ML20098E801

Proposed Tech Specs Deleting Listing of safety-related Snubbers,Providing for Functional Testing of 10% Sample of Snubbers & Incorporating Mgt Organizational Changes

ML20098E801
Person / Time
Site:	Point Beach
Issue date:	09/25/1984
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20098E785	List: ML20098E784 ML20098E801
References
TAC-54592, TAC-54593, NUDOCS 8410010392
Download: ML20098E801 (16)

Text

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6.

Direct commiunication between the control room and the operating floor of the containment shall be available whenever changes in core x

geometry are taking place.

7.

The cortainment vent and purge system, including the radiation monitors which initiate isolation shall be tested and verified to be operable immiediately prior to refueling operations.

8.

If any of the specified limiting conditions for refueling are not met, refueling of the reactor shall cease. Work shall be initiated to correct the violated conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be made.

B.

Limitations on Load Movements Over a Spent Fuel Pool

• 1.

A load of 1750 pounds shall be the maximum load allowed over either the north half or south half of the spent fuel storage pool when one or more spent fuel assemblies are stored in that half of the spent fuel pool.

2.

Auxiliary building crane bridge and trolley positive acting limit switches shall be installed to prevent motion of the main crane hook over that half of the spent fuel pool which contains stored spent fuel which has been suberitical for less than one year, i.

e 1

These are interim requirements pending completion and implementation of NRC Generic Task A-36 " Control of Heavy Loads Near Spent Fuel."

8410010392 840925 15.3.8-2 PDR ADOCK 05000266 P

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3.

Loads-greater than 1750 pounds but not exceeding 52,500 pounds may be carried over either pool half (or placed in the north half of the spent fuel pool) provided that no spent fuel assemblies are stored in that half of the spent fuel pool.

Basis The equipment and general procedures to-be utilized during refueling 2 are discussed in the Final Safety Analysis' Report. Detailed instructions, the above specified precautions, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no Lincident could occur during the refueling operations that would result in a hazard to public health and safety.II).

Whenever changes are not being made in core geometry one flux monitor is sufficient. This permits maintenance of the instrumentation. Continuous

. monitoring of radiation levels (A2 above) and neutron flux provides immediate indication of an unssfa condition. The residual heat pump is used to maintain a uniform boron concentration.

The shutdown margin indicated in Part A5 will keep the core suberitical, even if all control rods were withdrawn from the core. During refueling, the reactor refueling cavity is filled with approximately 275,000 gallons of borated water. The boron concentration of this water is sufficient to maintain the reactor 15.3.8-3

15.3.13 SHOCK SUPPRESSORS (SNUBBERS)

- Applicability Applies to the operability of safety related snubbers.

Obj ective To define those limiting conditions for operation for snubbers required to protect the primary coolant system and safety related systems.

Specification 1.

During all modes of operation except Cold Shutdown and Refueling Shutdown all safety related snubbers shall be operable except as noted in items 15.3.13.2 through 15.3.13.4 below.

3 2.

If a snubber is determined to be inoperable, reactor operation may be continued no longer than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the time the snubber was determined to be inoperable.

3.

If the requirements of 15.3.13.1 and 15.3.13.2 cannot be met, an orderly shutdown shall be initiated and the reactor put in the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

p 4.

If a snubber is determined to be inoperable shile the reactor is in the Cold Shutdown or Refueling Shutdown condition, the snubber shall be made operable prior to reactor startup.

15.3.13-1 i

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15 4.13- -Shock Suppressors (Snubbers)

Applicability Applies to the periodic inspection and. testing requirements of safety related

.Caubbers.

Objective To verify the operability of the snubbers.

Specifications The following surveillance requirements apply to safety related snubbers.

1.

All snubbers shall be visually inspected to verify operability in

-accordance with the following schedule:

Number of Snubbers Next Required Found Inoperable Inspection During Inspection Interval or During Inspection Interval 0

18 months i 25%

1 12 months i 25%

2 6 months i 25%

3, 4 124 days i 25%

5, 6, 7 62 days i 25%

8 31 days i 25%

The required' inspection interval shall not be lengthened more than one step at a time.

During each refueling shutdown, a representative sample of approximately 10%

2.

of the snubbers shall be functionally tested for operability. The hydraulic snubber functional test shall verify:

r Activation (restraining action) is achieved within the specified range

-a.

of velocity or acceleration in both tension and compression, b.

Snubber bleed, or release rate, where required, is within the specified range in compression or tension. For snubbers specifically required not to displace under continuous load, the ability of the l

snubber to withstand load without displacement shall be verified.

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15.4.13-1

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15.4.13

- Shock Suppressors (Snubbers) (Continued)

For each snubber found to be inoperable, an additional 10% of c.

that type snubber shall be tested until no more failures are found or all units have been tested.

3..

A record of the service life of 'esch snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained. Concurrent with the next inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance records for all safety related snubber shall be re-viewed to verify that the indicated service life has not been exceeded If the indi-prior to the next schedu' 4 snubber service life review.

cated service life will be exceeded prior to the next scheduled snubber service life review,.the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review.

This reevaluation replacement or reconditioning shall be indicated in the records.

' Bases All safety related snubbers are visually inspected for overall integrity and operability. The inspection will include verification of proper crientation, g.

adequate hydraulic fluid level and proper attachment of snubber to piping To further increase the assurance of snubber reliability, and structures.

functional tests are performed once each refueling cycle on a representative 15.4.13-2 r

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- ALL CORE ALTERATIONS af ter the initial fuel loading e.

shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator LimitedEto Fuel Handling who has no other concurrent responsibilities during this operation, f.

A Fire Brigade of at least 5 members shall be maintained onsite at all times *. This excludes 3 members of the minimum shift crew necessary for safe shutdown of a unit and any personnel. required for other essential functions during a fire emergency.

15.6.2.3 Duty & Call Superintendent To assist and counsel the Shift Superintendent in case s.

of significant operating events, a Duty & Call Super-intendent group has been established. The Duty & Call Superintendent group shall consist of qualified persons designated in writing by the Manager.

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b.

In the event of a reportable occurrence, the Shift Superintendent shall communicate with at least one Duty

& Call Superintendent before taking other than the immediate on-the-spot action required. One Duty & Call Superintendent will be assigned to be "on call" at all times.

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• Fire Brigdde composition may be less than the minimum requirements for a t.

period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of Fire Brigade members provided immediate action is taken to I

restore the Fire Brigade to at least the minimum requirements.

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of the Board i

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Vice Chairman of the Board

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President t

Senior I

Vice President

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l Superintendent j

i Vice President System Security i

l Treasurer Nuclear Power and Fire Protection 1

Superintendent General Superintendent Manager System Fire Risk Management Nuclear Systems Engineering

. Point Beach Protection Officer and Analysis Section Nuclear Plant l

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WISCONSIN ELECTRIC POWER COMPANY OFF-SITE MANAGEMENT FIRE PROTECTION ORGANIZATION

..... Policy, Procedure, Design Coordination Figure 15.6.2-3

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Administrative Organization Fire Protection Organization i

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15.6.5.1.2 The Manager's Supervisory Staff shall be selected from the following:

Chairman: Manager - Point Beach Nuclear Plant Member:

General Superintendent Member:

Superintendent - Operations Member:

Superintendent - Maintenance &

Construction Member:

Superintendent - Engineering, Quality &

Regulatory Services Member:

Superintendent - Training Member:

Superintendent - Technical Services Member:

Superintendent - Chemistry & Health Physics Member:

Superintendent - Reacter Engineerins, Member:

Health Physicist Member:

Superintendent - Instrumentation &

Control

.15.6.5.1.3 Alternate members may be appointed by the MSS Chairman to serve on a temporary basis; however, no more than two alEernates shall vote in MSS at any one time. Such appointment shall be in writing.

15.6.5.1.4 The MSS shall meet at least once per calendar month and as convened by the MSS Chairman.

15.6.5.1.5 A quorum of the MSS shall consist of the Chair-man or his designated alternate and four members including alternates.

15.6.5.1.6 The MSS shall have the following duties:

Review procedures as required by these a.

Technical Specifications. Review other procedures or changes therete which affect nuclear safety as determined by the Manager.

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b.

Review all proposed testa and experiments f

related to nuclear safety and the resulto thereof when. applicable.

15.6.5-2

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

Review all proposed changes to Technical c.

Specifications.

o d.

Review all proposed changes or modifica-tions to plant systems or equipment where changes affect nuclear safety, Periodically review plant operations for e.

nuclear safety hazards.

f.

Investigate violations or suspected viola-tions of Technical Specifications, such investigations to include reports, evalua-tions and recommendations, g.

Perform special reviews, investigations or prepare reports thereon as requested by the Chairman of the Off-Site Review Com-mittee.

h.

Review the Facility Fire Protection Pro-gram and implementing procedures at least once per 24 months.

1.

Investigate, review, and report on all reportable events.

15.6.5.1.7 The Manger's Supervisory Staff shall have the following responsibility:

Serve as an advisory committee to a.

the Manager.

b.

Make recommendations to the Manager for proposals under items (a) through (d) above.

In the event of disagreement between 15.6.5-3

a majority of the Supervisory Staff and decisions.by the Manager, the course of action will be determined by the Manager and the disagreement recorded in the Staff minutes.

c.

Make reconmendations as to whether or not proposals considered by the Staff involve unreviewed safety questions, d.

Review and approve the contents of a report for each reportable event.

Copies of all such reports shall be submitted to the Vic'e President - Nuclear Power and the Chairman of the Off-Site Review Committee.

Written. minutes of each meeting shall be e.

reviewed by staff members and copies shall be provided to the Vice President -

Nuclear Fower and Chairman of the Off-Site Review Committee.

15.6.5-4 i

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15.6.5.3 0FF-SITE REVIEW COMMITTEE (OSRC)

FUNCTIO _N 15.6.5.3.1 The Off-Site Review Committee shall function to provide independent review and audit of designated activities in the areas of:

'a) nuclear power plant operations J

b) nuclear engineering c) chemistry and radiochemistry d) metallurgy e) fastrumentation and control f) radiological safety g) mechanical and electrical engineering h) quality assurance practices i) environmental monitoring COMPOSITION 15.6.5.3.2 The Off-Site Review Committee is made up of a minimum of five regular Of the members appointed by the President and one or more ex-officio members.

five or more regular members, at least two will be persons not directly employed by the Licensee. All members will be experienced in one or more aspects of the nuclear industry.

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ALTERNATES 15.6.5.3.3 Alternate members may be appointed in writing by the OSRC Chairman to serve on a temporary basis; however, no more than two alternates shall parti-cipate in OSRC sctivities at any one time.

' CONSULTANTS' 15.6.5.3.4 Consultants shall be utilized as determined by the OSRC Chairman to provide expert advice to the OSRC.

15.6.5-5

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,7 h) Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

1) Reports and meeting minutes of the Manager's Supervisory Staff.

AUDITS 15.6.5.3.8 Audits of facility activities shall be performed under the cognizance of the OSRC. These audits shall encompass:

a) The conformance cf facility operation to provisions contained within the Technical Specifications and appli-cable license conditions at least once per year, b) The performance, training and qualifications of the licensed operating staff at least once per year.

c) The results of actions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least twice per year at approximately six month intervals.

d) The results of quarterly audits by the Quality Assurance Division on the performance of activities required by the Quality Assurance Program to meet the criteria of Appendix "B",

10 CFR 50, at least once per two years.

e) Any other area of facility operation considered appropriate by ths President.

AUTHORITY 15.6.5.3.9 The OSRC shall report to and advise the President on those areas of responsibility specified in Section 15.6.5.3.7 and 15.6.5.3.8.

15.6.5-7

RECORDS 15.6.5.3.10 Records of OSRC activities shall be prepared, approved and distributed as indicated below:

a)

Minutes of each OSRC meeting shall be prepared, approved and forwarded to the President within 14 days following each meeting.

b)

Reports of reviews encompassed by Section 15.6.5.3.7.e, f and g above, shall be prepared, approved and forwarded to the President within 14 days following completion of the review.

c)

Audit reports encompassed by Section 15.6.5.3.8 above, shall be forwarded to the President and to the management positions responsible for the areas audited within 30 days af ter completion of the audit.

I 15.6.5-8

<0 15.6.6 REPORTABI.E OCCURRENCE ACTION Specification The following action shall be taken for REPORTABLE OCCURRENCES K.

The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 15.6.9.2.

B.

Each REPORTABLE OCCURRENCE requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Commission shall be reviewed by the Manager's Supervisory Staff (MSS) and submitted to the Vice President - Nuclear Power and the Off-Site Review Committee (OSRC).

15.6.6-1

15.6.7 ACTION TO BE TAKEN IF A SAFETY LIMIT IS EXCEEDED Specification A.

If a safety limit is exceeded, the affected reactor shall be shut down and reactor operation shall not be resumed until approval is received from the NRC.

B.

An immediate report shall be made to the Vice President - Nuclear Power and the Chairman of the Off-Site Review Committee.

The Vice President - Nuclear Power shall report the circumstances to the C.

NRC.

D.

A Safety Limit Violation Report including a complete analysis of the cir-cumstances leading to and resulting from the occurrence, effects upon facility components, systems or structures, together with recommendations to prevent a recurrence, shall be prepared. This report shall be submitted to the Vice President - Nuclear Power and the Chairman of the Off-Site Review Committee. A Safety Limit Violation Report shall be submitted to the NRC by the Vice President - Nuclear Power within 10 days of the occurrence.

15.6.7-1

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