ML20086L361
| ML20086L361 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 01/27/1984 |
| From: | WISCONSIN ELECTRIC POWER CO. |
| To: | |
| Shared Package | |
| ML20086L343 | List: |
| References | |
| GL-83-43, TAC-54200, TAC-54201, NUDOCS 8402080199 | |
| Download: ML20086L361 (8) | |
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APPENDIX A TO FACILITY OPERATING LICENSE DPR-24 s
AND FACILITY OPERATING LICENSE DPR-27 FOR POINT BEACH NUCLEAR PLANT UNIT NOS. 1 AND 2 WISCONSIN ELECTRIC POWER COMPANY DOCKET NOS. 50-266 AND 50-301 t
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15.
TECHNICAL SPECIFICAT:0N AND BASES 15.1 DEFINITIONS The definitions for frequently used terms which are applied to the Point Beach Nuclear Power Plant, Units No. I and 2, are stated below, a.
Reportable Event A Reportable Event shall be any of those conditions specified in Section 50.73 to Title 10 Code of Federal Regulations (10 CFR)
Part 50.
b.
Quadrant Power Tilt Quadrant to average power tilt is expressed in percent as de-fined by the following equation:
p wer in any core quadrant
_3 100 x
average for all quadrants c.
Operability A system, subsystem, train, component, or device shall be operable or have operability when it is capable of performing its function (s) as' analyzed in the safety analysis report.
Implicit in this definition is the assumption that necessary instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubricatian or other auxiliary equipment required for the system, subsystem, train, component or device to perform its function (s) are capable of performing their related support function (s).
d.
Containment Integrity *_
Containment integrity is defined to exist when:
- 1) All non-automatic containment isolation valves and blind flanges are closed as required.
15.1-1 l
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- 2) The equipment hatch is properly closed.
- 3) At least one door in each personnel air lock is properly closed.
- 4) All automatic containment isolation valves are operable or are secured closed.
- 5) The uncontrolled containment leakage satisfies Specifi-cation 15.4.4.
e.
Protective Instrumentation Logic
- 1) Analog Channel An analog channel is an arrangement of components and modules as required to generate a single protective action signal when required by a plant condition. An analog channel loses its identity where single action signals are combined.
1
- Containment isolation valves are discussed in FFDSAR Section 5.2.
15.1-2
b)
Review all proposed tests and experiments related to safety and the results thereof when applicable.
c)
Review all proposed changes to Technical Specifications, d)
Review all proposed changes or modifications to plant systems or equipment where changes would require a change in operating or emergency procedu'res or that affect nuclear safety.
e)
Periodically review plant operations for industrial and nuclear safety hazards.
f)
Investigate violations or suspected violations of Technical Specifications, such investigations to include reports, evaluatior.s, and recommendations to prevent recurrence, to the Executive Vice President and to the Chairman of the Off-Site Review Committee, g)
Perform special reviews and investigations and prepare reports thereon as requested by the Chairman of the Off-Site Review Committee.
h)
Investigate, review, and report or, all reportable events.
i)
Cause to be conducted periodic drills on emergency procedures, including evacuation (partial or complete) of the site and check adequacy of communications with off-site support groups.
j)
Review the Facility Fire Protection Program and implementing procedures at least once per 24 months.
AUTHORITY 15.6.5.2.7 a)
The Supervisory Staff shall serve as advisory to the Manager-Nuclear Operations.
b)
The Supervisory Staff shall recommend to the Manager approval or disapproval of proposals under items a) through d) above.
In the event of disagreement between a majority of the 15.6.5-3
7 e
Supervisory Staff and decisions by the Manager, the course of action will be determirid by the Manager and the dis-agreement recorded in the Staff minutes. Records of the disagreement will be included in the minutes sent for review to the Off-Site Review Committee and the Executive Vice President.
c)
The Supervisory Staff shall make tentative recommendations as to whether or not proposals considered by the Staff involve unreviewed safety questions. These recommendations shall be subject to review and further recommendations by the Off-Site Review Committee. Minutes shall be kept of all meetings of the Staff and copies shall be sent to the Executive Vice President and to the Chairman of the Off-Site Review Committee.
d)
The Supervisory Staff shall review and approve the contents of a report for each reportable event. This report shall be prepared and submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50. Copies of all such reports shall be submitted to the Executive Vice President and to the Off-Site Review Committee.
RECORDS 15.6.5.2.8 The Manager's Supervisory Staff shall maintain written minutes of each meeting and copies shall be provided to the Executive Vice President and Chairman, Off-Site Review Committee.
15.6.5-4
MEETING FREQUENCY 15.6.5.3.5 The OSRC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least twice per year at approximately six month intervals thereafter.
QUORUM 15.6.5.3.6 A quorum of OSRC shall consist of the Chairman or his desig-nated alternate and three members.
No more than a minority of the quorum shall have line responsibility for operation of the facility.
REVIEW 15 6.5.3.7 The OSRC shall review:
a)
The safety evaluations for 1) changes to procedures, equipment or systems, and 2) tests or experiments com-pleted under the provision of 10 CFR, Section 50.59, to verify that such actions did not constitute an unreviewed safety question.
b)
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in 10 CFR, Section 50.59.
c)
Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR Section 50.59.
d)
Proposed changes in Technical Specifications or Licenses, e)
Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f)
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
g)
All reportable events.
15.6.5-6
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15.6.6 Reportable Event Action Specification The following action shall be taken for Reportable Events:
A.
The Connission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and B.
Each Reportable Event shall be reviewed by the Manager's Supervisory
' Staff (MSS) and submitted to the Off-Site Review Committee (OSRC) and l
the Vice President - Nuclear Power.
1 15.6.6-1
15.6.9.2 Deleted
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15.6.9.3 Unique Reporting Requirements The following written reports shall be submitted to the Director, Office of Nuclear Reactor Regulation, USNRC:
A.
Each integrated leak test shall be the subject of a summary technical report, including results of the local leak rate tests and isolation valve leak rate tests since the last report.
The report shall include analysis and interpreta-tions of the results which demonstrate compliance with specified leak rate limits.
B.
Deleted C.
Submission of a report within 60 days after January 1 and after July 1 each year for the six-month period or fraction thereof, ending June 30 and December 31 containing:
l-l 15.6.9-4 i
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