exercise of the site emergency plan is required by 10 CFR 50, Appendix E.

Appendix E also gives the details and periodicity for the participation requirements for these exercises.

Deleting this from the Administrative Controls section of the technical specifications will eliminate ambiguity from the technical specifications.

Drills on actions to be taken following failures of safety related systems or components are performed in the simulator during the continuing training which supports qualification and requalification of operators and supervisors.

The elimination of these requirements from the Administrative Controls section will make the BFN Technical Specifications similar to the industry standards but will not change any requirements.

REVISE SECTION 6.8.2.b TO READ:

"For site controlled procedures,"

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

JUSTI F ICATION:

The addition of "For site controlled procedures" provides clarification as to which procedures this TS is referring to.

This eliminates any confusion with other existing procedures used by other departments existing for requirements supporting the plant which are not, nor should they be, under the control of this TS.

DELETE TS 6.9.1.4 This Section is Intentionally left Blank DELETE TS 6.9.1.5 This Section is Intentionally left Blank Administrative CHANGES MADE THROUGHOUT SECTION 6.0 1.

The existing TS refer to Health Physics individuals/technicians.

The title of Health Physics has been changed 'to Radiological Control.

This change provides consistency with the titles used within TVA and the TS

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2.

The existing section 6.8.3 radiation control procedures was rewritten and is now numbered 6.12.

Even though rewritten, the overall content and intent of this section was not changed.

This change provided clarification to the existing wording.

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Page 17 of 17 3.

TS sections 6.14 and 6.1S (current TS sections 6.11 and 6.12) also added some words of clarification that do not change the intent of the TS.

These changes identify the specific TS section as to how updates to the Process Control Program (PCP) and Offsite Dose Calculation Manual (ODCM) shall be submitted to NRC.

By adding the words "without benefit of additional or supplemental information" requires that the sub]ect PCP and ODCM submittals are self contained.

This complies with current TVA practice.

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ENCLOSURE 3

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION BRONNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 DESCRIPTION OF PROPOSED TS AMENDMENT The proposed amendment would change the BFN Technical Specifications for units 1, 2, and 3.

SECTION 6.0 ADMINISTRATIVE CONTROLS These changes more accurately reflect the way in which BFN will control and conduct operations.

These changes provide additional clarification for those responsibilities and programs listed in the subject section.

Basis for Proposed No Significant Hazards Consideration Determination.

NRC has provided standards for determining whether a significant hazards consideration exists as stated in 10CFRS0.92(c).

A proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability of consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind if accident from an accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

1.

The proposed technical specification changes to BFN Section 6 do not involve a significant increase in the probability or consequence of any accident previously evaluated.

The proposed changes enhance the overall operation and support of the plant by more clearly defining responsibilities and qualifications of various plant and support personnel in addition to providing appropriate guidance on specific procedures,

programs, and processes.

The proposed changes are administrative in nature and do not involve any modifications to safety-related equipment currently installed in the plant.

The proposed changes will better align BFN with current industry and NRC practices which have been proven to support the overall safety of the plant.

The subject changes do not prevent or alter the operation of any equipment required to mitigate any accident in which BFN is licensed.

These changes do not involve any physical modification to the design or operation of the plant, therefore, they do not invalidate the assumptions made in the Final Safety Analysis Report (FSAR) or safety limits currently identified in the TSs.

These changes provide the required guidance and clarification to the plant staff on how specific practices are to be conducted.

This includes plant practices ranging from the review and approval of specific procedures and programs to the responsibilities of the Plant Operations Review Committee and Nuclear Safety Review Board.

For these

reasons, the proposed changes do not affect the probability or consequences of any accident previously analyzed.

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C Enclosure 3

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The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes are purely administrative in nature and do not modify or alter the configuration of the plant.

These changes do not create any new accident mode or release pathway of radioactive effluents to the environment.

These changes add assurance that safe plant practices, as specified in BFN Section 6, are provided for by requiring appropriate reviews'and approvals prior to their implementation.

3.

The proposed amendment does not involve a significant reduction in a margin of safety.

The proposed changes do not remove or diminish any elements of the nuclear organization or practices that are essential to the safe operation of BFN.

They do not involve any changes to plant operating systems or associated safety analyses.

The changes enhance the overall clarity of the key functions and responsibilities of those practices and personnel identified in'FN Section 6.

DETERMINATION OF BASIS FOR PROPOSED HO SIGNIFICANT HAZARDS Since the application for amendment involves a proposed change that is not encompassed by the criteria for which no significant hazards consideration

exists, TVA has made a proposed determination that the application involves no significant hazards consideration.

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