ML19257C576

Proposed Changes to Tech Specs 7.1.1,7.1.3 & 7.4 Re Administrative Controls & 4.3.10,Table 4.3.10-1 Re Limiting Conditions for Operations

ML19257C576
Person / Time
Site:	Fort Saint Vrain
Issue date:	01/11/1980
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML19257C560	List: ML19257C559 ML19257C566 ML19257C576
References
NUDOCS 8001290422
Download: ML19257C576 (15)

Text

7.1-2

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

2. A licensed operator must be in the control room at all times when fuel is in the reactor. During reactor startup, shutdown, and recovery from reactor trip, two licensed operators must be in the control room.

3. ALL CORE ALTERATIONS af ter the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reac-tor Operator limited to Fuel Handling who has no other concurrent respon-sibilities during this operation.

4. An operator or technician qualified in radiation protection procedures shall be present at the facility at all times that there is fuel on site.

5. A site Fire Brigade of at least 5 members shall be maintained on site at all times #. The Fire Brigade shall not include (3) members of the minimum shif t crew necessary for safe shutdown of the unit and any per-sonnel required for other essential functions during a fire emergency.

1. 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 accomodate unexpected ab-sence of Fire Brigade members provided immediate action is taken to restore the Fire Brigade ta within the minimum requirements.

Upon commencement of commercial operation the staffing of the plant shall be in accordance with American National Standards Institute N18.1-1971,

" Selection and Training of Personnel for Nuclear Power Plants" M

1827 210 800129o O2

7.1-3 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for co= parable position, except for the Health Physics Supervisor who shall meet or exceed the qualifications of Regu-latory Guide 1.8, September,1975.

A retraining and replacement training program for the facility staf f shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10CFR Part 55. Compliance with Section 5.5 of ANSI N18.1-1971 shall be achieved no later than six months following cor=lence-ment of commercial operation.

A training program for the Fire Brigade shall be maintained under the direction of the Training Supervisor and shall reet or exceed the requirements of Section 27 of the NFPA Code-1975, except for Fire Brigade training sessions which shall be held at least once per 92 days.

Specification AC 7.1.2 - Plant Operations Review Committee (PORC) , Adminis-trative Controls The organization, responsibilities, and authority of the PORC shall be as follows:

a. Membership The Plant Operations Review Cor=11ttee shall be composed of the following:

Chairman: Administrative Services Manager Operations Manager Superintendent Operations Health Physics Supervisor Results Engineering Supervisor Reactor Engineer Technical Services Supervisor ff Shift Supervisor

ATTAGMENT 2 182fr 212

7.1- 7

b. Membership The NFSC shall be composed of the following:

Chairman: Vice President Production Nuclear Project Manage r Manager of Safety and Security Quality Assurance Manager Manager Nuclear Production Consultants, as required and appointed by the Chairman

c. Alte rnates Alternate members shall be appointed in writing by the Chairman; however, no more than two alternates shall participate in NISC activities at any one time.

d. Consultan ts Consultants shall be utilized as determined by the Chairman, NISC, to provide expert advice to the NFSC.

e. Meeting Frequency The NFSC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.

f. Quorum A quorum of the NFSC shall consist of the Chairman or his designated alternate and a majority of the NESC members including alternates.

F 182p213

7.1-9 (h) Any indication that there may be a deficiency in some aspect of design or operation of structures, systems, or components, that affect nuclear safety.

(i) Reports and meeting minutes of the PORC.

2. Audits of fhcility activities shall be performed under the cognizance of the Nuclear Facility Safety Committee. These audits shall encompass:

(a) The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license conditions at least once per year.

(b) The performance, training, and qualifications, of the facility staf f 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 af fect nuclear safety at least once per six months.

(d) 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) The facility Emergency Plan and implementing procedures at least once per two years. .

(f) The facility Security Plan and implementing procedures at least once per two years.

(g) Any other area of facility operation considered appropriate by the NFSC or the appropriate Vice President.

(h) An audit of the Fire Protection Program including a fire protection and loss prevention inspection shall be performed annually, utilizing qualified off site licensee personnel, an outside fire protection firm, or an outside qualified fire consultant. This audit must be performed by an outside qualified fire consultant at invervals no greater than 3 years. -

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ATTACHMENT 4 1827217

7.0-1 7.4 PROCEDURES - ADMINISTRATIVE CONTROLS Applicability Applies to administrative procedures which will govern plant opera-tions .

Objective To ensure that written procedures will be maintained to define re-quirements for plant operation.

Specification AC 7.4 - Procedures, Administrative Controls

a. Written procedures shall be established, implemented and maintained covering the activities referenced below:

1. The applicable procedures recommended in Appendix A of Regula-tory Guide 1.33, November,1972.

2. Refueling operations.

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

4. Security Plan implementation.

5. Emergency Plan implementation.

b. Procedures and administrative policies of a. above, and changes thereto, shall be reviewed by the PORC and approved by the appro-priate Manager prior to implementation and reviewed periodically as set forth in Administrative Procedures.

Security Plan procedures, and changes thereto, shall be reviewed by the Plant Operations Review Committee and approved by the designated Plant Security Officer prior to implementation.

Security Plan procedures and changes thereto, shall be reviewed by the Fort St. Vrain Security Committee.

182$:218

O ATTACHMENT 5 182y219

~. .

Revise Table 4.3.10-1 of LCO 4.3.10 to delete BIS-298E.

Boiler Feed Snubbers - Emergency BFS-14E BFS-89E BFS-219E BIS-399E BFS-15E BIS-122E BIS-228E BFS-405E BFS-16E BFS-141E BIS -229E BIS-414E BFS-26E BFS-142E B FS-243E BFS-417E BFS-29E BFS-143E BFS-244E BFS-419E BIS-30E BFS-158E BFS-245E BIS-421E BFS-31E BIS-167E BFS-257E BFS-422E BFS-47E BFS-181E BFS-260E BIS-423E BFS-53E B FS-19 7E BFS-263E BFS-430E BFS-56E BFS-203E B FS-264E BIS-431E BFS-5 7E BIS-204E BFS-268E BFS-432E BFS-74E BFS-210E BFS-269E BFS-442E BFS-76E B FS-216E BFS-444E BFS-77E BIS-218E BIS-398E Reason for Chance -

Change Notice 473 deleted snubber.

The Dynaflex Computer Program showed a maximum DBE stress of 15,870 psi in the line 10" L-22291 at hanger BF-200 due to a seismic event. This is with-in the yield stress for the pipe. Corresponding displacements were very small and would cause no problem. No large seismic stresses nor displace-ments will occur by removing snubber BIS-298E. The Computer Program showed a maximum DBE stress of 17,381 psi in Line 10" L-21325 near snubber BFS-44E.

This is also within the yield stress of the pipe.

182y220-

ATTAGMENT 6 18?S 22L

SPECIFICATION DF6.1 - REACTOR CORE, DESIGN FEATURES The following discussion describes the design features which shall be incorporated in the reactor core:

Reactor Assembly (Remains unchanged)

Active Core (Remains unchanged)

Fuel 235 The fuel consists of fissile uranium highly enriched (93.15%) in U and fertile thorium. The initial fuel loading is about 773 Kg of uranium and 16,000 Kg of thorium. The initial core is loaded with 13 fuel compositions whose distribution within the core is designed to mock up the fuel content of the equilibrium cycle refueling regions and to shape the radial and axial power distribution. Fuel is designed for up to a six-year life. About one-sixth of the core will be replaced at each refueling interval. The fuel loading in a reload segment will be about 200 Kg of uranium and 2,300 Kg of thorium.

All uranium and thorium in the reference fuel elements is in the form of heavy metal carbide and pyrocarbon, referred to as coated fuel particles. The coatings form the primary fission product barrier. The coated fuel particles consist of two general types, fissile particles (TH:UC 2 ) and fertile (TH C2 ) particles. The fissile particles shall contain thorium and uranium in a weight ratio of about 3.6 to 1 (+1.2,

-0.2) of thorium to uranium. The fertile particles shall contain only thorium.

In addits an to the reference fuel elements, eight test fuel ele-ments are Laciuded in the reactor core. These eight test elements (FTEl-8) con .ain small quantities of test fuel particles that are in various ways different from the reference fuel. The description of the test fuel elements is contained in Table 6.1-1.

The coated fuel particles are bonded together with a carbonaceous material to form fuel rods. The fuel rods are completely surrounded and contained by graphite which forms the structural part of the fuel element and, in addition to the carbon contained within the fuel rods, also serves as the sole moderator. The reference fuel elements are fabricated from H-327 needle coke (anisotropic) graphite, as described in the Fort S t . Vrain FSAR, Section 3.0. The test fuel elements are fabricated from H-451 near-isotropic graphite in anticipation of quali-fying this material for future use in all reload fuel for the reactor.

1821222

Reflector (Remains unchanged)

Basis for Specification DF6.1 (Remains unchanged)

Table 6.1-1 (Remains unchanged)

Reason For Chante The Fort St. Vrain fuel specification was changed in 1971 to per-mit a reduction in the TH:U ratio from 4. 25 /1 (1 0. 5) to 3. 6/1 (+1. 2,

-0.2) However, a revision to the Technical Specifications was inad-vertently omitted. This change will correct that omission.

182$223'