ML14175A738

Amend 70 to License DPR-23,amending Tech Specs to Modify Corporate & Plant Organizational Structures & Mods to Plant Nuclear Safety Committee

ML14175A738
Person / Time
Site:	Robinson
Issue date:	06/28/1982
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML14175A739	List: ML14175A738 ML14175A741 ML14175A742
References
DPR-23-A-070 NUDOCS 8207120210
Download: ML14175A738 (36)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION o

WASHINGTON, D. C. 20555 CAROLINA POWER AND LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UAIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. DPR-23

1. The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Carolina Power and Light Company (the licensee) dated July 10, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

8207120210 620628 PDR ADOCK 05000261 PDR

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-23 is hereby amended to read as follows:

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

3. This license amendment is effective 45 days from the date of its issuance.

FR THE NUCLEAR REGULATORY COMMISSION teven A. Var a, h f Operating Reacto Branch No. 1 Division of Licen ing

Attachment:

Changes to the Technical Specifications Date of Issuance: June 28, 1982

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 70 TO FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Revise Appendix A as follows:

Remove Pages Insert Pages 6.2-1 thru 6.2-4 6.2-1 thru 6.2-3 6.3-1 6.3-1 6.4-1 6.4-1 6.5-1 thru 6.5-12 6.5-1 thru 6.5-16 6.6-1 6.6-1 6.7-1 6.7-1 6.8-1 6.8-1 6.9-1 thru 6.9-8 6.9-1 thru 6.9-9 6.10-2 6.10-2

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

Facility Staff 6.2.2 The facility organization shall be as shown on Figure 6.2-2 and:

a.

The shift complement shall consist of at least one Shift Foreman holding a Senior Reactor Operator's-License, three control operators each holding at least a Reactor Operator's license, one additional shift member, and one Shift Engineer (shift technical advisor).

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

c. At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown, and during recovery from reactor trips.

d. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

e. ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f. A Plant Fire Brigade of at least 5 members shall be maintained on site at all times.

This excludes three members of the minimum shift crew necessary for safe shutdown of the plant and any personnel required for other essential functions during a fire emergency.

6.2-1 AMENDMENT NO.

70

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6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed ANSI N18.1-1971 with regard to the minimum qualifications for comparable positions.

6.3.2 The Manager - Environmental and Radiation Control shall meet or exceed the qualifications of Regulatory Guide 1.8, September, 1975.

The Guide says that he shall have a bachelor's degree or equivalent in a science or engineering subject. Equivalent in this case is defined as follows:

(a) 4 years of formal schooling in science or 'engineering, (b) 4 years of applied radiation protection experience at a nuclear

facility, (c) 4 years of operational or technical experience/training in nuclear power, or (d) Any combination of the above totaling 4 years.

This requirement is in addition to the requirement for five years of professional experience in applied radiation protection.

6.3.3 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 response and analysis of the plant for transients and accidents.

6.3-1 AMENDMENT NO.

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6.5 REVIEW AND AUDIT 6.5.1 The license organization's review and approval process shall assure that the nuclear safety of the facility is maintained.

6.5.1.1 Procedures, Tests, and fxperiments 6.5.1.1.1 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, Rev. 2, February 1978.

b. Refueling operations.

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

d. Security Plan implementing procedures.

e. Emergency Plan implementing procedures.

f. Fire Protection Program implementing.

6.5.1.1.2 A safety analysis shall be prepared for all procedures, tests, and experiments covering the activities identified in 6.5.1.1.1 and procedures that affect nuclear safety.

The analysis shall include a written determination of whether or not the procedure, test, or experiment is a change in the facility as described in the FSAR, involves a change to the Technical Specification, or constitutes an unreviewed safety question as defined in 10CFR50.59(a)(2).

This analysis constitutes a first party safety review and may be accomplished by the individual who prepared the document.

6.5-1 AMENDMENT NO. 70

6.5.1.1.3 Prior to approval, a second safety review shall be performed on all procedures, tests, or experiments that affect nuclear safety.

This review shall be performed by an individual other than the individual who was the original preparer.

6.5.1.1.4 Following the two-party review, procedures, tests, and experiments and permanent changes th'ereto (other than editorial, or.

typographical) which have been determined neither to involve an unreviewed safety question as defined in 10CFR50.59(a)(2), nor a change to the Technical Specifications, shall be approved prior to implementation by one-of the following:

a..Plant General Manager, or designated alternate, or

b. The Manager of the functional area affected by the procedures tests, and experiments and permanent changes thereto as previously designated by the Plant General Manager.

The individual approving the procedure, test, or experiment or change thereto shall be other than those who performed the required reviews.

The Plant General Manager or other designated manager approving the review activities of the two-party review shall assure that the reviewers collectively possess the background and qualifications in all of the disciplines necessary and important to the specific review.

To assure that the individuals selected for the two-party review are qualified and have the background necessary, the Plant General Manager shall approve and maintain a list of qualified persons.

Included in this list will be individuals in addition to the first and second party reviewer whose expertise may be necessary during the review to assure that the reviewers collectively possess the background and qualifications in the disciplines necessary and important to the specific review.

6.5-2 AMENDMENT NO. 70

The list will include the disciplines for which each person is qualified.

6.5.1.1.5 Temporary changes to procedures, tests, or experiments may be approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator License if such change does not change the intent of the original procedure, -test, or experiment.

Temporary changes shall be documented and, within 21 days of receiving temporary approval, be reviewed in accordance with specification 6.5.1.1.2, 6.5.1.1.3 and 6.5.1.1.4 and incorporated as a permanent change or deleted.

6.5.1.1.6 Those procedures, tests, or experiments and changes thereto that constitute an unreviewed safety question, or involve a change to Technical Specifications shall be reviewed by the Plant Nuclear Safety Committee and submitted to the NRC for approval prior to implementation. All such procedures, tests, or experiments and changes shall be reviewed by the Corporate Nuclear Safety Section prior to implementation.

6.5.1.1.7 Procedures, tests, or experiments, which constitute a change to the FSAR shall also be reviewed by the Corporate Nuclear Safety Section. These reviews may be conducted after plant Management approval, and implementation may proceed prior to completion.of review as provided for by 10CFR50.59(a)(1).

6.5.1.2 Modifications 6.5.1.2.1 A safety analysis shall be prepared for all modifications that affect nuclear safety. The analysis shall include a written determination of whether or not the modification is a change in the facility as described in the FSAR, involves a change to the Technical Specification, or constitutes an unreviewed safety question as defined in IOCFR50.59(a)(2).

This analysis constitutes a first party safety review and may be accomplished by the individual who prepared the modification.

6.5-3 AMENDMENT NO.

70

6.5.1.2.2 Prior to approval, a second safety review shall be performed on all modifications that affect nuclear safety. This review shall be performed by a qualified individual other than the individual who was the original preparer.

6.5.1.2.3 Following the two party review, modifications that have been determined neither to involve an unreviewed safety question as defined in IOCFR50.59(a)(2) nor a change to the Technical Specifications shall be approved, prior to implementation, by one of the following:

a. Plant General Manager, or designated alternate, or

b. The Manager of Technical Support.

The individual approving these modifications shall be other than those who performed the required reviews.

The Plant General Manager or other designated manager approving the review activities of the two-party review shall assure that the reviewers collectively possess the background and qualifications in all of the disciplines necessary and important to the specific review.. To assure that the individuals selected for the two-party review are qualified and have the background necessary, the Plant General Manager shall approve and maintain a list of qualified persons. Included in this list will be individuals in addition to the first and second party reviewers whose expertise may be necessary during the review to assure that the reviewers collectively possess the background and qualifications in the disciplines necessary and important to the specific review. The list will include the disciplines for which each person is qualified.

6.5-4 AMENDMENT NO.

70

6.5.1.2.4 Modifications that are determined to either constitute an unreviewed safety question, as defined in 10CFR50.59(a)(2),

or a change to the Technical Specifications, shall be reviewed by-the Plant Nuclear Safety Committee and submitted to the NRC for approval prior to implementation.

All such modifications shall be approved by the Corporate-Nuclear Safety Section prtor-to implementation.

6.5.1.2.5 Modifications which constitute changes to the facility as described in the FSAR shall.also be reviewed by the Corporate Nuclear Safety Section. This review may be conducted after plant Management approval, and implementation may proceed prior to completion of review.

6.5.1.3 Technical Specifications and License Changes 6.5.1.3.1 Each proposed Technical Specification or Operating License change shall be reviewed by the Plant Nuclear Safety Committee and submitted to the NRC for approval.

6.5.1.4 Review of Technical Specification Violations 6.5.1.4.1 All violations of Technical Specifications shall be investigated and a report prepared that -evaluates the occurrence and that

.provides recommendations to prevent recurrence. Such reports shall be reviewed by the PNSC and approved by the Plant General Manager or his designee and submitted to the Vice President -

Nuclear Operations and to the Manager - Corporate Nuclear Safety.

6.5.1.5 Nuclear Safety Review Qualification 6.5.1.5.1 Individuals shall be designated by the Plant General Manager for the safety reviews of Specifications 6.5.1.1.2, 6.5.1.1.3, 6.5.1.2.1, and 6.5.1.2.2. These reviewers shall have a Bachelor of 6.5-5 AMENDMENT NO.

70

IS Science in engineering or related field or equivalent and two (2) years related experience.

6.5.1.6 Plant Nuclear Safety Committee (PNSC) 6.5.1.6.1

a. As. an effective means for the regular overview, qvaluation, and maintenance of plant operational safety, a Plant Nuclear Safety Coimittee (PNSC) is established.

b. The committee shall function through the utilization of subcommittees, audits, investigations, reports, and/or performance of reviews as a group.

6.5.1.6.2 The PNSC shall be composed of the following:

Chairman -

General Manager or designated alternate Member -

Manager -

Operations and Maintenance or designated alternate Member -

Manager -

Technical Support or designated alternate Member -

Assistant to General Manager Member -

Manager -

Environmental & Radiation Control or designated alternate Member -

Director -

QA/QC or designated alternate 6.5.1.6.3 Alternates shall be appointed in writing by the General Manager to serve on a temporary basis.

Alternates shall, as a minimum, meet the qualifications specified for professional-technical personnel in Section 4.4 of ANSI N18.1-1971.

6Property "ANSI code" (as page type) with input value "ANSI N18.1-1971.</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..5.1.6.4 The PNSC shall meet at least once per calendar month and as convened by the PNSC Chairman or his designated alternate.

6.5-6 AMENDMENT NO.

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6.5.1.6.5 A quorum of the PNSC shall consist of the Chairman, and three members, of which two may be alternates.

6.5.1.6.6 The PNSC activities shall include the following:

a. Perform an overview oi Specifications 6.1.1.1, and 46.5.1.2 to assure that processes are effectively maintained.

b. Performance of special reviews, investigations, and reports thereon requested by the-Manager -

Corporate Nuclear Safety.

c. Annual review of the Security Plan and.Emergency Plan.

d. Perform reviews of Specifications 6.5.1.1.6, 6.5.1.2.4, 6.5.1.3.1, and 6.5.1.4.1.

e. Perform review of all events requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the NRC.

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

6.5.1.6.7 In the event of disagreement between the recommendations of the Plant Nuclear Safety Committee and the actions contemplated by the General Manager, the course determined by the General Manger to be more conservative will be followed. The Vice President -

Nuclear Operations and the Manager -

Corporate Nuclear Safety will be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the disagreement and subsequent actions.

6.5.1.6.8 The PNSC shall maintain written minutes of each meeting that, at a minimum, document the results of all PNSC activities performed under the provisions of-these Technical Specifications; and copies shall be provided to the Vice President -

Nuclear Operations, and to the Manager -

Corporate Nuclear Safety.

6.5-7 AMENDMENT NO.

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6.5.2 Corporate Nuclear Safety Section -

Independent Review The Corporate Nuclear Safety Section of the Corporate Nuclear Safety & Research Department shall provide independent review of significant plant changes, tests, and procedures; verify that reportable occurrences are investigated in a timely manner and corrected in a manner that reduces the probability df recurrence of such events; and detect trends that may not be apparent to a day-to-day olfserver.

Specific review subjects are defined in Specification 6.5.2.1.d.

6.5.2.1 The Manager -

Corporate Nuclear Safety, under-the Vice President -

Corporate Nuclear Safety & Research, is charged with the overall responsibility for administering the independent review function as follows:

a. Approves selection of the individuals to conduct safety reviews under Specification 6.5.2.

b. Has access to plant records and operating personnel in performing independent reviews.

c. Prepares and retains written records of reviews.

d. Assures independent reviews are conducted on the following subjects:

(1)

Written safety evaluations of changes in the facility as described in the Safety Analysis Report, changes in procedures as described in the Safety Analysis Report, and tests or experiments not described in the Safety Analysis Report that are completed without prior NRC approval under the provisions of 10CFR50.59(a)(1). This review is to verify that such changes, tests, or 6.5-8 AMENDMENT NO. 70

experiments did not involve a change in the Technical Specifications or an unreviewed safety question as defined in 10CFR50.59(a)(2).

These reviews may be conducted after appropriate management approval, and implementation may proceed prior to completion of the review.

(2)

Proposed changes in procedures, proposed changes in the facility, or proposed tests or experiments, any of which involves a-change-in the Technical Specifications or an unreviewed safety question pursuant to 10CFR50.59(c).

Matters of this kind shall be referred to the Corporate..

Nuclear Safety Section by the Plant General Manager or by other functional organizational units within Carolina Power & Light Company prior to implementation.

(3)

Proposed changes to the Technical Specifications or this operating license, prior to implementation.

(4)

Violations, deviations, and reportable events -that require reporting pursuant to Specification 6.9.2.a.

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

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

Review of events covered under this paragraph shall include the results of any investigations made and the 6.5-9

-AMENDMENT NO. 70

recommendations resulting from such investigations to prevent or reduce the probability of recurrence of the event.

(5)

Any other matter involving safe operation of the nuclear power plant that the Manager -

Corporate Nuc-lear Safety Section, deems appropriate for consideration of which is referred to the Manager -

Corporate Nuclear Safety Section, by the on-site operating organization or by other functional organizations units within Carolina Power & Light Company.

(6)

Reports and minutes of the PNSC.

6.5.2.2 Results of Corporate Nuclear Safety reviews, including recommendations and concerns, shall be documented.

a. Copies of documented reviews shall be retained in the CNS files.

b. Recommendations and concerns shall be submitted to the plant General Manager and Vice President -

Nuclear Operations, within 14 days of determination.

c. A summation of Corporate Nuclear Safety recommendations and concerns shall be submitted to the Chairman/President; Vice Chairman; Executive Vice President -

Power Supply and Engineering & Construction; Senior Vice President -

Power Supply; Vice President -

Nuclear Operations; Vice President -

Nuclear Safety & Research; plant General Manager; and others, as appropriate on at least a bimonthly frequency.

d. The Corporate Nuclear Safety Review program shall be conducted in accordance with written, approved procedures.

6.5-10 AMENDMENT NO. 70

6.5.2.3 Personnel

a. Personnel assigned responsibility for independent reviews shall be specified in technical disciplines and shall collectively have the experience and competence required to review problems in the following areas:

(1)

Nuclear power plant operations (2)

Nuclear.e.ngineering (3)

Chemistry and radiochemistry (4)

Metallurgy (5)

Instrumentation and control (6)

Radiological safety (7)

Mechanical and electrical engineering (8)

Adminsitrative controls (9)

Seismic and environmental (10)

Quality assurance practices

b. The following minimum experience requirements shall be established for those persons involved in the independent safety review program:

(1)

Manager of CNSS -

Bachelor of Science in engineering or related field and ten (10) years' related experience, including five (5) years' involvement with operation and/or design of nuclear power plants.

(2)

Reviewers -

Bachelor of Science in engineering or related field or equivalent and five (5) years' related experience.

c. An individual may possess competence in more than one specialty area. If sufficient expertise is not available 6.5-11 AMENDMENT NO. 70

within the Corporate Nuclear Safety Section, competent individuals from other Carolina Power & Light Company organizations or outside consultants shall be utilized in performing independent reviews and investigations.

d. At least three persons, qualified as discussed in Specificftion 6.5.2.3.b, shall review each item submitted under the requirements of Section 6.5.2.1.d.

e. Independent safety reviews shall be performed by personnel not directly involved with the activity or responsible for the activity.

6.5-12 AMENDMENT NO. 70

6.5.3 Performance Evaluation Unit -

Audit 6.5.3.1 The Performance Evaluation Unit of the Corporate Quality Assurance Department shall perform audits of plant activities.

Specific audit subjects are defined in Specification 6.5.3.2.d.

6.5.3.2 The Principal QA Specialist -

Performance Evaluation Unit under the Manager Corporate Quality Assurance is charged with the overall responsibility for administering the quality assurance audit programs as follows:

a. Approves selection of the individual(s) to conduct quality assurance audits.

b. Has access to the plant operating records and operating personnel in performing the quality assurance audits.

c. Prepares and retains written records of audits.

d. Assures quality assurance audits are conducted on the following subjects:

(1)

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

(2)

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

(3)

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

6.5-13 AMENDMENT NO. 70

(4)

The verification of compliance and implementation of the requirements of the Quality Assurance Program to meet the criteria of Appendix B, 10CFR50, at least once per 24 months.

(5)

The Emergency Plan and implementing procedures at least once per 24 months.

(6)

The Security Plan and implementing procedures at least once per Z4-months.

(7)

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

(8)

Any other area of facility operation considered appropriate by the Corporate Quality Assurance Performance Evaluation Unit; the Executive Vice President -

Power Supply and Engineering & Construction; or the Senior Vice President -

Power Supply.

e. Distribute reports and other records to appropriate managers.

.6.5.3.3

a. Audit personnel shall be independent of the area audited.

Selection for auditing assignments is based on experience or training that establishes that their qualifications are commensurate with the complexity or special nature of the activities to be audited.

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

6.5-14 AMENDMENT NO.

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b. Qualified outside consultants or other individuals independent from those personnel directly involved in plant operation shall be used to augment the audit teams when necessary. Individuals performing the audits may be members of the audited organization; however, they shall not audit activities for which'they have immediate..responsibility, and while performing the audit, they shall not report to a managemeat representative who has immediate responsibility for the activity audited.

6.5.3.4 Results of plant audits are approved by the Principal QA Specialist -

Performance Evaluation Unit, and transmitted to the.

Executive Vice President -

Power Supply and Engineering &

Construction; the Senior Vice President -

Power Supply; Vice President -

Nuclear Operations; General Manager; and the Vice President -

Corporate Nuclear Safety & Research; and others, as appropriate within 30 days after the completion of the audit.

6.5.3.5 The Corporate Quality Assurance Audit Program shall be conducted in accordance with written, approved procedures.

6.5-15 AMENDMENT NO. 70

6.5.4 Outside Agency Inspection and Audit Program 6.5.4.1 An independent fire protection and loss prevention inspection and audit shall be performed at least once per 12 months utilizing either qualified offsite personnel or an outside fire protection firm.

6.5.4.2 An inspectiop 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.

6.5-16 AMENDMENT NO.

70

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

a. The NRC shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.2.

b. Each REPORTABLE OCCURRENCE requiring 24-hour notification to the NRC shall be reviewed in accordance with 6.5.1.6.6 and submitted to the Manager -

Corporate Nuclear Safety Section, and the Vice President -

Nuclear Operations.

6.6-1 AMENDMENT NO. 70

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

a. The provisions of 10CFR50.72 shall be complied with.

b. The provisions of 10CFR50.36(c)(1)(i) shall be complied with.

c. The Safety Limit violation shall be reported to the NRC Region II, the Vice President -

Nuclear Operations, and to the Manager -

Corporate Nuclear Safety Section, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d. A Safety Limit Report shall be prepared. The report shall be reviewed in accordance with specification 6.5.1.6.6.

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.

e. The Safety Limit Violation Report shall be submitted to the NRC Region II, Vice President - Nuclear Operations, and the Manager -

Corporate Nuclear Safety Section within 14 days of the violation.

6.-A AMENDMENT NO. 70

0w 6.8 NOT USED 6.8-1 AMENDMENT NO.

70

6.9 REPORTING REQUIREMENTS Information to be reported to the NRC, in addition to the reports required by Title 10, Code of Federal Regulations, shall be as indicated in the following sections.

Reports shall be addressed to the Director of the -appropriate Regional Office of Inspection and Enforcement unless otherwise noted.

6.9.1 Routine Reports

a.

Startup Report. A summary report of plant startup and power escalation shall be submitted following (1) amendment to the license involving a planned increase in power level, (2) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (3) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

The report shall address each of the tests performed related to the startup 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.

6.9-1 AMENDMENT NO. 70

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

i:

'he startup report does not cover all three events (i.e.,

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

b.

Annual Report Prior to March 1 of each year a report.shall be submitted which provides a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exp6sures greater than 100 mrem/yr and their associated man rem exposure according to work and job functions(1), e.g., reactor operations and surveillance, inservice 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 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 work functions.

(1)This tabulation supplements the requirements of §20.407 of 10CFR Part 20.

6.9-2 AMENDMENT NO. 70

c.

Monthly Operating Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis. The report fcrmats set forth in Appendices B, C, and D to Regulatory Guide 1.16 shall be completed in accordance with the instructions provided.

The complete.forms should be sub mnitted by the tenth of the month following the calendar month covered by the report to the Director, Office of Management and Program Analysis, U. S. Nuclear Regulatory Commission, Washington, D. C.

20555, with a copy to the appropriate NRC.

Regional Office.

6.9.2 -

Reportable Occurrences The Reportable Occurrences of Specifications 6.9.2.a and 6.9.2.b below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC.

Supplemental reports may be required to fully describe final resolution of the occurrence.

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

a.

Prompt Notification With Written Followup The types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the appropriate Regional Office of Inspection and Enforcement or his designate no later than the first working day following the event, with a written followup report within two weeks.

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

6.9-3 AMENDMENT NO.

70

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

(1) Failure of the reactor protection system, or other systems subject to limiting safety system settings to initiate the required protective function by'the time a' m2 nitored parameter reaches the setpoint specified as the limiting safety system setting in the Technical Specifications ox failure to complete the required protective function.

Note:

Instrument drift discovered as a result of testing need not be reported under this item (but see 6.9.2.a(5), 6.9.2.a(6), and 6.9.2.b(1) below.

(2) Operation of the unit or affected systems when any parameter or operation subject to a limiting condition.

for operation is less conservative than the least conservative aspect of the limiting condition for operation established in the Technical Specifications.

Note:

If specified action is taken when a system is found to be operating between the most conservative and least conservative aspects of a limiting condition for operation listed in the Technical Specifications, the limiting condition for operation is not 'considered to have been violated and no report need be submitted under this section (but see 6.9.2.b(2) below).

6.9-4 AMENDMENT NO. 70

(3) Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary or primary' containment.

Note:

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

(4) Reactivity anomalies involving disagreement with predicted value.of reactivity balance under steady state**

conditions during power operation greater than or equal to 1% Ak/k; a calculated reactivity balance indicating a shutdown margin less conservative than specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor startup rate greater than 5 dpm, or if subcritical, an unplanned reactivity insertion of more than 0.5% Ak/k; or any unplanned criticality.

6.9-5 AMENDMENT NO. 70

(5) Failure or malfunction to one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.

(6) ?ersonnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional recuirements of systems required to cope with atcidents analyzed in the SAR.

Note:

For 6.9.2.a(5) and 6.9.2.a(6) reduced redundancy that does no-t-result -in loss of system function need not be reported under this section (but see 6.9.2.b(2) and 6.9.2.b(3) below).

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

(8) Errors discovered in.the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the.

Technical Sopecifications that have or could have permitted reactor operation in a manner less conservative than assumed in the analyses.

6.9-6 AMENDMENT NO. 70

(9) Performance of -structures, systems or components that require remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis reDort or Technical Specifications bases or discovery during plant life of conditions not specifically considered in the safety analysis report or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.

Note:

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

b.

Thirty-day Written Reports. The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate NRC Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of the licensee event report form, used for entering data into the NRC's computer-based file of information concerning licensee events.

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

(1) Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those established by the Technical Specifications but which do not prevent the fulfillment of the functional requirements of affected systems (but see 6.9.2.a(1) and 6.9.2.a(2) above).

6.9-7 AMENDMENT NO..70

(2) Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation (but see 6.9.2.a(2) above).

Note:

Roitine surveillance testingp instrument calibration or preventive maintenance which require system configurations described in 6.9.2.b(1) and 6.9.2.b(2) above need not be reported except where test results themselvest.

reveal a degraded mode &s described above.

(3) Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems (but see 6.9.2.a(6) above).

(4) Abnormal degradation of systems other than those specified in 6.9.2.a(3) above designed to contain radioactive material resulting from the fission process.

Note:

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

6 6.9-8 AMENDMENT NO. 70

6.9.3 Special Reports Special reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement within the time period specified for each report.

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

Area Reference Submittal Date

a.

Containment Leak 4.4 Upon completion of Rate Testing each test

b.

Containment Sample 4.4 Upon completion of Tendon Surveillance the inspection at 25 years of operation

c.

Post-operational 4.4.

Upon completion of Containment the test at 20 years Structural Test of operation

d.

Fire Protection 3.14 As specified by System limiting condition for operation.

e.

Overpressure Pro-3.1.2.le Within 30 days of tection System.

operation.

Operation S6.9-9 AMENDMENT NO. 70

b.

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

c.

Records of facility radiation and contamination surveys.

d.

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

e.

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

f.

Records.of transient or operational cycles for -those -facility components designed for a limited number of transients or cycles.

g..

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

h.

Records of in-service inspections performed pursuant to these Technical Specifications.

i.

Records of Quality Assurance activities required by the QA Program.

j.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10CFR50.59.

k.

Records of meetings of the ?NSC and of the independent reviews performed by the Corporate Nuclear Safety Section.

6.10-2 AMENDMENT NO. 70