ML20210E979

Biweekly Notice Re Applications & Amends to OLs Involving NSHCs

ML20210E979
Person / Time
Issue date:	02/05/1987
From:	Houston R Office of Nuclear Reactor Regulation
To:
References
NUDOCS 8702100416
Download: ML20210E979 (112)

Text

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l 7590-01 i

NUCLEAR REGULATORY COMMISSION BI-WEEKLY NOTICE APPLICATIONS AND AMENDMENTS TO OPERATING LICENSES INVOLVING NO SIGNIFICANT HAZARDS CONSIDERATIONS I. Backaround Pursuant to Public Law (P.L.)97-415, the Nuclear Regulatory Comission (the Comission) is publishing this regular bi-weekly notice. P.L.97-415

- revised section 189 of the Atomic Energy Act of 1954, as amended (the Act),

to require the Comission to publish notice of any amendments issued, or proposed to be issued, under a new provision of section 189 of the Act.

This provision grants the Comission the authority to issue and make im-mediately effective any amendment to an operating license upon a determina-tion by the Comission that such amendment involves no significant hazards consideration, notwithstanding the pendency before the Comission of a request for a hearing from any person.

This bi-weekly notice includes all amendments issued, or proposed to be issued, since the date of publication of the last bi-weekly notice which was l published on January 28, 1987 (52 FR 2870) through February 2, 1987.

8702100416 PDR ORO 870205

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NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE AND PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION AND OPPORTUNITY FOR HEARING The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the 1 Comission's regulations in 10 CFR 50.92, this means that operation of the facility in accordance with the proposed amendments would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or'different kind of accident from any accident previously evaluated; or (3) involve a signifie r t reduction in a margin of safety. The basis for this proposed detr.rmination for each amendment request is shown below.

4 The Commission is seeking public coments on this proposed determina-tion. Any comments received within 30 days after the date of publication of this notice will be considered in making any final determination. The Commission will not normally make a final determination unless it receives a request for a hearing.

Written comments may be submitted by mail to the Rules and Procedures Branch, Division of Rules and Records, Office of Administration, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, and should cite the publication date and page number of this FEDERAL PEGISTER notice. Written comments may also be delivered to Room 4000, Maryland National Bank Building, 7735 Old Georgetown Road, Bethesda, Maryland from 8:15 a.m. to 4

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5:00 p.m. Copies of written coments received may be examined at the NRC Public Document Room, 1717 H Street, NW, Washington, DC. The filing of requests for hearing and petitions for leave to intervene is discussed below.

By March 13, 1987, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written petition for leave to intervene. Requests for a hearing and petitions for leave to intervene shall be filed in accordance with the Commission's " Rules of Practice for Domestic Licensing Proceedings" in 10 CFR Part 2. If a request for a hearing or petition for leave to intervene is filed by the above date, the Commission or an Atomic Safety and Licensing Board, designated by the Commission or by the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of hearing or an appropriate order.

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As: required by 10.CFR 62.714, A. petition for leave-to intervene shall set forth with particularity the interest of the petitioner in the proceeding, and how that interest may be affected by the results of the proceeding. The petition shoulo specifically explain the reasons why intervention should be permitted with particular reference to the following l

l factors: (1) the nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial, or other interest in the proceeding; and (3) the l

possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect (s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to i l

intervene or who has been admitted as a party may amend the petition without i requesting leave of the Board up to fifteen (15) days prior to the first prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than fifteen (15) days prior to the first prehearing

- conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter, and the bases for each contention set forth with reasonable specificity. Contentions shall be .

limited to matters within the scope of the amendment under consideration. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to particip6te as a party.

.. ,.Thosa_pennitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.

If a hearing is requested, the Commission will make a final determination on the issue of no significant hazards consideration. The final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves no 1

significant hazards consideration, the Commission may issue the amendment i

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and make it imediately effective, notwithstanding the request for a hearing. Any hearing held would take place after issuance of the amendment.

If the final determination is that the amendment involves a significant hazards consideration, any hearing held would take place before the issuance of any amendment.

Normally, the Comission will not issue the amendment until the expira-tion of the 30-day notice period. However, should circumstances change during the notice period such that failure to act in a timely way would result, for example, in derating or shutdown of the facility, the Comission may issue the license amendment before the expiration of the 30-day notice period, provided that its final detennination is that the amendment involves no significant hazards consideration. The final determination will consider all public and State comments received before action is taken. Should the

. Commission take this action, it will publish a notice of issuance and provide for opportunity for a hearing after issuance. The Comission expects that the need to take this action will occur very infrequently.

A request for a hearing or a petition for leave to intervene must be l

filed with the Secretary of the Co:nmission, U.S. Nuclear Regulatory l

l Comission, Washington, D.C. 20555, Attention: Docketing and Service i

Branch, or may be delivered to the Comission's Public Document Room,1717 H Street, N.W., Washington, D.C., by the above date. Where petitions are filed during the last ten (10) days of the notice period, it is requested that the petitioner promptly so inform the Comission by a toll-free telephone call to Western Union at (800) 325-6000 (in Missouri (800) 342-6700). The Western Union operator should be given Datagram Identification Number 3737 and the following message addressed to

(Branch Chief): petitioner's name and telephone number; date petition was mailed; plant name; and publication date and page number of this FEDERAL REGISTER notice. A copy of the petition should also be sent to the Executive Legal Director, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be entertained absent a determination by the Commission, the presiding officer or the presiding Atomic Safety and Licensing Board, that the petition and/or request should be granted based upon a balancing of factors specified in 10 CFR 2.714(a)(1)(1)-(v) and 2.714(d).

For further details with respect to this action. see the application for amendment which is available for public inspection at the Commission's Public Document Room, 1717 H Street, N.W., Washington, D.C., and at the local public document room for the particular facility involved.

Arizona Public Service Company et al., Docket Nos. STN 50-528 and STN 50-529, Palo Vsids Nuclear Gensritind'Stition'(PVNGS),' Units 1"and 2f Marif6pa County,

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Arizona Date of amendment request: January 19, 1987 Description of amendment request: The proposed amendments consist of a proposed change to the Technical Specifications (Appendix A to Facility Operating License Nos. NPF-41 for PVNGS, Unit 1 and NPF-51 for PVNGS, Unit 2).

The proposed change would revise Table 3.6-1 in the Technical Specifications to remove the four main steam line isolation valves (MSLIV)

from the list of valves that are subjected to the requirements of Technical Specification 3/4.6.3, " Containment Isolation Valves". Technical Specification 3/4.7.1.5, " Main Steam Line Isolation Valves," specifically provides operability and surveillance requirements for the MSLIVs which are consistent with the assumptions for MSLIV operability used in the PVNGS safety analyses. Removal of the MSLIVs from Table 3.6-1 would provide an additional four hours to restore an inoperable valve to operable status before a plant power reduction is required. However, the proposed amendment would not change the valve design, performance requirements, or any of the

- instrumentation and control settings associated with these valves.

At the request of the licensees, the change is also being considered by the NRC staff in the development of the Technical Specifications for PVNGS, Unit 3 (STN 50-530), which is currently under review for an operating license. PVNGS, Unit 3 is of the same design as PVNGS, Units 1 and 2.

Basis for Proposed No Sianificant Hazards Consideration Determination: The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility involves no significant hazards considerations if operation of the facility in accordance with a proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

A discussion of the proposed change, as it relates to these standards is presented below.

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Standard 1 - Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated The proposed amendments do not change any of the design requirements or accident evaluations involving the MSLIVs. The change wculd only delete the valves from the requirements of Technical Specification 3/4.6.3, however, Technical Specification 3/4.7.1.5 will continue to provide requirements which are consistent with the assumptions used in the PVNGS safety analyses.

Therefore, the proposed amendments do not significantly increase the probability or consequences of an accident previously evaluated.

Standard 2 - Create the Possibility of a New or Different Kind of Acciden't From Any Accident Previously Evaluated The proposed amendments would not vary or affect any plant operating condition or parameter. Therefore, the amendments do not create the possibility of a new or different kind of accident previously evaluated.

Standard 3 - Involve a Sionificant Reduction in a Maroin of Safety The proposed amendments would not change any of the design basis or operating conditions for the plants. Therefore, the proposed amendments do not involve a significant reduction in any margins of safety. . ,

Based on the above considerations, the Commission proposes to determine that the proposed change does not involve a significant hazards consideration.

Local Public Document Rnom location: Phoenix Public Library, Business, Science and Technology Department, 12 East McDowell Road, Phoenix, Arizona 85004.

Attorney for licensees: Mr. Arthur C. Gehr, Snell & Wilmer, 3100 Valley Center, Phoenix, Arizona 85007.

NRC Pro,1ect Directorate: George W. Knighton I

Arizona Public Service Company et al., Docket Nos. STN 50-528 and STN 50-529, Palo Verde Nuclear Generating Station (PVNGS), Units 1 and 2. Maricopa County, Arizona Date of amendment request: January 23, 1987 Description of amendment request: The proposed amendments consist of a proposed change to the Technical Specifications (Appendix A to Facility Operating License Nos. NPF-41 for PVNGS, Unit 1 and NPF-51 for PVNGS, Unit 2). The proposed change would make the surveillance requirements in

- Technical Specification 4.6.4.2 for the hydrogen recombiner power control cabinets consistent with the expanded testing reofmen prescribed by the j vendor. The function of the hydrogen recombiner is to maintain the hydrogen concentration within containment below its flammable limit in the event of a loss-of-coolant accident. The purpose of the surveillance requirements is to ensure the operability of the equipment in the event that it is needed.

l At the request of the licensees, the change is also being considered by the NRC staff in the development of the Technical Specifications for PVNGS, l Unit 3r(4T4 AQ d lQ,), g igh ,4s gur_,r gt)y,,ygygr rg q q.4 f grjlp ,pp y,g i g n m q e

l license. PVNGS, Units 1, 2 and 3 will use the same hydrogen recombiner system.

l Basis for Proposed No Significant Hazards Consideration Determination: The Commission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing l

certain examples (51 FR 7751) of amendments that are considered not likely to involve significant hazards considerations. Example (ii) in 51 FR 7751 is a change that constitutes an additional limitation, restriction or i

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control not presently included in the Technical Specifications; e.g., a more stringent surveillance requirement. The proposed amendment request is similar to Example (ii) in 51 FR 7751 since the request involves expanded surveillance testing for the hydrogen recombiner power control cabinets to be consistent with vendor recommendations. Therefore, the Commission proposes to determine that the proposed amendment does not involve any significant hazards considerations.

Local Public Document Room location: Phoenix Public Library, Business, Science and Technology Department,12 East McDowell Road, Phoenix, Arizona

. 85004.

Attorney for licensees: Mr. Arthur C. Gehr, Snell & Wilmer, 3100 Valley Center, Phoenix, Arizona 85007.

NRC Project Director: George W. Knighton Arkansas Power and Light Company, Docket No. 50-313, Arkansas Nuclear One, Unit No. 1, Pope County, Arkansas Date of amendment request: November 17, 1986 Description;of, amendment request:gNfg,,Geneyjen ytter 83-43 dated December 19, 1983, requested licensees to amend their Technical Specifications (TSs) to roflect changes in reporting requirements of 10 CFR 50.72 and 50.73. A model TS was enclosed showing revisions to be made in the " Administrative Control" and " Definitions" sections of the TSs. The generic letter further requested that other conforming changes to TSs be made in order to reflect the revised reporting requirements.

The purpose of this amendment request is to revise the reporting requirements of the TSs for Arkansas Nuclear One, Unit 1, to be consistent with the rule changes to 10 CFR 50.72 and 50.73.

Basis for proposed no significant hazards consideration determination: The Consnission's staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. The Commission has provided guidelines pertaining to the application of the three no significant hazards consideration standards by listing specific examples in 51 FR 7750. The proposed amendment is being made to comply with

. reporting requirements in 10 CFR 50.72 and 50.73. The proposed amendment is in the same category as example (vii) of amendments that are considered not likely to involve significant hazards considerations, i.e., a change to make a license conform to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations.

Based on the above, the Commission makes a proposed determination that this amendment request does not involve significant hazards considerations.

Local Public Document Room location:, Tomlinson Library, Arkansas Tech University, Russellville, Arkansas 72801 Attorney for licensee: Nicholas S. Reynolds, Bishop, Liberman, Cook, Purcell and Reynolds, 1200 17th Street, N.W., Suite 700, Washington, D.C.

20036 NRC Project Director: John F. Stolz

Carolina Power & Light Company, Docket No. 50-325 Brunswick Steam Electric Plant, Unit No.1. Brunswick County, North Carolina Date of application for amendment: December 16, 1986 Description of amendment request: The proposed amendment would change the Technical Specifications (TS) for Brunswick Steam Electric Plant, Unit 1 (Brunswick-1). The proposed change to TS Table 3.6.3-1 revises the description of valve 1-CAC-V172 from drywell purge exhaust isolation valve to suppression chamber purge exhaust isolation valve.

The proposed revision reflects modifications being made to the .

containment atmospheric dilution (CAD) system during the upcoming Brunswick-1 Reload 5 outage. The Brunswick-1 CAD system was previously modified to meet the requirements of NUREG-0737 Item II.E.4.1 during the Reload 4 outage. As installed, the primary means of containment purging is through purge exhaust isolation valve 1-CAC-V172, which discharges from the drywell through a 2-inch bypass line to the standby gas treatment system.

The proposed change modifies the CAD system so that containment purging takes place through the suppression chamber rather than the drywell. This designntreconsistentcwithtthatcapprovediforaBrunswick-2teniMays5ge19861when l Amendment 125 to Brunswick-2 TS was issued.

Basis for proposed no significant hazards consideration determination: The

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Commission has provided standards for detemining whether a significant hazards detemination exists as stated in 10 CFR 50.92(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 i

. or consequences of an accident previously evaluated, or (2) Create the possibility'of a new or different kind of accident from any accident previously evaluated, or (3) Involve a significant reduction in a margin of safety.

The licensee has evaluated the proposed amendment against the standards in 10 CFR 50.92 and has determined the following:

1. The proposed amendment does not involve a sionificant increase in the probability or consequences of an occident previously evaluated. The CAD system provides fN post-accident control of i combustible gases. Since this system provides an accident mitigation function, its modification can not involve an increase

- in the probability of an accident previously evaluated. The proposed modification allows for controlled venting of the .

containment through the suppression chamber under post-accident conditions. This is the preferred venting pathway because the suppression chamber water will act as a scrubbing agent for potential gaseous radioactivity, thereby, reducing the consequences of an accident involving a radioactive release to the primary containment. The modification does not affect the redundancy or the single failure design of the CAD system. The i

modified piping and components will be safety grade and the electrical equipment will conform to Class IE requirements. The modified CAD system meets the requirements of NUREG-0737. Item II.E.4.1, and its design is consistent with that approved for Brunswick-2 on May 5, 1986, when Amendment 125 to Brunswick-2 TS was issued.

2. The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The CAD system provides post-accident combustible gas control through purging of the containment. Post-accident venting of the l containment can still be accomplished via three pathways, only one of which is necessary during post-accident operation. The modification merely makes the primary pathway through the suppression chamber rather than from the drywell, this is the preferred venting pathway because the suppression chamber water will act as a scrubber for potential gaseous radioactivity. This design has been reviewed and approved by the NRC for Brunswick-2 on May 5, 1986, when Amendment 125 to the Brunswick-2 TS was <

i issued.

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3. The proposed amendment does not involve a significant reduction in a margin of safety. Post-accident venting of the containment through the suppression chamber water provides scrubbing of potential radio-active gases, thereby, reducing the activity of potential releases. The modification does not affect the redundancy or the single failure design of the CAD system. The r modified piping and components will be safety grade and the electrical equipment will conform to Class 1E requirements. As such, the proposed amendment will increase the margin of safety.

Based on the above reasoning, the licensee has detennined that the proposed amendment does not involve a significant hazards consideration.

The NRC staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. Based on this review, the staff therefore proposed to detennine that the propos'ed amendment does not involve a significant hazards consideration.

Local Public Document Room location: University of North Carolina at Wilmington, William Madison Randall Library, 601 S. College Road, Wilmington, North Carolina 28403-3297.

Attorney for licensee: Thomas A Baxter, Esquire, Shaw, Pittman, Potts &

Trowbridge, 2300 N Street, N. W., Washington, D. C. 20037.

NRC Project Director: Daniel R. Muller (np+ + >ho maa tn &rreenm a Comonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, Byron Station Units 1 and 2. Ogle County, Illinois Date of application for amendments: October 29, 1986 1

Description of amendments request: The amendment would add Technical Specifications to require the availability of the essential service water system crosstie between Units 1 and 2 whenever either or both units are in i Modes 1, 2, 3 or 4. This cmendment is required because a Probabilistic Risk l

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Assessment done by the staff determined that the two-pump ESW system for each unit was the main contributor to a core melt frequency estimate of 10-3/ year . Havino an ESW pump available from the other unit reduces this core melt frequency. This amendment supplements the August 15, 1986 amendment request which was noticed in the FEDERAL REGISTER on September 24, 1986 (51 FR 33946).

Basis for proposed no significant hazards consideration detemination: The staff has evaluated this proposed amendment and detemined that it involves no significant hazards considerations. According to 10 CFR 50.92fc), a .

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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 or consequences of an accident previously evaluated; or

2) Create the possibility of a new or different kind of accident from any

accident previously evaluated; or 31 Involve a significant reduction in a margin of safety.

! This proposed amendment is a new technical specification which requires the availability of the essential service water system crosstie whenever either or both units are in Modes 1, 2, 3 or 4. The proposed amendment only addresses the probability and consequences of a loss of all essential service water to one of the units. This event has not been previously i

evaluated. The probability or consequences of accidents which have been previously evaluated are not affected by requiring the availability of additional components in the essential service water system.

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As stated above, the sole purpose of this amendment is to address loss of all essential service water to one of the units. Loss of all essential service water to one of the units is a new or different kind of accident which has not been previously evaluated. Instead of creating this accident, this proposed amendment minimizes the possibility of its occurrence.

j Since this proposed amendment requires additional availability of components in the essential service water system, there is no reduction in any margin of safety.

Based on the preceding assessment, the staff believes this proposed ,

amendment involves no significant hazards considerations.

Local Public Document Room location: Rockford Public Library, 215 N. Wyman l

l Street, Rockford, Illinois 61103 l

l Attorney to licensee: Michael Miller, Isham, Lincoln & Beal, One First i National Plaza 42nd Floor, Chicago, Illinois 60603 NRC Project Director: Steven A. Varga l Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, Byron Station, Unit Nos. I and 2. Oole County, Illinois Date of application for amendments: January 6, 1987.

Description of amendments reouest: The amendment would revise Technical Specification Design Features Section 5.3.1 on Page 5-4. The change would allow for the reconstitution of a fuel assembly by insertion of a filler rod fabricated from stainless steel or Zircaloy-4 or leaving a vacancy when a rod is leaking or failed.

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Basis for proposed no sianificant hazards consideration determination: The staff has evaluated this proposed amendment and determined that it involves no significant hazards consideration. According to 10 CFR 50.92(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 or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The proposed amendment revises the Design Features Section 5.3.1 ,

regarding the fuel assemblies description to allow for fuel assembly reconstitution. The revision allows for each fuel assembly to contain 264 l fuel rod locations. A fuel rod location may consist of a fuel rod, a filler rod, or a vacancy as determined in accordance with the cycle specific reload l

analysis.

The reconstituted fuel assemblies meet essentially the same design o ..> u. i c ous.co ..o 3 .s... . . w ..< . .w. , u 3 w.,3 . u e, n . o.. u n w ....... u .u... u

• requirements, satisfy the same design criteria as the original fuel assembly, and the use of reconstituted assemblies will not result in a change to existing safety criteria and design limits. Therefore, they do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The actual reconstitution of a fuel assembly and the procedures by I which it is accomplished will be in accordance with the approved cycle l

specific reload analysis. Separate 10 CFR 50.59 safety evaluations will be performed to verify the reconstitution process does not involve an unreviewed safety question before it is implemented. The reload safety ,

I analysis will evaluate the effect of the actual reconstitution on core operation of equipment important to plant safety. Revising the Technical Specification ensures that if reconstitution is required and approved, it i

would not be inconsistent with the fuel assembly design description wording currently in the Technical Specifications. Since no more than one assembly is reconstituted at one time, the consequences of an accident are bounded by the fuel handling accident which is the most severe accident related to fuel manipulation. This proposed amendment will not create the possibility of a new or different kind of accident.

The margin of safety as defined in the Technical Specification basis has not been reduced since the existing safety criteria design limits will not be changed. Allowing the use of reconstituted assemblies in Design Features Section 5.3.1 does not directly affect any safety system or the 1

safety limits, and thus, does not affect the plant margin of safety.

Therefore, based on the above considerations the-staff-has determined that these changes do not involve significant hazards considerations.

Local Public Document Room location: Rockford Public Library, 215 N. Wyman Street, Rockford, Illinois 61103.

Attorney to licensee: Michael Miller, Isham, Lincoln and Beale, One First National Plaza, 42nd Floor, Chicago, Illinois 60603.

NRC Project Director: Steven A. Varga 1

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station, Unit Nos. I and 2, Benton County, Illinois Date of application for amendments: December 5, 1986

, Description of amendments request: Technical Specification Sections 3.3.2, and 4.3.2 contain the reactor coolant system heatup and cooldown limitation curves that are presently applicable to 8 effective full power years (EPFY).

Zion Units 1 and 2 are approaching 8 EFPY and require revisions to the curves for continued operation. Additionally, portions of Sections 3.3.2, and 4.3.2 are being updated to reflect current methods, conditions, and analyses used in generating the new heatup and cooldown curves. Specific reasons for the updating and explanation of changes follow:

1. Bases: Changes to the heatup, cooldown curves have been made to conform with requirements of the Standardized Technical Specifications.

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2. Page 79, Specification 3.3.2.A.1.a: Heatup rates are changed to reflect the revised heatup curve (p.84) to extend curve applicability to 15 EFPY.

3. Page 84: The heatup curve is being updated to derive the 15 EFPY curve as addressed in WCAP-11247.

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4. Page 85: The cooldown curve is being updated to derive the 15 EFPY curve as addressed in WCAP-11247.

5. Page 86: The current figure 3.3.2-3 has been removed and replaced with a Neutron Fluence vs EFPY figure for Unit 1. Information pertaining to the old figure is contained in WCAP's-10902, 11247, and 7924-A. This

! information is required only for deriving the limitation curves and is used in the WCAP's.

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Page 87: Figure 3.3.2-4 is being replaced by more accurate curves for

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Unit 2 which were used in the analysis of the 15 EFPY heatup, cooldown curves. Actual derivation of the fluence curves is explained in WCAP-10902 and are referenced in WCAP-11247. This information also j applies to the new figure 3.3.2-3.

7. Page 88, 89: Table 3.3.2-1 is being replaced by more current infonnation used in deriving the limitation curves. The values in the' x 50 ft-lb/35 MIL and TRANS USE columns are different due to the use of methods in NUREG-0800. These values replace the old ones which were ,

generated by Westinghouse and were accepted by the NRC at that time.

Some of the values in the RTNDT column changed due to influence of the 50 ft-lb/35 MIL column. The table currently in the Technical

Specifications is only for Unit 1; no information was available at that time for Unit 2. These new tables are referenced on page 90 of the proposed Technical Specifications.

8. Pages 90-93: The bases section on fracture toughness properties is being changes to conform with the Standardized Technical -

Specifications. -Changes in this section were also made to correspond with the proposed LCO section (e.g., 15 EFPY curves, figures, tables).

It should also be noted that equation 2.0 Kyg + 1.25 kit is less than

or equal to K has been modified to reflect the equation contained in IR the Standardized Technical Specifications. ,

9. Page 93a: The listing of references has been updated.

Basis for proposed no significant hazards consideration determination

The Comission has provided standards for detennining whether a significant '

i hazardsconsiderationexists(10CFR50.92(c)). A proposed amendment to an i

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A operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed L amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possiblity of a new or different kind of accident from any accident l

previously evaluated; or (3) involve a significant reduction in a margin of

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The licensee provided the following discussion regarding the above three criteiia:

Criterion 1 The purpose of this amendment request is to update the pre-existing heatup and cooldown limitations as the Zion Units accumulate exposure. The purpose of these curves is to ensure that the reactor vessel is not subjected to excessive levels of stress during the heatup and cooldown phases of reactor operation.

The revised limitations provide an equivalent level of protection to the previous limitations. The acceptance criteria for the calculations

= performed have-not-been sJgnifJcantly. altered....Thus, there.will.be. nom._

change in the probability of vessel failure through crack propagation.

This amendment will not effect the performance of any Zion's safety systems or structures beyond ensuring the continued integrity of Zion reactor vessels as discussed above. Thus, the consequences of all previously evaluated accidents will be unaltered.

Criterion 2 The updating of the administrative controls regarding the heatup and cooldown rates allowable at Zion Station has no effect on any of Zion's i

P systems or stru'ctures. In addition, the imposition of a more c oservative heatup and cooldown rate will not interact with any other phase of Zion's operation.

The analyses contained in the Zion FSAR were examined for any potential alterations . Based upon the lack of system and component interaction discussed above, the specific accident sequences will not be affected.

Thus, this revision of the current limitations to include the higher levels of exposure will not create the possibility of any new or different kind of accident.

. Criterion 3 WCAP-112a7 addressed the criteria for the acceptability of these calculations. The revised heatup and cooldown limitations for Zion Station provide an equivalent level of safety to that which currently exists.

The allowable stresses that the reactor vessel could be subjected to i

have not been altered from the currently existing levels. Thus, there will l be no change in the margin of safety at Zion Station.

The staff has reviewed the licensee's no significant hazards l (t0RSVeratdpgjfetpination c and agrees with the licensee's analysis. The proposed amendment updates the exposure limitations for the heatup and cooldown curves at Zion Station from eight EFPY to fifteen EFPY. The mechanical r.cceptance criteria for the allowable stresses has not been l

significantly altered. Thus, example (i) is applicable in this instance.

Example (i) reads as follows:

(i) A purely administrative change to Technical Specifications; for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.

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Accordingly, the Commission proposes to detennine that the proposed changes to the Technical Specification involve no significant hazards consideration.

Local Public Document Room location: Waukegan Public Library, 128 N. County Street, Waukegan, Illinois 60085.

Attorney to licensee: P. Steptoe, Esq., Isham, Lincoln and Beale, Counselors at Law, Three First National Plaza, 51st Floor, Chicago, Illinois 60602.

NRC Project Director: Steven A. Varga Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion Nuclear Power Station, Unit Nos. I and_2, Bec. ton County, Illinois Date of application for amendments: December 16, 1986 Description of amendments request: These amendments provide clarifications, corrections of typographic errors and errors of inadvertent ommissions or

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deletions in the recently approved Zion Radiological Effluent Technical Specifications (RETS).

Basis for proposed no sianificant hazards consideration determination: n The Commission has provided standards for determining whether a significant hazardsconsiderationexists(10CFR50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1)Involveasignificantincreaseintheprobability or consequence of an accident previously evaluated; (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

The licensee provided the following discussion regarding the above three criteria:

Criterion 1 The proposed change involves clarifications to the exiting RETS and the correction of minor typographical errors. None of the proposed changes involve the alteration of any definition or of any limiting condition for opera tion.

None of these administrative clarifications will effect any of Zion's systems or the integrity of any of its structures. In addition, these clarifications will not affect the probability of occurrence of an external event such as a tornado or an earthquake.

Since these clarification will not involve any system interaction, will not alter Zion's operation, or affect the performance of Zion's structures, the pre-existing safety analysis will remain valid. Thus, this proposed change will not involve a significant increase in the probability or consequences of any accident previously evaluated.

Criterion 2 AW'8Tstested'8Bev4;'these ddminittP#tive=ctartffcattentewfit' hot affect the performance of any of Zion's systems or structures. They are

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administrative clarifications or corrections of minor errors.

These clarifications could not conceivably have any significant effect on Zion's operation during either normal or abnormal conditions. The specific accident sequences contained in Zion Safety Analysis have been reviewed. Based upon the lack of interaction discussed above, none of these sequences will be altered by this proposed change.

Thus, the proposed amendments will not create the possibility of a new or different kind of accident from any accident previously evaluated.

Criterion 3 These clarifications and corrections will not affect the perfonnance of any of Zion's systems or structures. Thus, all of Zion's systems will continue to perform their intended functions.

Since all of Zion's individual components and systems will continue to perform their intended safety function, there can be no change in the plant's overall performance during normal and abnormal operations. Thus, this proposed change will not involve a significant reduction in the margin of safety.

. The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis. This proposed change consists of administrative clarifications and minor corrections. Thus, example (i) is applicable in this instance. Example (i) reads as follows:

(i) A purely administrative change to technical specifications: for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.

Accordingly, the Commission proposes to determine that the proposed l changes to the Technical Specification involve no signi'fcant hazards consideration.

i Local Public Document Room location: Waukegan Public Library, 128 N. County l Street, Waukegan, Illinois 60085.

Attorney to licensee: P. Steptoe, Esq., Isham, Lincoln and Beale, i Counselors at Law, Three First National Plaza, 51st Floor, Chicago, Illinois 60602.

I NRC Project Director: Steven A. Varga l

Consolidated Edison Company of New York, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2 Westchester County, New York Date of amendment request: November 26, 1986 Description of amendment request: The proposed amendment would revise the Technical Specifications for Indian Point Unit Nos.1 and 2 to incorporate changes to the Facility Organization. The amendment would revise the organ-izational figures contained in the Technical Specifications and revise the affected titles of the members of the Station Nuclear Safety Committee.

As a result of organizational restructuring, the offsite Technical .

Support Organization is now known as the offsite Independent Safety Review and Audit Organization. The responsibilities of the Vice President of Nuclear Engineering and Quality Assurance have been split between the Director of Power Generation Services taking over Quality Assurance and Reliability, offsite, and the Vice President of Nuclear Power taking over Nuclear Engineering, onsite.

The Nuclear Power Generation Department now includes an Operations Manager and staff instead of a Chief Operations Engineer, elimination of the position of Operations Superintendent and the addition of a new Planning and Project Manager and staff. The Operations Manager is responsible for managing the activities of the Senior Watch Supervisors, the Support Facilities Supervisors, the rotating shifts and has additional personnel who assist in the day-to-day management of Operations. The Projects and Planning Manager and staff organization includes the transfer of the Major Projects Manager and staff from the Technical Support Department to the Nuclear Power Generation Department.

All fire protection responsibilities of the Chief Technical Engineer of the Technical Support Department have been transferred to the newly created

position of Fire Protection, Safety and Security, and transfers them from the Security Administrator.

The proposed revision includes a change in the following titles:

Maintenance Engineer to Maintenance Manager, Chief Technical Engineer to Chief Plant Engineer, Radweste General Supervisor to Radwaste Manager, Test and Perfomance Engineer to Test and Performance Manager, Nuclear Training Director to Nuclear Training Manager, and Emergency Planning Director to Emergency Planning Manager. The Nuclear Environmental Monitoring Engineer and staff has become a functional responsibility and has been absorbed by the Manager of Emergency Planning.

The Technical Support Department has been merged with the Nuclear Engineering Department resulting in the addition of the Safety Assessment and Nuclear Analysis Managers to the Technical Support Department .

Basis for proposed no sionificant hazards consideration determination: The Commission has provided standards for determining whether a significant hr.zards consideration exists (10 CFR 50.92(c)). A proposed arrendment to an operating license for a facility involves no significant hazards consideratfpD jf.pDereliODs gf,the_fqpility.ip,ac,cpt ange,wjth,,thg,p,ropgged d

amendment would not: (1) involve a significant increase in the probability or consequences of an accident from any accident previously evaluated (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The licensee provided the following discussion regarding the above criteria. The proposed change does not involve a significant hazards consideration because operating of Indian Point Units 1 and 2 in accordance with this change would not:

l l

d (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. This is an administrative change which reflects a recent facility reorganization with consolidation and realignment of responsibilities. Therefore, this change cannot increase in the probability or consequences of an accident.

(2) Create the possibility of a new or different kind of accident from any previously analyzed. It has been determined that a new or different kind of accident will not be possible due to this change. An administrative change such as a facility re-organization does not

- create the possibility of a new or different kind of accident.

(3) Involve a significant reduction in a margin of safety. The administrative change as described above in the safety assessment discussion has been deternined not to impact the preservation or reduction of any margin of safety.

The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis.

Accordingly, the Commission proposes to determine that the proposed changes

( to the Technical Specification involve no significant hazards consideration.

Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York, 10610.

Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, l New York, New York 10003 NRC Project Director: Steven A. Varga i

l Consolidated Edison Company of New York, Docket No. 50-247, Indian Point Nuclear Generating Unit No. 2. Westchester County, New York j Date of amendment request: December 8, 1986 i

Description of amendment request: The proposed amendment would modify the Indian Point 2 Technical Specifications using provisions regarding the limiting conditions of operation for inoperable main steam safety relief valves contained in the Standard Technical Specifications for Westinghouse Pressurized Water Reactors, NUREG-0452, Revision 4.

Currently Technical Specification 3.4.A(1) requires that a minimum of twenty ASME code approved main steam safety valves (MSSV's) be operable when the reactor coolant system is above 350 F. If this requirement cannot be met within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the Technical Specifications require the reactor to be placed in the hot shutdown condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and subsequently cooled below 350 using normal operating procedures. The proposed change would allow up to three inoperable main steam safety valves on any steam generator on the basis of the reduction of secondary system steam flow and thermal power required by reduced reactor trip settings of .

the Power Range Neutron Flux Channels.

Basis for proposed no sianificant hazards consideration determination: The Commission has provided standards for determining whether a significant

! hazardsconsiderationexists[10CFR50.92(c)]. A proposed amendment to an operating license for a facility involves no significant hazards j consideration if operation of the facility in accordance with the proposed I -

amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in margin of safety.

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The licensee provided the following discussion with regard to the above three criteria. The proposed change to the limiting condition of operation for the main steam safety valves would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated, since the specified limits on power (per Table 3.4-1) with one or more inoperable main steam safety valves will not increase the consequences of a full loss of load accident and will ensure meeting the required heat removal capability at the reduced power level.

- (2) Create the possibility for a new or different kind of accident from any previously evaluated, since the proposed change only involves a reduction in power in order to compensate for any reduction in steam relieving capability with inoperable main steam safety valves. This is no different than operating at full power with 20 MSSVs.

(3) Involve a significant reduction in the margin of safety, since the power reduction compensates for any reduction in heat removal capability at full power caused by the inoperability of one, two or three main steam safety valves on any operating steam generator, thereby ensuring the same margin of safety.

The staff has reviewed the licensee's no significant hazards consideration determination and agrees with the licensee's analysis.

Accordingly, the Commission proposes to determine that the proposed changes to the Technical Specifications involve no significant hazards consideration.

Local Public Document Room location: White Plains Public Library, 100 Martine Avenue, White Plains, New York, 10610.

_ Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, New York, New York 10003 NRC Project Dire,ctor: Steven A. Varga Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie Plant, Unit No. 2, St. Lucie County, Florida Date of amendment request: December 2, 1986 Description of amendment request: The proposed amendment would revise the steam generator low water level trip setpoints specified in Table 2.2-1 and Table 3.3-4 of the Technical Specifications (TS) and the associated Bases'.

Specifically, the Reactor Protection System (RPS) trip setpoint would be reduced from 39.5% narrow range (NR) to 20.5% NR and the Engineered Safety Features Actuation System (ESFAS) trip setpoint would be reduced from 20.6%

NR to 19.0% NR. Similarly, the allowable values for the RPS and ESFAS trip setpoints would be reduced from 39.1% NR to 19.5% NR and from 20.0% NR to 18.0% NR, respectively. The proposed reduction of the RPS and ESFAS trip 1

! setpoints is expected to decrease the likelihood of an unnecessary reactor trip or Auxiliary Feedwater Actuation System (AFAS) initiation on low steam generator level.

The purpose of the steam generator low water level reactor trip is to provide protection against a loss of normal feedwater flow incident. The reactor trip setpoint should provide allowance that there will be sufficient water inventory in the steam generators at the time of the trip to provide a sufficient rargin before emergency feedwater is required. Automatic actuation of the AFAS is initiated when several parameters, including the l steam generator water level, reach the ESFAS trip values.

~~

Basis for proposed no significant hazards consideration detennination: The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR Part 50.92(c). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. The proposed changes will not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

The current RPS trip setpoint corresponds to a Cycle 1 analysis assumption of steam generator low level RPS trip setpoint of 30% NR for the most limiting event, Loss of Feedwater, and 5% NR setpoint for all other accidents. The current RPS setpoint also includes an uncertainty factor of 9.5% due to the uncertainties associated with the equipment, response times, and RPS cabinet, thereby arriving at the current setpoint value of 39.5%. The ESFAS setpoint value in the TS's reflects the 5% NR assumption value plus the respective uncertainty value.

However, the Safety Analyses presented for Cycles 2 and 3 incor-porated the assumed setpoint value of 5% NR (and the related instrument uncertainties) for the RPS and ESFAS. The proposed setpoints would incorporate tht 5% NR limit as used in the Cycle 2 and 3 analyses plus the respective instrument uncertainties, as derived using accepted methodology for instrument uncertainty calculations. Therefore, the 1

proposed setpoints do not impact the result of the Safety Analyses presented for Cycles 2 and 3, but nerely reflects a change in the analysis assumptions made for Cycle 1 versus Cycles 2 or 3. The Cycle 2 and 3 setpoint was set at 5% NR to ensure actuation prior to reaching 0% under accident conditions. The Cycle 2 Reload Safety Evaluation used the 5% NR setpoint and evaluated the results for all events which could be impacted, namely the Loss of Feedwater event, the Feedwater Line Break event, the Steam Line Break event, and the Inadvertent Opening of a Steam Generator Safety Valve or Atmospheric Dump Valve.

. The results of the analyses for the Cycle 2 Reload Safety Evaluation demonstrated that all key parameters were below the acceptance criteria. Therefore, on the basis that the proposed amendment has been addressed by the existing Analyses of Record, the proposed amendment would not involve a significant increase in the probability or

consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of accident from any accident previously evaluated.

The setpoints for the RPS trip and the AFAS have been established such that they ensure actuation of these functions before the instrumentation goes off scale. This starts with a minimum level of 5%

NR and includes appropriate errors for an inside containment accident such as a Steam Line Break. The difference between the setpoint and allowable values accounts for the instrumentation drift over the specified surveillance interval. As stated above, the proposed setpoints correspond to values for which safety analyses have previously been evaluated. Therefore, the proposed amendment would not

create the possibility of a new or different kind of accident from any accident pre'viously evaluated.

(3) Involve a significant reduction in the margin of safety.

The proposed setpoints incorporate the same narrow range setpoints as those incorporated in the Cycle 2 Reload Safety Evaluation for which the results of the analyses for the events presented demonstrated that all key parameters were below the acceptance criteria. Based on this evaluation, it can be concluded that the proposed amendment has been addressed by the existing Analyses of Record and therefore, would not involve a significant reduction in the margin of safety. -

Based on the above discussion, the staff proposes to determine that this proposed amendment involves no significant hazards consideration.

Local Public Document Room location: Indian River Junior College Library, 3209 Virginia Avenue, Fort Pierce, Florida 33450 Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzinger, 1615 L Street, N.W., Washington, D.C. 20036 NRC Project Director: Ashok C. Thadani g neral Public Utilities Nuclear Corporation. Docket No. 50-320. Three Mile Island Nuclear Station Unit 2 (TMI-2), Dauphin County, Pennsylvania l Date of amendment request: June 18, 1985 and July 31, 1985, as supplemented November 20, 1985, February 26, 1986 and May 20, 1986.

Description of amendment request: The proposed amendment would revise TMI-2 i

Operating License No. DPR-73 by modifying Appendix A Technical Specifications 3.7.4, 3.7.7, 3.7.10, 3.8.1, 3.8.2, 3.9.12.1, 3.9.12.2, 3/4.7.4, 3/4.7.7, 4.3, 4.7.4, 4.7.7, 4.8.1, 4.8.2, 4.9.12.1, and 4.9.12.2.

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Specifically, the proposed amendment would: (1) delete all operability requirements for the emergency diesel generators and supporting systems, (2) modify the operability requirements for the control room emergency air cleanup system so that a backup onsite AC power source (the diesel generators) is not required, and (3) modify the requirements for both the fuel handling building and auxiliary building air cleanup systems, by reducing the number of required operable exhaust fans from two to one for each system.

TMI-2 is currently maintained in a stable, long-term cold shutdown mode without a need for forced cooling of the core. Active safety systems pre-viously required to assure safe shutdown of the plant, such as the decay heat removal system and the high pressure safety injection system, have been deleted from the Technical Specifications by earlier license amendments.

Therefore, the need for emergency backup power to these systems no longer exists and the emergency diesel generators are not necessary for the

• maintenance of the facility in a safe, cold shutdown condition.

l Consequently, retaining the operability requirements for the diesel generators and supporting systems (e.g., cooling water, fire suppression) places an undue burden and expense on the licensee without significantly contributing to plant safety.

- Control room actions are not necessary to maintain the current safe shutdown of TMI-2, however, habitability of the TMI-2 control room in the event of an accident is assured through operation of the control room emergency air cleanup system. The licensee and the staff have evaluated a broad spectrum of accidents for TMI-2 in the current recovery mode. The only accident scenario having a potential to affect habitability of the l

, 36 & -

TMI-2 control room is a TMI-1 loss-of-coolant accident (LOCA) coincident with a loss of offsite power (LOOP) to TMI-2. In this extremely unlikely scenario, without backup power provided by the diesel generators, the control room emergency air cleanup system would be inoperable and evacuation of control room personnel could become necessary. The staff has previously determined that offsite power to TMI-2 can be restored within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and concurs in the licensee's assessment that temporary evacuation of the TMI-2 control room will not affect the maintenance of the plant in a safe shutdown condition. Therefore, the operability requirements for the TMI-2 control room emergency air cleanup system can be modified so that backup power fr'om the emergency diesel generators is not required, without impacting plant safety.

The proposed amendment would also reduce from two to one the number of operable exhaust fans required for both the fuel handling building air cleanup system and the auxiliary building air cleanup system. Regardless of the number of operable fans, both systems would still be required to j maintain a 1/8 inch water gauge negative pressure in the respective buildings to limit exfiltration in the event of radioactive releases. The staff has added a requirement to perform a daily check of pressure indication or the pressure alarm to assure early detection of system degradation. Thus, this change represents an improvement to safety since the negative pressure requirement remains the same and system surveillance is enhanced.

Basis for proposed no sionificant hazards consideration determination: The Commission has provided standards for determining whether a significant hazardsconsiderationexistsin10CFR50.92(c). A proposed amendment to an

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operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1)involveasignificantincreaseintheprobability J

or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of

safety.

TMI-2 is -in a long-term cold shutdown for accident recovery.

Short-lived fission products which make up the preponderance of the source ,

tenn for operating reactors have decayed to negligible levels. The decay P

heat produced by the core has now dropped to less than 10 kilowatts and forced cooling of the core has not been required or used since 1981.

j Consequently, in previous license amendments, the staff has determined that

! the potential accidents analyzed for TMI-2 in the current recovery mode are bounded in scope and severity by the range of accidents originally analyzed in the facility FSAR.

The proposed changes do not significantly increase the probability or consequences of an accident previously evaluated because no changes to l

current safety systems or setpoints are proposed. No active systems are required to maintain TMI-2 in its current safe shutdown condition and temporary evacuation of the control room for the duration of a loss of offsite power event, if necessary, will not impact plant safety. As the l requirements for those safety systems formerly provided with emergency power t

( by the diesel generators have been deleted by previous license amendments, l

deletion of the requirements for the diesel generators will r.ot have a J

significant impact on the prevention or mitigation of previously analyzed

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accidents. The proposed change to the operability requirements for the exhaust fans for the auxiliary and fuel handling building air cleanup systems will not have a significant impact on the probability or consequences of previously evaluated accidents because system performance requirements and required corrective actions would remain unchanged.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because no new modes of operation or new equipment are being introduced. Deletion of the license requirements for the diesel generators is consistent with those previous license amendments that deleted requirements for safety systems' fonnerly dependent on the diesel generators for emergency backup power. The functions of the auxiliary and fuel handling building air cleanup systems will remain unchanged.

The proposed changes do not involve a significant reduction in a margin of safety because, as mentioned previously, no active components are required to maintain the current safe shutdown of THI-2 and the diesel generators are not needed to provide emergency power to safety systems.

Also, the negative pressure required to be maintained in the auxiliary and fuel handling buildings remains unchanged.

Based on the above considerations, the Commission proposes to determine that the proposed changes do not involve a significant hazards consideration.

Local Public Document Room location: State Library of Pennsylvania Government Publications Section, Education Building, Coninonwealth and Walnut Streets, Harrisburg, Pennsylvania 17126.

a e , .-, --m--,,- - -,--

Attorney for licensee: Ernest L. Blake, Jr., Shaw, Pittman, Potts and Trowbridge, 2300 N Street, N. W., Washington, DC 20037 NRC Project Director: William D. Travers indiana and Michigan Electric Company, Docket No. 50-315, Donald C. Cook Nuclear Plant, Unit No. 1, Berrien County, Michigan Date of amendment request: January 9, 1987 Description of amendment reouest: The proposed amendment would change the Technical Specifications to allow certain tests or surveillances to be

. delayed until the end of the next refueling outage currently scheduled to begin during the second quarter of 1987. These tests include battery charger and battery service tests on the AB and CD batteries, response time testing of eauipment which actuates on an engineered safety feature (ESF) signal, reserve power transfer tests, tests of the monitoring system for ice condenser lower inlet doors, channel calibration of seismic monitoring instrumentation, power operated relief valve (PORV) calibration, inspection of containment penetration seals on doors and hatches, channel calibration of the safety valve position indicator acoustic monitor, and tests on the pressurizer heaters. The proposed amendment is similar to two previous requests which were approved by license amendment 100 dated December 20, 1986 (published 52 FR 1561).

Basis for proposed no significant hazards consideration determination: The Commission's standard for detennining wnether a significant hazards l consideration exists is as stated in 10 CFR 50.92. A proposed amendment to i

an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with a proposed I

amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

The changes proposed by the licensee will extend certain tests to the next refueling outage beginning about mid-April rather than shutdown the plant or place the plant or personnel in a less safe condition. The safety significance, therefore, is in extending the tests for one or two months and the confidence that the components or systems will still perform their

. intended function with the extension of time.

An additional consideration is the risk of unwarranted reactor trips and plant transients if the licensee attempts to perform some tests at power with new or revised procedures. ALARA concerns are also weighed against the short period of the extension. We have revised the licensee's submittal and sumarize below our agreement with the licensee's findings.

The AB and CD batteries are new and have passed all factory and post-installation tests and the subcomponent testing performed weekly and l quarterly. The test proposed for extension is the capacity test (without charger) which discharges the battery below the technical specification minimum operable level. The test is required to be done at shutdown. Based on the results of the tests on these new batteries to date and the batteries are not normally subject to operation that would limit or reduce their life, the short extension of time to perform the tests should have no discernible effect on the batteries. The extension will not significantly increase the probabilities or consequences of a previously analyzed accident nor will it create a new or different kind of accident. There should be no significant l

reduction in a margin of safety.

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The proposed amendment would extend the surveillance tests frequency for response testing of certain equipment which activates on an ESF signal.

Most of the tests must be done at shutdown and the remaining tests would require removing en entire train of safety equipment from operation. The surveillance history of the equipment has been good; the only problem with limit switch wiring was corrected. The systems have not been found unable to perform their safety function and the short extension should have little if any effect on response times. The reserve power transfer is somewhat like the response time testing in that the automatic transfer capability from the normal auxiliary source to the preferred reserve source and the' manual transfer to the alternate reserve source are measures of the capabilities of these systems which are not usually operated and subject to operational degradation. This testing also removes entire trains of safety equipment from service and is not considered feasible (or desirable) at power. The extension of time to perfonn the test is small and should have' little effect on system operation. The extension of time should not significantly increase the probabilities or consequences of a previously analyzed accident nor will it create the possibility of new accidents.

There is no significant reduction in any margin of safety.

The proposal to extend the test period for the ice condenser lower inlet door position monitoring system and the inspection of the containment personnel access door and equipment hatch seals is due to ALARA considerations while the plant is at power. The recent inspection history of these systems and the short extension period until shutdown supports the licensee's position that exposing the personnel needlessly to radiation is unwarranted. For the short period of the extension, these systems should L _ - - - -. _ -. .__ -_ .-

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continue to function with little or no safety consequences. We agree there should be no significant increase in the probabilities or consequences of previously analyzed accidents nor should the extension create any new or different accidents. There should be no significant reduction in any margin i

of safety.

The test of the pressurizer heaters is to demonstrate the capability to l

power the heaters from the emergency diesel generators. This test at power would require operation in an abnormal mode with emergency power on one train of safety related equipment. The successful tests of the individual I ,

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components of this system and the short period of the extension provides' I

l confidence in the operability or capability of the emergency diesel generatets to power the heaters if called upon to do so. The short

! extension of time should not significantly increase the probabilities or i

i consequences of any previously analyzed accident nor would it create any new or different kind of accident. There should be no significant reduction in f

any margin of safety.

The remaining proposed changes relate to extending the period for calibration of seismic monitoring channels, PORY block valve limit switches, and safety valve position indicator - acoustic monitor channels. The latter two items have monthly channel checks which have not indicated any problems with operation or adverse calibration effects. The seismic channels involved in the extension period are two (out of eight) strong motion triaxial accelographs which are located in the lower containment and inaccessible due to high radiation during operation. There is no reason to believe these instruments will not functina adequately for the short extension period. For all the calibrations, the short extension of the test

period will not significantly increase the probabilities or consequences of any previously analyzed accident nor will it create any new or different kind of accident. There should be no significant reduction in any margin of safety.

Based on the above, the staff proposes to determine that the requested changes do not involve a significant hazards consideration.

_L,ocal o Public Document Room location: Maude Preston Palenske Memorial Library, 500 Market Street, St. Joseph, Michigan 49085 Attorney for licensee: Gerald Charnoff, Esquire, Shaw, Pittman, Potts and

. Trowbridge, 2300 M Street, N.W., Washington, D.C. 20037

NRC Project Director

__ B.J. Youngblood Lona Is. land Lighting Company, Docket No. 50-322, Shoreham Nuclear Power S_tation. Suffolk County, New York Date of amendment request: January 21, 1987 Description of amendment request: The proposed amendment would revise the Technical Specifications regarding the water level setpoint for closure of theMainSteamIsolationValves(MSIVs). This setpoint would be changed from reactor low-low water level (Level 2) to reactor low-low-low water level (Level 1). Since the setpoint of water level instruments controlling the MSIVs also controls the main steam line drain valves the setpoint for closure of these valves is also changed by the proposed amendment.

i The proposed change would reduce the possibility of spurious MSIV closure due to variations in reactor water level. This, in turn, would inc'r ease plant availability, reduce challenges to the safety relief valves (with attendant suppression pool heatup), and simplify plant operation.

j

Basis for proposed no sianificant hazards consideration determination: In accordance with the Commission's Regulations in 10 CFR 50.92, the Commission has made a determination that the proposed amendment involves no significant hazards considerations. To make this determination the staff must establish that operation in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

All accidents or events previously evaluated in the Final Safety Analysis Report have been reevaluated using the proposed, revised water-level setpoint. These include abnormal operational transients, loss of coolant accidents and anticipated transients without scram (ATWS). In each case, reanalysis shows that the probability or consequence of the previously evaluated accident or event is not increased as a result of the proposed l

revision. Therefore, criterion (1) above is satisfied.

The reactor water-level setpoint is a safety feature which, through closure of the MSIVs, serves to mitigate the consequences of previously evaluated events. A change in this setpoint will not create the possibility of a new or different kind of accident from any previously evaluated since the change does not entail a hardware modification or any change in plant operating procedures, nor would the change in setpoint create a new accident sequence. Criterion (2) above is therefore satisfied.

The effects of the proposed setpoint change on operating parameters indicative of plant safety margin such as minimum critical power ratio and peak vessel pressure have been evaluated. Also evaluated were the effects

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on radiation release and shutdown capability during abnormal operational ,

transients, fuel cladding integrity during a loss of coolant accident, and the reactor response during an ATWS event. These analyses have indicated 4 that the proposed revision would not cause a reduction in safety margin.

Furthermore, since the proposed revision would decrease the possibility of spurious MSIV closure due to variations in reactor water level, challenges to the safety relief valves and attendant suppression pool heatup would be reduced, resulting in an improved safety margin. Therefore, criterion (3) 4 above is satisfied.

Because it has been established that plant operation in accordance with the proposed amendment would satisfy the three above stated criteria, the staff has, therefore, made a proposed determination that the proposed amend-ment involves no significant hazards consideration.

Local Public Document Room location: Shoreham-Wading River Public Library, Route 25A, Shoreham, New York 11786 Attorney for licensee: W. Taylor Reveley, III, Esq., Hunton and Williams, P. O. Box 1535, Richmond, Virginia 23212 NRC Project Director: Walter R. Butler i

i i Maine Yankee Atomic Power Company, Docket No. 50-309. Maine Yankee Atomic Power Station. Lincoln County, Maine Date of amendment request: --

January 12, 1987 and January 21, 1987 l Description of amendment request: The proposed amendment would modify the i

Technical Specifications (TS) to reflect the operating limits for the Cycle i

10 reload core as follows:

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

1. The steady state peak linear heat rate would be modified to reflect Cycle 10 specified acceptable fuel design limits for prevention of centerline melting.

2. The text would be modified to clarify that each fuel type has its own linear heat generation rate limit.

3. The figure in the TS which describes the Power Dependent Insertion Limit would be modified to reflect Cycle 10 control element assembly insertion limit.

4. The figure in the TS which describes the Allowable Unrodded Radial Peak versus Cycle A Average Burnup would be modified to reflect Cycle 10 -

i radial' peaking, j 5. The figure in the TS which describes the Allowable Power Level versus Increase In Total Radial Peak would be modified to reflect Cycle 10 1

power distributions and Reactor Protection System setpoints.

Basis for proposed no significant hazards consideration de_temination: The Commission has provided standards for detemining whether a significant hazardsconsiderationexists(10CFR50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from an accident previously evaluated; or (3) involve a significant reduction in margin of safety.

f i The licensee has evaluated the proposed changes to the Technical l Specifications for the Cycle 10 operation of the Maine Yankee plant against

! l l

1

the standards of 10 CFR 50.92 and has determined they do not involve a significant hazards consideration. The licensee's discussion follows.

These proposed changes do not:

1. Involve a significant increase in the probability or consequence of any accident previously evaluated. The Cy:le 10 refueling will involve the discharge of 73 fuel assemblies and insertion of 72 new assemblies and one previously irradiated assembly. The new fuel assemblies are fabricated by Combustion Engineering and are not significantly different from those previously used at Maine Yankee. In previous reload cores at Maine Yankee and other facilities, the NRC has found' the fuel design to be acceptable. The Control Element Assembly (CEA) pattern for Cycle 10 is identical to that used in Cycle 9. Also, the thermal, thermal-hydraulic, and physics characteristics for Cycle 10 are not significantly different from those of Cycle 9. Therefore.

these proposed changes which support the operation of Maine Yankee for Cycle 10 do not increase in probability of an accident previously evaluated.

The Cycle 10 design has been evaluated to demonstrate the accept-ability of events previously evaluated in the Maine Yankee FSAR. The acceptance criteria for the evaluation are identical to those which were employed for Cycle 9. Furthermore, the analytical methods used to demonstrate conformance of the Cycle 10 design are identical to those used in Cycle 9 except for the removal of the flux augmentation factor.

The removal of the flux augmentation factor was proposed by Maine YankeeAtomicPowerCompany(NYAPCo)byletterdatedMay 20, 1986 and approved by the NRC by letter dated September 15, 1986.

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s.

- 48 T The effects of Cycle 10 operation on the consequence of accidents previously evaluated in the Maine Yankee FSAR were analyzed in MYAPfo's submittal.

For these transients where the parameters for Cycle 10 are not bounded by previous safety analyses, a new or revised analysis was performed. They are:

1) Boron Dilution

2) Ejection

3) CEA Withdrawal

. 4) CEA Drop

5) Loss of Load Other transients that required a partial rearalysis or review included:

1) SeizedReactorCoolantPump(RCP) Rotor

2) Excess Load

3) Loss of Feedwater

4) Loss of Coolant Flow

5) Steam Line Rupture

6) Steam Generator Tube Rupture In each case, the reanalysis demonstrated that the applicable acceptance criteria for the accident or transient continue to be met.

For the remainino transients, the parameters were bounded by previous safety analyses and therefore are not adversely affected by the reload.

The evaluation of accidents previously analyzed in the FSAR has demonstrated that all applicable acceptance criteria continue to be met.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated. As indicated in Item 1, above, and the Cycle 10 Core Performance Analysis Report (CPAR), the reload core for Cycle 10 operation is similar in fuel design, CEA placement, and thermal, thermal-hydraulic, and physics characteristics to that of Cycle 9. It is concluded that Cycle 10 does not create the possibility of a new cr different kind of accident from any previously evaluated.

3. Involve a significant reduction in a margin of safety. As indicated in Item 1, above, and the Cycle 10 CPAR, the design of Cycle 10 is similar

- to Cycle 9. The methods used to analyze Cycle 10 are the same as were used for Cycle 9 or they have been previously approved by the NRC staff. Additionally, the acceptance criteria for Cycle 10 are the same as Cycle 9. It is demonstrated that these acceptance criteria continue to be met. It is therefore concluded that Cycle 10 operation does not involve any reduction in a margin of safety.

Based on the above discussion, the staff agrees with the licensee's evaluation and proposes to detemine that the proposed amendment would involve no significant hazards considerations.

Local Public Document Room locatjon,: o Wiscasset Public Library, High Street, Wiscasset, Maine Attorney for licensee: J. A. Ritscher, Esq., Ropes and Gray, 225 Franklin Street, Boston, Massachusetts 02210 NRC Project Director: Ashok C. Thadani Neb'raska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaha County,_ Nebraska Date of amendment request: November 24, 1986

.oe_scription of amendment request: The proposed amendment would revise the Appendix A Technical Specifications as follows:

1. Section 3.6.B.2 would be revised to state that the reactor coolant system water chemistry 71mits apply "when the reactor is pressurized (i.e. equal to or greater than 212 deg F) and during operation up to 10% of rated power." The present 3.6.B.2 applies " prior to startup and during operation of the reactor up to 10% of rated power, and during hot standby."

2. Section 3.6.B.2 would be reyfsed to change the reactor water conduc-

. tivity limit from less than 5 rnicromho/cm at 25 deg.C to equal to or 1ess than 2.0 micromho/cm at 25 deg.C.

3. Section 3.6.B.2 would be revised such that should a conductivity, chloride, or pH limit be out-of-limit for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval, the fccif f ty must be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and (or if already in hot shutdown) cold shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The present requirement states that the facility shall be shut down if pH is out-of-limit for a for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

4. Section 3.6.B.3 and Section 3.6.B.4 would be combined to include, in one specification, normal limits for conductivity, chloride, and pH; time limits for chloride, conductivity, and pH; and maximum limits for conductivity and chloride.

5. Section 3.6.B.3 would be revised to require, that in event pH is out-of-limit for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during one continuous time interval, the reactor be placed in startup/ hot standby within the next eight hours. The existing Section 3.6.B.4 requires, in such event, that the reactor be shut down if pH is out-of-limit for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6. Section 3.6.B.3 would be revised to specify that, should the normal chloride or conductivity limits be out-of-limit for more than 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> per year or 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during one continuous interval, the reactor shall be placed in startup/ hot standby within the next eight hours. The existing Section 3.6.B.4 requires the reactor to be placed in cold shutdown if the normal limits are exceeded for two weeks / year.

7. Section 3.6.B.3 would be revised to specify that should the maximum conductivity or chloride limits be exceeded, the reactor shall be placed in cold shutdown witain 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The existing Section 3.6.B.4 requires, in such case, the reactor be "placed in the cold shutdown' condition."

8. Section 3.6.B.4 would be revised to specify reactor coolant chemistry limits that apply "at all other times." This specification would replace the existing Section 3.6.B.5 which applies when the reactor is not pressurized. The new Section 3.6.B.4 would include limits on pH.

The existing Section 3.6.B.4 does not include pH limits. The new 3.6.B.4 would also require the licensee to perform an engineering evaluation in event the conductivity or pH limit is out-of-limit for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during one continuous interval or the chloride limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

9. A new Section 3.6.B.5 would be added. Section 3.6.B.5 would state that the provisions of Specification 1.0.J are not applicable.

Specification 1.0.J defines actions to be taken for circumstances in excess of the limiting conditions for operation.

10. Section 4.6.B.2 would be revised to indicate that, in the event the continuous conductivity monitor is inoperable, reactor coolant samples

shall be periodically analyzed; every four hours prior to startup, during operation, and during hot standby; and every 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> when not pressurized.

11. Section 4.6.B.3.d would be revised to delete a requirement for periodic sampling of the reactor coolant system when the continuous conductivity monitor is inoperable. That requirement would be included in Section 4.6.B.2 as described above.

12. Section 4.6.B.4 would be revised to include pH as one of the paraneters to be sampled every 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> when the reactor is not pressurized. At present, only chloride and conductivity are required to be sampled when depressurized.

13. Note 8 of Table 3.2.A would be revised to delete the statement that reactor water cleanup system high system temperature is a group 3 isolation signal, '

14. Section 6.4.2.H h,uld be revised to state that life retention would be required for " Records of current individual plant staff members showing qualifications and the completion of training." This would replace a requirement for life retention of " Records of individual plant staff members showing qualifications, training and retraining."

Basis for proposed no significant hazards consideration determination: The j Commission has provided guidance for the application of criteria for no significant hazards consideration determination by providing examples of amendments that are considered not likely to involve significant hazards l considerations (51FR7751).Theseexamplesinclude:

(i) A purely administrative change to technical specifications: for

! example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature.

(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications, e.g., a more stringent surveillance requirement.

(vi) A change which either may result in some increase to the probability or consequences of a previously-analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan, e.g., a change resulting from the application of a small refinement of a previously used calculational model or design method.

Change 1 eliminates the use of the vague term " prior to startup" replacing it with the more definitive tenn "when pressurized." Since pressurization occurs coincident with startup, there is no substantial -

change as to when the specification actually becomes applicable. The change serves only to provide clarification and is therefore within the scope of example (1).

Change 2 re. vises a limit to specify a more restrictive value. It is therefore within the scope of example (ii). -

Change 3 would allow additional time to correct an out-of-limit condition prior to commencing a required shutdown. This may in some way possibly decrease a safety margin. However, the additional time is consistent with the criteria of the Standard Review Plan which references NUREG-0123 " Standard Technical Specifications for BWRs" for the staff

. position on Technical Specifications for BWRs. The proposed change is consistent with Section 3.4.4.b of NUREG-0123 and is within the scope of example (vi).

Change 4 is a reformatting or editorial change and is within the scope ofexample(1).

Change 5 would allow additional time to correct an out-of-limit condition prior to commencing a required shutdown. This may in some way

7 . _.

possibly decrease a safety margin. However, the additional time is consistent with the criteria of the Standard Review Plan which references NUREG-0123 " Standard Technical Specifications for BWRs" for the staff position on Technical Specifications for BWRs. The proposed change is consistent with Section 3.4.4.a.2 of NUREG-0123 and is therefore within the scope of example (vi).

Change 6 would provide an additional restriction, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> continuous time limit, and is therefore within the scope of example (ii).

Change 7 adds a definitive time limit within which the licensee must

. accomplish a required action. This provides a clarification or additional restriction and is therefore within the scope of example (1).

Change 8 would provide clarification as to which chemistry limits apply during various reactor conditions, assures that all reactor conditions are covered by appropriate chemistry limits, and adds new restrictions. It is therefore within the scope of example (ii).

Change 9 would provide clarification by giving the new specifications precedence over the general, less restrictive requirements of Technical Specification 1.0.J. This change is within the scope of example (1).

Changes 10 and 11 are editorial changes. A requirement is being relocated from one section to another without being changed. They are therefore within the scope of example (1).

Change 12 is an additional restriction and is therefore within the scopeofexample(ii).

Change 13 corrects an error made in a previous amendment. In Amendment 62 the Staff approved the deletion of the reactor water cleanup system high temperature signal as being a Group 3 isolation signal. Note 8 of Table

3.2.A should have been changed as.part of Amendment 62. This change corrects the error and is therefore within the scope of example (1).

Change 14 is a clarification intended to ensure that permanent records are kept for plant staff individuals documenting the completion of training activities. The change clarifies that it is not the intent of the requirement that each such record thoroughly describe the subject matter covered during the training.

Since the application for amendment involves proposed changes that are encompassed by examples for which no significant hazards consideration .

~

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

Local Public Document Room location: Auburn Public Library,11815th Street, Auburn, Nebraska 68305 Attorney for licensee: Mr. G. D. Watson, Nebraska Public Pmver District, Post Office Box 499, Columbus, Nebraska 68601.

NRC Project Director: Daniel R. Muller Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, Millstone Nuclear Power Station, Unit No. 1, New London County, Connecticut Date of amendment request: December 22, 1986 Description of amendment request: The proposed amendment would change the operating license expiration date for the Millstone Nuclear Power Station, Unit No. 1 (Millstone 1), from May 19, 2006 to October 6, 2010, which would f be forty years from its October 7, 1970 date of issuance.

i Basis for proposed no sionificant hazards consideration determination: The l

l currently licensed term of operation for Millstone 1 is 40 years beginning

with issuance of the construction permit. Accounting for the more than 52 months required for plant construction results in an effective current operatinglicensetermof35yearsandsevenplus(7+) months. The NNECO (the licensee) application requests a 40-year operating license term for Millstone 1 in place of the currently shorter tem. The licensee's request for extension of the operating license is based primarily on the fact that a 40-year service life was considered during the design and construction of the plant.

The licensee has determined that the structures, systems and components

, are designed and maintained to perfom for the full 40 year operating ter'm i

i sndaresubhecttodetailedinspection,surveillanceandmaintenance requirements that assure detection of any abnormal degradation and provide for appropriate corrective action.

Only the reactor pressure vessel (RPV) is considered to be a non-replaceable component. However, the licensee has determined that the RPV,

(' consistent with its original design, will maintain its functional capability for the full 40-year operating term requested. The original design of the RPV and associated internals considered the effects of 40 years of operation within the cyclic limits identified in the Millstone 1 Final Safety Analysis j Report (FSAR). These cyclic limits equate to 40 years of operatiop at full power (2011 MW themal) with a plant capacity factor of 80%, including expected operational and themal transients. The cummulative capacity factor at this point in time is 68%. The design of the RPV meets the intent

of 10 CFR 50, Appendix A, General Design Criteria (GDC) 31, " Fracture Prevention of Reactor Coolant Pressure Boundary", and is therefore accept-able. In addition, the licensee has determined that total neutron fluence

over the 40 year term of the proposed operating license will be no greater than 16 percent of the design limit for neutron fluence in the RPV wall.

Limiting conditions of operation and surveillance requirements are identi- - '

fied in Technical Specifications 3.6 and 4.6. Included is the requirement for RPV test specimens (TS 4.6.B.5). In accordance with 10 CFR Part 50, Appendix H, the changes that appear in the mechanical and impact properties of the material contained in the surveillance capsules inside the reactor vessel are used to monitor the radiation-induced changes in the mechanical i

and impact properties of the RPV materials. This program provides a means of assuring that the cumulative effects of power operation are within des'ign allowances.

In accordance with 10 CFR 50.49, the licensee has established programs for electrical equipment requiring environmental cualification. These programs assure that electrical equipment important to safety is maintained in a qualified state at all times during plant operation, regardless of ths

_ length of operation. The programs include: aging analysis to determine i

qualified life, maintenance designed to preserve qualification, surveillance according to procedures, qualification of affected portions of plant modifi-cations and scheduled replacement of the equipment at the end of its quali-fled life. These programs provide assurance that equipment will perform as required, if called upon to mitigate design basis events, regardless of the

, term of the license.

Based upon the above, the licensee concluded that extension of the operating license for Millstone to allow a 40-year service life is consist-ent with the safety analysis in that all issues associated with plant aging have already been addressed. Since the proposed amendment involves no 6

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changes in the technical specifications or safety analyses, the licensee concludes that the proposed amendment would not: (1)involveanysignifi-l cant increase in the probability or consequences of an accident previously evaluated; or (ii) create a possibility of a new or different kind of accidentfromanyaccidentpreviouslyevaluated;or(iii)involveany reduction in the margin of safety.

The staff has reviewed the licensee's determination and concurs with its conclusions. Therefore, based on the above, the Commission proposes to determined that the proposed amendment, which would provide a 40 year operating life for the Millstone Nuclear Power Station, Unit No.1, involves no significant hazards considerations.

Local Public Document Room location: Waterford Public Library, 49 Rope Ferry Road, Waterford, Connecticut 06385.

Attorney for licensee: Gerald Garfield, Esquire, Day, Berry, & Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-3499.

NRC Project Director: Christopher I. Grimes.

Omaha Public Power District, Docket No. 50-285. Fort Calhoun Station, Unit No., 1. Washington Coun,ty, Nebraska Date of amendment request: January 7,1986 [ Sic] (1987)

Description of amendment request: Section 2.9.1(2)d of the Fort Calhoun Technical Specifications states that the hydrogen and oxygen monitors shall be u nitoring the in-service gas decay tank during the transfer of waste gases to the gas decay tank. Whenever the monitors are inoperable, transfer of waste gases to a gas decay tank may continue provided grab samples are taken and analyzed in accordance with the frequency specified in Section

t 2.9.1(2)d. The purpose of monitoring the hydrogen and oxygen concentrations is to assure that these concentrations will be maintained below the flammability limits of these gases to prevent a potential rapid release of the radioactive waste gas. The proposed amendment would modify Section 3.12.1(2)d(i) of the surveillance requirements for the hydrogen and oxygen monitor!ng system for the gas decay tanks to clarify that a daily channel check is required for this system when it is in service.

Basis for proposed no significant hazards consideration determination: The Commission has provided standards for determining whether a significant

. hazards consideration exists as stated in 10 CFR 50.92(c). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2)

create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. The proposed changes will not

l (1) Involve.a significant increase in the probability or consequences of an accident previously analyzed. Specifically, the proposed change does not increase the likelihood of an explosive concentration of hydrogen and oxygen in the waste gas decay tank. The proposed amendment would

( change the requirement for performing channel checks of the hydrogen and oxygen monitors from " daily" to " daily when in service." This i

chance, for example, would not make it necessary to perform a daily l ' channel check if the hydrogen and oxygen monitoring system was inoperable. As discussed above, whenever the hydrogen and oxygen l

i

~

monitors are inoperable, transfer of waste gases to a gas decay tank may continue provided grab samples are taken and analyzed in accordance withSection2.9.1(2)doftheTechnicalSpecifications. Therefore, this change does not affect the current requirements for monitoring hydrogen and oxygen concentration in the waste gas decay tanks. This change does not increase the probability or consequences of a previously evaluated accident.

(2) Create the possibility of a new of different kind of accident from any accident previously evaluated. Because the proposed change does not modify the present design, the possibility of a different type of accident other than that previously anelyzed has not been created.

(3) Involve a significant reduction in the marain of safety. Because the proposed change does not affect the requirement'to monitor the hydroacn and oxygen concentrations during the transfer of waste gases to the gas decay tank, the proposed change will not involve a significant reductinn in a margin of safety.

Based on the above discussion, the staff proposes to detennine that the l

proposed change does not involve a significant hazards consideration.

. Local Public Document Room location: W. Dale Clark Library, 215 South 15th Street Omaha, Nebraska 68102 Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1333 New Hampshire Avenue, N.W., Washington, D.C. 20036 l

NRC Project Direc g : Ashok C. Thadani Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna Steam Electric Station, Unit 1, Luzerne County, Pennsylvania D. ate of amendment req.ug: December 12, 1986 1

l l

Description of amendment request: The proposed amendment would revise the Susquehanna Steam Electric Station (SSES) Unit 1 Technical Specifications (TS) to incorporate changes to: (a) General Electric (GE) fuel and Exxon Nuclear Company (ENC) fuel MAPLHGR Limits; (b) MCPR operating limits reflecting the results of the XCOBRA-T analyses; (c) set the single loop operation (SLO) MAPLHGR multiplier for ENC fuel to 0.0 to preclude extended operation with one recirculation loop out-of-service, and (d) change bases to provide consistency between Unit 1 and Unit 2 Technical Specifications.

Basis for proposed no significant hazards consideration determination: The

. Commission has provided standards for determining whether a significant '

hazardsconsiderationexists(10CFR50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability

or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The staff has reviewed and concurs with the following basis provided by the licensee in its December 12, 1986 submittal for its 50.92 determination.

(a) MAPLHGR Limits (i) GE Fuel The current GE MAPLHGR limits are provided in the Technical Specifications to an average planar exposure of 33,069 MWD /MT.

However, the GE fuel bundles are projected to attain average planar exposures greater than 33,069 MWD /MT in both Cycle 3 and Cycle 4 operation.

GE performed a LOCA analysis to determine the MAPLHGR limits for an average planar exposure of 40,675 MWD /MT. The analysis was performed based on the same plant conditions and systems analysis that were used as a basis for the derivation of the current MAPLHGR limits defined in the FSAR. Using a MAPLHGR limit of 9.2 kw/ft at an average planar exposure of 40,675 MWD /MT, the Peak Cladding Temperature (PCT) is 1621 F and the local oxidation fraction is 0.3 percent. These values are not significantly different from the previous values (PCT range 1753 F to 1823*F and local oxidation 0.5 to 0.6 percent), when f . compared to 10 CFR 50.46 limits of 2200*F and 17 percent local oxidation rate.

In addition, the projected peak bundle average exposure at the End-of-Cycle (EOC) 3 and EOC 4 was calculated to assure the fuel does not exceed the limits of the fuel mechanical design analysis (NEDE-20944-P, Revision 1. " Licensing Topical Report: BWR-4 and BWR-5 Fuel Design," October 1976). The projected E0C3 and EOC4 peak bundle average exposure for the GE fuel is 26,000 MWD /ST (28,600 MWD /MT) which l is well within the design bundle average exposure limit of 30,000 .

MWD /ST.

(ii) ENC Fuel .

The current MAPLHGR limits for ENC fuel are provided to assure that the PCT resulting from the LOCA analysis is less than 2000*F and that during steady state operation, the fuel remains within the assumptions of the fuel mechanical design analyses. In the Unit 1 Cycle 3 (U1C3) reload analysis Technical Specification revision (Amendment 57 to license No. NPF-14, dated April 11,1986), a Linear o

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Heat Generation Rate (LHGR) Limit was added for the Exxon fuel to assure the fuel operates within the assumptions of the mechanical design analysis for steady state operation and anticipated operational occurrences. Therefore, the proposed MAPLHGRs are based only on the LOCA analysis.

The ENC LOCA Analysis showed that the LHGR limit currently in the i

l Technical Specifications assures that the fuel remains within the assumptions of the fuel mechanical design analysis and the 10 CFR 50.46 limits are not exceeded with the revised MAPLHGR limits.

(b) MCPR Operatino Limits During the review of the Unit 2 Cycle 2 reload analysis, the ENC method for calculating the MCPR operating limits was shown to be non-conservative for transients which involved recirculation pump trips (RPT). Therefore, administrative MCPR limits were implemented to assure that the Unit 1 core did not exceed the Safety Limit MCPR during

- a transient which resulted in a recirculation pump trip.

ENC has reanalyzed the UIC3 transients with an improved computer code, XCOBRA-T, to determine conservative MCPR operating limits. As

! shown in the XCOBRA-T documentation (SN-NF-84-105, Volume 1 and Revision 1 of Supplements 1 and 2, "XCOBRA-T: A computer Code for BWR

~

Transient Thermal-Hydraulic Core Analysis," May, 1985 and March, 1986),

XCOBRA-T has been benchmarked against power increase and flow decrease boiling transition tests. The results show that XCOBRA-T is conservative for predicting the onset of transition boiling. From the results of the XCOBRA-T analyses the transient delta CPR and the Rod Withdrawal Error delta CPR are added to the MCPR Safety Limit to obtain

~ ~~ ~~

- 64 : '

the MCPR operating limits shown in the new Technical Specification Figures 3.2.3-1 and 3.2.3-2. In addition, the Rod Block Monitor (RBM) setpoint Technical Specification can be changed since the MCPR operating limit is not decreased for a reduced setpoint. With these MCPR operating limits, 99.9 percent of the fuel rods are not expected to experience transition boiling during normal operation and anticipated operational occurrences. This result is the same as the previous calculated value of percent rods not expected to experience l transition boiling, i .

(c) MAPLHGR Multipliers for SLO i

The Unit 1 Technical Specifications were revised to allow extended i

operation with one recirculation loop out-of-service. GE performed i

single loop safety analyses and ENC provided justification that the GE operating limits were applicable to the ENC fuel. However, while implementing the SLO Technical Specification, it was found that no explicit analyses were available to support ENC's justification of the applicability of the GE operating limits. Therefore, administrative Ifmits were implemented which precluded extended single loop operation.

This Technical Specification change set the SL0s MAPLHGR limit for ENC i

l fuel to 0.0 which precludes extended operation with one recirculation pump out-of-service until ENC can perform the required analyses to determine operating limits for the ENC fuel.

(d) TS Bases Section 3/4 7.8 Bases were revised to provide a direct reference to the transient analyses. This change was made to provide a consistent description of the Bases between the Unit I and Unit 2 f Technical Specifications.

l l

l l

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Based on the above licensee evaluation, the staff finds that:

^

(1) The proposed changes do not involve a significant increase in the prob-ability or the consequences of an accident previously evaluated because:

(a) changes in the MAPLHGR for both GE and ENC fuels, when included in the LOCA analysis, result in PCT and local oxidation fractions well within the 10 CFR 50.46 limits; (b) with proposed MCPR operating limits, the percentage of the fuel rods not expected to experience transition boiling during normal operation and anticipated operational occurrences is not changed; (c) the TS for SLO has been changed to preclude extended operation in SLO mode until analyses can be performed to detemine appropriate operating limits; and

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(d) changes to bases do not affect the probabilities and consequences of previously evaluated accidents.

(2) The proposed changes do not create the possibility of a new or

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different kind of accident because:

! (a) the proposed MAPLHGR changes do not create any new events; (b) the use of improved calculation methods do not create the possibility of any new events; (c) the specification has been changed to preclude extended operation in SLO mode to prevent operation without appropriate operatirg ,

limits; and j (d) the changes in the bases are administrative in nature and do not create any new events.

(3) The proposed changes do not involve a significant reduction in a margin of safety because:

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(a) the LOCA analyses show that MAPLHGR changes will not result in a significant reduction in PCT and local oxidation safety margin; (b) the use of the staff approved improved methodologies to determine MCPR limits were used to show that the changes will not result in i a significant reduction in a margin of safety; (c) precluding extended operation in SLO mode is not expected to result in significant reduction in a margin of safety; and (d) the changes in the bases are administrative in nature and do not affect any margin of safety.

. Based on the above considerations, the Commission proposes to determine that the proposed changes do not involve a significant hazards consideration.

Mc_al Public Document Roo.n location: Osterhout Free Library, Reference Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701.

! Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts and l

Trowbridge, 1800 M Street N. W., Washinoton, D.C. 20036 NRC Project Director: Elinor G. Adensam Pennsylvania Power and Lioht Company, Docket No. 50-387 and 50-388, Susquehanna Steam Electric Station,. Unit 1 and 2. Luzerne County, Pennsylvania Date of amendments request: February 10, 1986, as supplemented March 4, June 24, and August 29, October 1, 1986, and January 21, 1987.

Description of amendments request: The licensee's request for the proposed amendments to revise the Susquehanna Steam Electric Station (SSES) Unit 1 and Unit 2 Technical Specifications (TS) was published in a bi-weekly FEDERAL REGISTEP, notice dated September 24,1986,(51F.R.33956). As a l-1

result of the staff review of the licensee's request and subsequent staff discussions with the licensee, the licensee submitted a letter dated January 21, 1987, revising the request for TS changes.

In its January 21, 1987 submittal, the licensee requested (a) to add two tests of diesel startup using simulated signals for LOCA and loss of offsite power (LOOP), (b) a one time exception from testing requirements to verify diesel generator E's capability to synchronize with the offsite power source while the generator is loaded with emergency loads, and a I

substitution of the excepted test by an equivalent startup test, and (c)

. that it be permitted to change the earlier rating for the E diesel genera' tor from 5000 kw to 4000 kw for the purposes of testing.

B, asis for proposed no sianificant hazards consideration determination: The Commission has provided standards for determining whether a significant hazardsconsiderationexists(10CFR50.92(c)). A proposed amendment to an operating license for a facility involves no significant hazards I

consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident l previously evaluated; or (3) involve a significant reduction in a margin of safety.

The staff has reviewed the licensee's request and concurs with the licensee's basis provided in January 21, 1987 submittal.

Based on its evaluation, the staff finds that:

(1) the proposed change to the ifcensee's February 10, 1986 request, as revised does not involve a significant increase in the probability or

consequences of an accident previously evaluated because (a) the added tests are the same as those for the existing diesels, (b) one time exception from testing requirement to verify diesel generation E's capability to synchronize with the offsite power source while the generator is loaded with emergency loads will be substituted by an equivalent startup test which demonstrates that the intent of the test will be satisfied, and (c) derating of the E diesel generator from 5000 kw to 4000 kw still provides the diesel generator capability identical to each of the four existing diesel generators; (2) the proposed revision does not create a possibility of a new or different kind of accident because (a) the added tests are the same as those for the existing diesels, (b) one time exception from testing requirement to verify diesel generator E's capability to synchronize with the offsite power source while the generator is loaded with emergency loads.will be substituted by equivalent startup tests, and (c) derating of the E diesel generator from 5000 kw to 4000 kw will still provide rated power equal to any of the existing diesel generators; and (3) the proposed revision does not involve a significant reduction in a margin of safety because (a) the added tests are the same as those for the existing diesels, (b) one time exception from a diesel generator E' synchronization test is substituted by an equivalent startup test, and (c)deratingofthedieselgeneratorEfrom5000kwto4000kwwill still provide diesel generator E capability equivalent to any of the four existing diesel generators.

I Based on the above considerations, the Commission proposes to determine that the proposed changes, as revised, do not involve a significant hazards consideration.

Local Public Document Room location: Osterhout Free Library, Reference Department, 71 South Franklin Street, Wilkes-Barre, Pennsylvania 18701.

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts and Trowbridge, 1800 M Street N. W., Washington, D.C. 20036 NRC Project Director: Elinor G. Adensam Philadelphia Elec.tric Company, Public Service Electric and Gas Company, Delmarva Power and Light Company, and Atlantic City Electriq Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania Date of application for amendments: October 24, 1986 Description of amendment request: The proposed amendments would modify the Peach Bottom Technical Specifications (TS) by changing the currently specified range for the drywell temperature indicator and recorder. The 1

i changes are necessary to reflect the new equipment to be installed during the next refueling outages to meet the NRC's requirements for wide-range accident monitoring instrumentation as specified in Regulatory Guide 1.97 (InstrumentationforLightWater-CooledNuclearPowerPlantstoAssessPlant l

and Environs During and Following an Accident).

Basis for proposed no sianificant hazards consideration determination- The Commission has made a proposed determination that the amendment request involves no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility in l

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accordance with the proposed amendments would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any previously evaluated; or (3) involve a significant reduction in a margin of safety. A discussion of these standards as they relate to the proposed amendments follows:

Currently, drywell temperature is displayed in the control room on a recorder which has a range of 0 to plus 240*F. This recorder will be replaced with new recorder which has a range of 40 to 440 F. The

. temperature indicator, which monitors drywell temperature and feeds its input to the recorder, has a current range of minus 150 to plus 300 F. This temperature indicator will be modified to have a range of minus 300*F to plus 750 F. As a result of the TMI-2 accident, NRC has required that all plants have " wide-range" accident monitoring instrumentation so that during both normal and postulated accident conditions it would be highly unlikely that offscale readings will occur. The present TS specify not only the operability requirements on surveillance instrumentation such as the drywell temperature indicator and recorder but also the range over which this l

l instrumentation must be periodically calibrated.

In the same table (Table 3.2.F) listing the above drywell temperature instrumentation, there is an error in how the range for the suppression chamber water level is specified. The licensee has proposed a change to the TS to clarify this parameter. This is an administrative type change which does not change the actual range in which the water must be maintained and has no safety significance.

The proposed changes to the TS on the drywell temperature instrumentation are necessary to reflect the new modified instrumentation l

being installed to meet an NRC requirement. The changes do not involve a significant increase in the probability en consequences of any previously evaluated accident, create the possibility of a new or different kind of accident or involve a significant reduction is a margin of safety. In fact, in imposing the requirements for wide-range accident monitoring

• instrumentation, the NRC concluded that this would enhance the margin of safety in the event of an accident.

Accordingly, the staff proposes to determine that the proposed changes to the Technical Specifications do not involve a significant hazards

. consideration.

Local Public Document Room location: Government Publications Section, State Library of Pennsylvania, Education Building, Convronwealth and Walnut Streets, Harrisburg, Pennsylvania 17126 Attor.ney for Licensee: Troy B. Conner, Jr.,1747 Pennsylvania Avenue, N.W.

Washington, D.C. 20006 NRC Project Director: Daniel R. Muller

Portland General Electric Company, et al., Docket No. 50-344, Trojan Nuclear Plant, Columbia County, Oreoon Date of amendment request

October 31, 1986 as supplemented December 8,

, 1986.

Description of amendment request: This amendment proposes to revise the Trojan Technical Specifications, Section 3.1.3.4, " Shutdown Rod Insertion Limit," and 3.1.3.5, " Control Rod Insertion Limit," to redefine fully withdrawn rods as greater than 225 steps instead of the currently specified 228 steps. This change is proposed to minimize the likelihood of localized

fretting wear which results from rubbing contact between the rod and its alignment card caused by flow induced vibration. Concentrated fretting wear which can occur at 228 steps may be eliminated by subjecting a different portion of the rod to wear against the alignment card, thereby significantly extending the rod life.

Basis for proposed no significant hazards consideration determination: The Commission has provided guidance concerning the application of the standards j for determining whether license amendments involve no significant hazards considerations by providing certain examples (March 6, 1986, 51 FR 7751).

4 One example of an amendment that is considered not likely to involve significant hazards considerations is Example (vil a change which either may result in some increase to the probability or consequences of a previously analyzed accident or may. reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan.

Rod insertion limits are imposed during operation to ensure that power distribution limits are not exceeded, and that minimum shutdown margin is maintained. Since neither the power distribution limit or minimum shutdown I.

margi,n is changed, the proposed amendment does not result in an increase in the probability or consequences of an accident previously evaluated, nor does it create the possibility of a new or different kind of accident from any accident previously evaluated.

Redefining " fully withdrawn rods" to be greater than 225 steps represents approximately one inch of rod insertion. Because of the low rod worth in the top region of the core, the represented insertion has a negligible effect on power distribution. The available excess shutdown 4

margin of approximately 400 pcm margin would be decreased by a negligible amount (4 to 6 pcm), and would not encroach the minimum shutdown margin requirement. As such, the proposed change does not significantly reduce a margin of safety. Since the proposed change does not alter the existina ,

power distribution limit or minimum shutdown margin, the change is acceptable with respect to the criteria specified by SRP Section 4.3.

The staff has reviewed the licensee's no significant hazards analysis and concludes that the proposed change is within the scope of the Connission's cited example. Thus, the staff proposes to determine that the

. requested change does not involve a significant hazards consideration.

, Local Public Document Room location: Multnomah County Library, 801 S. W.

10th Avenue, Portland, Oregon Attorrey for licensee: J. W. Durham, Senior Vice President, Portland General Electric Company,121 S. W. Salmon Street, Portland, Oregon 97204 NRC Project Director: Steven A. Varga Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna Nuclear Power Plant, Wayne County, New York Date of amendment request: December 17, 1986.

Description of amendment request: The proposed changes to the technical specifications requested by Rochester Gas and Electric Corporation (the licensee) involve administrative changes in the licensee's Engineering organization and the Electric Transmission and Distribution organization.

The Engineering organization changes involve the elimination of the position of Assistant Chief Engineer and the creation of a Research and Science Group. The function that reported to the Assistant Chief Engineer would

under the proposed change report to the Chief Engineer directly or through the Research and Science Director or Manager of Divisional Services. The proposed changes associated with the Electric Transmission and Distribution organization involve duties currently assigned to Engineering and Electric Meter and Laboratory which would be shared under the proposed changes with the Electric Substations department. Another proposed change in the Ginna organization consists of the Ginna Quality Control Inspection Supervisor

! reporting directly to the Nuclear Assurance Manager in lieu of the Quality Control Engineer. Specifically, the proposed changes modify four figures in

. the Administrative Section of the technical specifications that describe the management organization, the Ginna Station organization, the Health Physics i

and Chemistry Group and the Maintenance and Nuclear Assurance Groups.

Basis for proposed no significant hazards consideration determination: The Commission has provided guidance for the application of standards for

determining if a no significant hazards consideration exists by providing

examples of amendments that are considered not likely to involve significant hazardsconsideration(51FR7751). One of these examples (1) is a purely administrative change to technical specifications

for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature. The proposed changes consist primarily of changes in nomenclature involving administrative functions that are normally under an orcanizational position and would be transferred to a new Research and Science Group. Other changes involving the sharing of duties between two organizations do not involve a reduction in the level of plant safety. These changes are administrative in nature and do not impact in any way the independent review of safety issues nor do they reduce the

level of plant safety. In addition, the proposed changes do not relax the intent of the TS; involve a significant increase in the probability or consequences of an accident previously evaluated; or create the possibility of a new or different kind of accident from any accident previously evaluated. As a consequence of the above, the staff has made a proposed determination that the application for amendment involves no significant hazards consideration.

Local Public Document Room location: Rochester Public Library, 115 South Avenue, Rochester, New York 14610.

. Attorney for licensee: Harry Voigt, LeBoeuf, Lamb, Leiby and McRae, Suite 1100, 1333 New Hampshire, N.W., Washington, D.C. 20036 NRC Project Director: George E. Lear, Director.

l Southern California Edison Company, et al., Docket No. 50-206, San Onofre Generating Station, Unit No.1_, San Dieoo County, California Date of amendment request: November 12, 1986.

Description of amendment request: The proposed change to the Technical Specifications would lower the pressurizer high level reactor trip setpoint l

from 70% to 50% in order to achieve consistency with the recently revised safety analysis for the loss of normal feedwater (LONF) accident. The previous LONF safety analysis relied upon the steamflow/feedflow mismatch reactor trip; however, recent operational experience has revealed that this reactor trip function is not single-failure proof. Thus, the safety analysis has been revised to rely upon the high pressurizer level trip signal, which is single-failure proof; however, the setpoint for this trip has been lowered from 70% to 50% to ensure that previously established acceptance criteria are n'et.

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' Basis for proposed no sionificant hazards consideration determination: The Commission has provided standards (10 CFR 50.92(c)) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license for a facility involves no significant hazards -

consideration if operation of the facility in accordance with the proposed I amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the i

possibility of a new or different kind of accident from any accident

! previouslyevaluated;or(3)involveasignificantreductioninamarginof

- safety. The licensee's November 12, 1986 submittal states the following with respect to these three factors:

This proposed change does not significantly increase the probability or consequences of an accident previously evaluated. As discussed in Attachment 1, the reanalysis of the events affected by the failure of PT-459, demonstrates that for the LONF event reducing the Pressurizer i High Level Reactor Trip setpoint to 50% will result in a system

response which is within the previously established acceptance l criteria. This change does not affect the system response to a '

l Feedwater Line Break event. In addition, the change does not impact l the probability of occurrence of the affected transients since the reduction of the trip setpoint is not related to any of the event

initiators such as loss of offsite power, feedwater system equipment failures and pipe break.

The proposed change does not creata the possibility of a new or different kind of accident. The change will only lower the initiation setpoint of a safety trip system to provide for earlier initiation of the trip during a transient. The change is a move in the conservative direction with respect to this safety function.

i l Operation of the facility in accordance with this proposed change does not involve a significant reduction in a margin of safety. The '

reduction in the Pressurizer High Level Trip setpoint to 50% results in a transient system response for the affected event (LONF) which is I within the previously established acceptance criteria for the event and therefore equivalent to the previously analyzed margin of safety.

The staff has reviewed the licensee's determination and agrees with

h x-their analysis. Therefore, the staff proposes to determine that the proposed amendment does not involve a significant hazards consideration.

Local Public Document Room location: Main Library, University of California, P. O. Box 19559, Irvine, California 92713.

Attorney for licensee: Charles R. Kocher, Assistant General Counsel, James Be'oletto, Esquire, Southern Edison Company, P.O. Box 800, Rosemead, California 91770 NRC Project Director: George E. Lear

. Southern California Edison Company, et al, Docket Nos. 50-361 and 50-362, San Onofre Nuclear Generatina Station, Units 2 and 3, San Diego County 2 California Dates of amendment request: March 17, and June 13, and September 30, 1986 (Reference PCN-192)

Description of amendment request: The proposed change would revise l Technical Specification (T.S.) 3/4.8.1.1 " Electrical Power Systems, AC Sources" to reduce the required number of fast cold start surveillance tests of the emergency diesel generators. The prcposed change also would modify diesel fuel oil testing requirements to more accurately determine the quality of the diesel fuel oil and its ability to power the diesel generators. In addition, the proposed change would clarify the requirements for offsite power for each engineered safety features (E5F) train.

The purpose of Technical Specification 3/4.8.1.1 is to ensure that sufficient power will be available to supply the safety related equipment req'u ired for (1) the safe shutdown of the facility and (2) the mitigation t

n and control of accident conditions within the facility. The proposed change b to this technical specification consists of the following parts:

(a) Technical Specification 3.8.1.1.a currently requires two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system. The proposed change would require two physically independent circuits between the offsite transmission network and each Class IE 4kV bus. This change would also require that Action Statement "a" associated with T.S. 3.8.1.1 be entered on a bus-by-bus' basis rather than a circuit-by-circuit basis. The proposed i

- change would also modify Technical Specification 4.8.1.1.1 to be t.onsistent with T.S. 3.8.1.1.a. This change would remove the existing '

requirement for cold start testing of a diesel generator on an unaffected Class IE 4kV bus, thereby reducing the number of fast cold start tests of the diesel generators.

(b) For Modes 1, 2, 3 and 4, existing Technical Specification 3.8.1.1 requires that if a diesel generator has become inoperable, it be restored to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant be brought to cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. However, the existing technical specification does not provide any limitation on the frequency of diesel inoperability or the total number of days lost per year due to inoperability.

The proposed change would provide a limit of 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> on the combined out-of-service time available to the two diesel generators in one calendar year. Should additional time be needed in a specific situation, the proposed change would require that the NRC be notified of the circumstances. Having thus established a minimum availability 1

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goal, this proposed change would increase the existing 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> individual out-of-service limit to 7 days (168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />), thereby pemitting greater flexibility in handling diesel generator malfunction and/or servicing needs without recourse to plant shutdown. Both of these proposed limits would be applicable in Modes 1, 2, 3 and 4 only.

(c) Technical Specification 3.8.1.1 Action Statements (a) and (b) require the diesel generators to be demonstrated operable by fast cold start testing within one hour and once per eight hours thereafter when either one offsite AC circuit and/or diesel generator is inoperable. The

. proposed change would reduce the number of diesel generator fast col'd start tests by requiring only one test of the diesel ger.erators within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when one diesel generator or one offsite AC circuit is inoperable.

(d) Technical Specification 3.8.1.1 Action Statement (d) requires that with two offsite AC sources inoperable, two diesel generators must be demonstrated operable by fast cold start testing within one hour and once per eight hours thereafter. The prcposed change would modify this l

action statement by requiring the two diesel generators to be verified operable by start testing within eight hours unless the diesels are already operating. This would reduce the number of fast cold start tests of the diesel generators.

(e) Technical Specification 4.8.1.1.2.a.4 currently requires the diesel  !

generators to be verified operable by fast cold start testing in accordance with the frequency specified in Table 4.8.1. This testing l

requires the diesel generators to start from ambient conditions and accelerate to 900 rpm in less than or equal to 10 seconds, i

Additionally, the generator voltage and frequency are required to be at-4,360 1 436 volts and 60 1.2 Hz within 10 seconds. The proposed change would require a fast cold start from ambient conditions only once per 184 days. For all other surveillance starts, the proposed change would allow the diesel generators to be started in accordance with the manufacturers recommendations regarding engine prelube and warmup procedures and allow the diesel generator to be gradually loaded. The proposed change would also specify that the diesel

! generators are to be started for the purpose of surveillance testing by

- the following signals .only: (1) manual,(2)simulatedlossofoffsite power by itself and (3) simulated loss of offsite power in conjunction with an ESF actuation test signal.

(f) Technical Specification Table 4.8.1 prescribes the test frequency for 4

diesel generators based on the number of failures in the last 100 valid demands. The proposed change would revise the diesel generator test base from the last 100 valid demands to the last 20 valid demands. The proposed change would also delete the last two tiers of test frequency, reducing the most frequent diesel generator testing from 3 days to 7 days.

(g) Technical Specification 4.8.1.1.2.c requires that diesel fuel oil be tested for water and sediment content, viscosity, and insolubles once every 92 days as well as from new fuel prior to addition to the fuel storage tanks. The proposed change would upgrade the testing methods to be consistent with current industry practice and would replace the l current test for insolubles with a more accurate and effective test.

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Basis for Proposed No Significant Hazards Consideration Determination: The NRC staff proposes to determine that the proposed change does not involve a

significant hazards consideration because, as required by the criteria of 10 CFR 50.92(c), operation of the facility in accordance with the proposed amendment would not: (1)Involveasignificantincreaseintheprobability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in the margin of safety.

1. The NRC staff proposes to determine that operation of the facility in accordance with these proposed changes does not involve a significant

- increase in the probability or consequences of any accident previously evaluated for the reasons given below.

(a) Surveillance Starts As noted above, the proposed change affects only the surveillance a

requirements pertaining to the diesels and not those pertaining to the I offsite circuits. Upon loss of required AC power, only one 1

l surveillance start is deemed necessary to confirm the operability of a diesel generator. By eliminating the repeat diesel surveillance starts at less than or equal to eight hour intervals, the proposed change would prevent premature diesel engine degradation and contribute to enhanced plant safety over the long tenn. Whereas the existing l technical specifications require demonstration of diesel generator operability within one hour of the initial power loss, the proposed change permits a delay of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after losing one source and 8 I hours after losing two sources. These new time limits conform to Generic Letter 84-15 and are consistent with the goal of minimizing l wear on the diesel engine parts. These limits would pennit the l

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inoperable power source (s) to be repaired and restored, if possible, while avoiding an unscheduled diesel start. Although the new limits are a relaxation from the existing surveillance requirements, it is not considered a significant relaxation, in light of the requirement to test the offsite circuits within I hour of the initial power loss and every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter for the duration of the loss. If the inoperable power source cannot be restored to service within the specified time interval, the technical specifications would require plant shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

. By enhancing both long tenn diesel reliability and imediate p1 ant safety requirements under different power loss situations, a decrease in the probability and/or consequences of an accident will result from the proposed change.

(b) Out of Service Time Limits

Increasing the individual out of service limit from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 7 days does not involve a significant increase in the probability or ,

consequences of any accident previously evaluated, considering that:

(1) The safety requirement to be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> if i the out-of-service limit has been exceeded and the inoperable i

power source remains inoperable is unchanged.

(2) The annual limit will insure that the actual out-of-service time is in all cases within reasonable limits and unnecessary diesel l

, out-of-service time is avoided.

l (3) In the history of San Onofre Units 2 and 3, the switchyard has never been completely de-energized. Presently, eight offsite j transmission circuits serve San Onofre, whereas only two circuits are required by the technical specifications.

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1 The proposed 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit on the total annually allowed diesel out-of-service time in Modes 1, 2, 3 and 4 instead of the unlimited number of 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> outages currently allowed will serve as an incentive in scheduling and completing all diesel maintenance in such a manner that diesel availability remains high. If downtime in excess of the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit is needed, the technical specifications would require notification of the NRC instead of requiring plant shutdown. This provision is based on the recognition that exceeding the 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> limit in itself does not represent an unsafe condition. Rather, each individual case would be evaluated in the light of all the relevant factors and concerns. Based on the above, it is concluded that the introduction of an 800 hour0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> annual out-of-service limit will not result in the probability or consequences of an accident previously evaluated being increased.

(c) Basis t_o Technical Specification 3.8.1.1 The changes to the Basis are only for the purposes of updating and clarifying the text to be consistent with the proposed configuration of the diesel generator systems.

(d) Diesel Fuel Oil Testing Requirements By changing the current diesel fuel oil testing requirements to those that are in current industry use and that more accurately detennine fuel oil quality, the probability of degraded fuel is reduced. Therefore, the probability or consequences of previously evaluated accidents would not be increased.

2. The NRC staff proposes to detennine that operation of the facility in accordance with the proposed changes does not create the possibility of I

a new or different kind of accident from any accident previously evaluated because the proposed changes would not change the configuration of the plant, or its manner of operation. Rather, in order to prolong diesel engine life and provide better diesel maintenance, the proposed changes would reduce the amount of diesel testing and increase the time allowed for diesel repair and maintenance in individual cases. The safety requirement to complete cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> if a 'simiting condition for operation is not met would not be changed. Based on these considerations, the proposed changes would not create the possibility of a new or different kind of accident from any previously evaluated.

3. The NRC staff proposes to determine that operation of the facility in accordance with these proposed changes does not involve a significant reduction in a margin of safety because the proposed changes affect

! only the surveillance requirements for fast cold starts of the diesel engines and fuel oil testing. The proposed changes would reduce premature diesel engine degratition and increase assurance of fuel oil quality and thus increase the overall reliability of the diesel l generators. Therefore, operation in accordance with these proposed changes would not involve a reduction in a margin of safety.

Local Public Document Room Location: General Librbry, University of California at Irvine, Irvine, California 92713.

Morneyforlicensees: Charles R. Kocher, Esq., Sc'uthern California Edison Company, 2244 Walnut Grove Avenue, P. O. Box 800, Rosemead, California 91770 and Orrick, Herrington & Sutcliffe, Attn.: David R. Pigott, Esq., 600 Montgomery Street, San Francisco, California 94111.

NRC Project Director: George W. Knighton

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear

$nt, Units 1and2,HamiltonCounty, Tennessee Date of amendment request: December 17, 1986 Description of amendment request: The licensee proposes to increase the loading sequence delay for the containment spray pumps onto the emergency power supply powered by the diesel generators. This technical specification change would increase the response time in seconds by 150 seconds from 58 to 208 seconds.

Basis for proposed no sianificant hazards consideration determination: The Comission has provided guidance (51 FR 7744) concerning the application 'of standards for determining whether a significant hazards consideration exists. Example (v) identifies a proposed change as having no significant hazard if it is as follows:

Upon satisfactory completion of construction in connection with an operating facility, a relief granted from an operating restriction that was imposed because the construction was not yet completed satisfactorily. This is intended to involve only restrictions where it is justified that construction has been completed satisfactorily.

Since the proposed amendment makes the Technical Specifications con-sistent with the change of loading sequence of the containment spray, it is l

similartoExample(v).

Therefore, the Commission has proposed to find that the amendment contains no significant hazards consideration, l.oc_al Public Document Room location: Chattanooga-Hamilton County i Bicentennial Library, 1001 Broad Street, Chattanooga, Tennessee 37401 Attorney for licensee: Mr. Lewis E. Wallace, Acting General Counsel, Tennessee Valley Authority, 400 Comerce Avenue, E11833, Knoxville, Tennessee 37902 NRC Project Director: B. J. Youngblood

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah Nuclear Rant, Units 1 and 2, Hamilton County, Tennessee Date of amendn.ent request: January 20, 1987 Description of amendment r.equest: The proposed change would add a requirement to limiting condition for operation (LCO) 3.3.1.1 for Unit 1 and 3.3.1 for Unit 2. This additional requirement in the LCOs would demonstrate the operability of the shunt trip attachment to the reactor trip breaker.

Also, the amendment request would incorporate testing requirements in surveillance requirement 4.3.1.1 for both units for the reactor trip bypass breakers. This request was previously published on July 2, 1986 (FR 51 F'R 24264).

Basis for proposed no significant hazards consideration determination: The Consnission has provided guidance in the form of examples of amendments that are not considered likely to involve a significant hazards determination (FR 517744). Example (ii) states "a change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications: for example, a more stringent surveillance requirement."

As discussed above, the licensee's request to add a requirement for the operability of the new shunt trip attachment to the reactor trip breakers to LC0 3.3.1.1 for Unit 1 and LCO 3.3.1 for Unit 2 would be based on NRC Generic Letters 83-28 and 85-09. Therefore, this proposed change is similar to example (ii) and does not involve a significant hazard consideration.

Local Public Document Room location: Chattanooga-Hamilton County Bicentennial Library,1001 Broad Street, Chattanooga, Tennessee 37401 Attorney for licensee: Lewis E. Wallace, Esquire, Acting General Counsel Tennessee Valley Authority, 400 Commerce Avenue, E11B33, Knoxville, l

Tennessee 37902 N3 Project Director: B.J. Youngblood

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NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE During the period since publication of the last bi-weekly notice, the Commission has issued the following amendments. The Commission has determined for each of these amendments that the application complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Comission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility' Operating License and Proposed No Significant Pazards Consideration Determination and Opportunity for Hearing in connection with these actions was oublished in the FEDERAL REGISTER as indicated. No request for.a hearing or petition for leave to intervene was filed following this notice.

Unless otherwise indicated, the Comission has determined that these amendmentssatisfythecriteriaforcategoricalexclusibninaccordancewith 10 CFR 51.?2. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. If the Comission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a determination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the applications for amendments, (2) the amendments, and (3) the Commission's related letters, Safety Evaluations and/or Environmental Assessments as indicated.

. " ' s'

. . . + , . .

1 All of these items are available for public inspection at the Commission's l l l Public Document Room, 1717 H Street, N. W., Washington, D. C., and at the j l I

local public document rooms for the particular facilities involved. A copy '

l of items (2) and (3) may be obtained upon request addressed to the U. S.

Nuclear Regulatory Commission, Washington, D. C. 20555, Attention:

Director, Division of Licensing. <

Carolina Power & Light Company, Dockets Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North Carolina Date of application for amendments: March 17, 1986, as supplemented October 13, 1986.

Brief description of amendments: The amendments change the Technical -

Specifications (TS) by deleting nonsafety-related instruments from the surveillance requirements related to reactor coolant system leakage and leakage detection systems. The nonsafety-related instrument tag numbers are deleted from the TS Sections 4.4.3.1.b and 4.4.3.2.a.

Date of issuance: January 21, 1987 Effective date: January 21, 1987 Amendments Nos.: 103 and 133 Facility Operatino Licenses Nos. DPR-71 and DPR-6?. Amendment revised the Technical Specifications.

Date of initial notice in Federal Reaister: May 21, 1986 (51 FR 18679)

The October 13, 1986 submittal provided clarifying information and did not change the determination of the initial notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 21, 1987

^

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l No significant hazards consideration connents received: No. ,

Local Public Document Room location: University'of North Carolina at Wilmington, William Madison Randall Library, 601 S. College Road, Wilmington, North Carolina 28403-3297

?

f Carolina Power and Light Company, Docket No. 50-261 H. B. Robinson Steam 4

Electric Plant, Unit No. 2, Darlington County, South Carolina Date of application for amendment: October 13, 1986, as supplemented December 11, 1986 Brief description of amendment: The amendment revises Technical Specification Section 5.3.13 to: increase the fuel enrichment from 3.5 w/o to 3.9 w/o; reformat and rewrite section 5.4 to mention the previously

approved 21-inch center-to-center spacing of the new fuel storage racks; allow storage of fuel with a maximum axial plane enrichment of 3.9 w/o in both new and spent fuel racks; inclusion of the design k,ff for worst accident conditions; adding boron concentration for the spent fuel pit during fuel handling; and revising Table 4.1-2 to correct an error and specify a sampling requirement prior to new fuel movement in the spent fuel storage pit.

i Date of issuance: January 20, 1987 Effective date: January 20. 1987

Arr.cndment No. 112 Facility Operating License No. DPR-23. Amendment revised the Technical .

Specifications.

Date of initial notice in Federal Register: November 19, 1986 (51 FR 41846)

The December 11, 1986 letter, which supplemented the application did not '

change the initial determination published in the FEDERAL REGISTER.

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

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A The Commission's related evaluation of the amendment is contained in a .

Safety Evaluation dated January 20, 1987.

No significant hazards consideration comments received: No Local Public Document Room location: Hartsville Memorial Library, Home and Fifth Avenues, Hartsville, South Carolina 29535 i

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, Ryron Station, Units 1 and 2, Ogle County, Illinois Date of application for amendments: August 13, 1986 and August 27, 1986 Description of amendments : The amendment approves changes to the Technical Specifications that (1) replaces "86% of total volume" with "50%" for the water level in the ultimate heat sink cooling tower basin; (2) permits a crosstie between Units 1 and 2 Class IE 125-vde buses; and (3) deletes two f

pages that are no longer effective.

Date of issuance: December 12, 1986 Effective date: December 12, 1986 Amendment No.: 5 3 Facility Operating License Nos. DPR-37 and DPR-60 . Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: October 8, 1986 (50 FR 36084)

The Commission's related evaluation of the amendments is contained in a i Safety Evaluation dated December 12, 1986.

No significant hazards consideration comments received: No Local Public Document Room location: Rockford Public Library, 215 N. Wyman Street, Rockford, Illinois 61103 t

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Comonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, l' nits 1 and 2, Rock Island County, Illinois Date of application for amendments: June 28, 1985 Brief description of amendments: The amendments change the technical specifications to delete certain emergency diesel generator surveillance testing requirements. A portion of the amendment request has been denied by the Ccmission and a separate Notice of Denial of Amendment has been forwarded to the Office of the Federal Register for publication.

Date of issuance: January 21, 1987 Effective date: January 21, 1987 Amendment Nos.: 99 and 96 Facility Operating License Nos. DPR-29 and DPR-30. Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: August 14, 1985 (50 FR 327901.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 21, 1987.

No significant hazards consideration comments received: No.

Local Public Document Room location: Moline Public Library, 504 - 17th Street, Moline, Illinois 61265.

l l

l Commonwealth Edison Company, Docket No. 50-265, Guad Cities Nuclear Power Station, Unit 2, Rock Island County, Illinois Date of application for amendment: September 18, 1986, as clarified December 10 and 23, 1986.

Brief description of amendment: The amendment reflects Cycle 9 reload fuel

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transient analysis and amends the license to provide for Single Loop Ooeration as part of the Technical Specifications and not a specific License Condition.

Date of issuance: January 16, 1987 Effective date: January 16, 1987 Amendment Nos.: 95 Facility Operating License No. DPR-30. Amendments revised the license and the Technical Specifications.

Date of initial notice in Federal Register: November 5, 1986 (51 FR 40278).

Py letters dated December 10 and 23, 1986, Commorwealth Edison submitted clarifying infomation and written confirmation of comitments made to NRC regarding related plant operation. These submittals did not significantly change the initial application nor did they change the initial no significant hazards consideration determination. Therefore, no renotice of the application was warranted.

The Comission's related evaluation of the amendment is contained in a Safety Evaluation dated January 16, 1987.

No significant hazards consideration comments received: No.

l Local Public Document Foom location: Moline Public Library, 504 - 17th i

l Street, Moline, Illinois 61265.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, Charlevoix County, Michigan Date of application for amendment: August 15, 1986 Brief description of amendment: The amendment changes the Technical Specifications in support of the installation of new source range neutron

, monitoring instrumentation by changing the associated terminology of the previous existing system.

Date of issuance: January 28, 1987 Effective date: January 28, 1987 Amendment No. 87 Facility Operating License No. DPR-6 This amendment revised the Technical Specifications.

Date of initial notice in Federal Reaister: September 24, 1986 (51 FR 33948).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 1987.

No significant hazards consideration comments received: No.

Local Public Document Room location: North Central Michigan College, 1515 Howard Street, Petoskey, Michigan 49770.

I -

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, Michigan Date of application for amendment: September 30, 1986 Brief description of amendment: This amendment revises the Fermi-2 Technical Specifications changing the upper limit for the pressure l detector's alarm calibration setpoint of each control rod drive scram accumulator to be greater than or equal to 940 psig.

Date of issuance: January 28, 1987 l

Effective date: January 28, 1987 Amendment No.: 5 Facility Operat.ing License No. NPF-43: Amendment revised the Technical 4

Specifications.

f Date of initial notice in Federal Register: December 3,1986(51FR43679) i l

l

The Comission's related evaluation of the amendment is contained in a

Safety Evaluation dated January 28, 1987 No significant hazards consideration coments received

No Local Public Document Room Location: Monroe County Library System, 3700

! South Custer Road, Monroe, Michigan 48161 Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station Units 1 and 2. York County. South Carolina Date of application for amendments: April 9, 1986, as supplemented June 5 and September 2,1986 Brief description of amendments: The amendments modify Technical Specification Table 3.3-6, Table Notations, to allow changing the alarm / trip setpoint for the containment radiation monitor EMF-39.

Date of issuance: January 29, 1987 Effective date: January 29, 1987 Amendment Nos.: 21 and 11 Facility Operatino License Nos. NPF-35 and NPF-52. Amendments revised the j Technical Specifications.

Date of initial notice in Federal Register: August 27, 1986 (51 FR 30565)

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 29, 1987.

No significant hazards consideration coments received: No Local Public Document Room location,: York County Library, 138 East Black Street, Rock Hill, South Carolina 29730 l

pukePowerCompany,DocketsNos. 50-269, 50-270 and 50-287, Oconee Nuclear Station Units Nos.1, 2 and 3, Oconee County, South Carolina

, Date of application for amendments: January 14, 1986, as supplemented on April 10 and June 18, 1986, and January 15, 1987 Brief description of amendments: These amendments extended the duration of the licenses to 40 years from the date of issuance of the full pcwer licenses. Therefore, the Oconee Operating Licenses were extended to rebruary 6, 2013 for Unit 1; to October 6, 2013 for Unit 2; and to July 19, 2014 for Unit 3. All three Oconee Units would have expired on November 6, 2007 without these amendments.

Date of issuance: January 30, 1987 Effective date: January 30, 1987 Amendments Nos.: 153, 153 and 150 Facility Operating Licenses Nos. DPR-38, DPR-47 and DPR-55. Amendments revised the Facility Operating Licenses.

Date of initial notice in Federal Reaister: July 2, 1986 (51 FR 24254).

Since the date of the initial notice, the licensee submitted clarifying information dated June 18, 1986, and January 15, 1987. This information did not change the original application in any way and, therefore, did not warrant renoticing. The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated January 30, 1987.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501 West Southbroad Street, Walhalla South Carolina 29691 Fl.orida Power Corporation, et al., Docket No. 50-302, Crystal River Unit No. 3 Nuclear Generatino Plant, Citrus County, Florida Date of application for amendment: June 18, 1986, as amended July 23, 1986

~

Brief description of amendment: This amendment increased the high pressure trip setpoint from 2300 psig to 2355 psig and added anticipatory reactor trips on turbine trip and trip of both main feedwater pumps.

Date of issuance: January 21, 1987

, Effective date: January 21, 1987 Amendment No.: 95 Facility Operating License No. DPR-72. Amendment revised the Technical Specifications.

Date of initial notice in Federal Reafster: October 22,1986(51FR37509)

Since the date of the initial notice, the licensee submitted clarifying information dated October 24, 1986. This information did not change the original application in any way eid, cherefore, did not warrant renoticing.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 21, 1987.

No significant hazards consideration connents received: No.

Local Public Document Room location: Crystal River Public Library, 668 N.W.

First Avenue, Crystal River, Florida 32629

}

Iowa Electric Light and Power Company, Docket No. 50-331 Duane Arnold, Energy Center, Linn County, Iowa Date of application for amendme_nt: May 22, 1986

, Brief Description of amendment: The amendment revises the Duane Arnold Technical Specifications relative to diesel generator (DG) testing to make them responsive to GL 83-30 and 84-15 and Information Notice 85-32 and includes changes which meet the intent of GL 84-15 to reduce unnecessary DG

! testing. ,

I l

Date of issuance: January 20, 1987 Effective date: January 20, 1987 Arendment No.: 139 Facility Operating License No. DPR-49. Amendment revised the Technical Specifications.

Date of initial notice in Federal Reaister: July 2, 1986 (51 FR 24257)

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 20, 1987.

No significant hazards consideration coments received: No

. Local Public Document Room location: Cedar Rapids Public Library, 500 First Street, S.E., Cedar Rapids, Io,wa 52401.

l Iowa Electric Light and Power Company Docket No. 50-331, Duane Arnold I

DnergyCenter,LinnCounty, Iowa Date of application for amendment: November 14, 1986 as clarified January 2, 1987.

Brief Description of amendment: The amendment revises the current TS requirements to allow an extension, on a one-time-only basis, of approximately 10 weeks to the surveillance test intervals for the functional testing of snubbers, the local leak rate testing of primary containment isolation valves and penetrations and the replacement of the T-ring seals in the primary containment purge and vent valves.

Date of issuance

January 30, 1987 l

i Effective date: January 30, 1987 Amendment No.: _

140 Facility Operatina License No. DPR-49. Amendment revised the Technical Specifications.

- . - _ _ . - _ _ = _ - . . . _ . - - _ _ . . _ . . - . . - . _ _ _ . . . _ _ _ - _ _ _ _

Date of initial notice in Federal Register: December 17, 1986 (51 FR 45204)

The licensee's January 2,1987, letter furnished clarification fer staff review and decreased the extent of the original request. It did not, in any way, change the staff's conclusion in the above notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 30, 1987.

No significant hazards consideration comments received: No Local Public Document Room location: Cedar Rapids Public Library, 500 First Street, S.E., Cedar Rapids, Iowa 52401.

l =

Louisiana Power and Light Company, Docket No. 50-382 Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana Date of applications for amendment: October 1,1986, as supplemented by letters dated October 29 and November 19, 1986.

Brief description of amendment: The amendment chhnged the Technical Specification by revising the core protection calculator DNBR setpoint; (2) revising the core operating limit supervisory system out-of-service DNBR limits; (3) revising the peak linear heat rate; (4) revising the reactor protection instrumentation response times; and (5) revising the control element assembly insertion limits.

Date of issuance: January 16, 1987

_ Effective date: January 17, 1987 Amendment No.: 12 Facility Operating License No.: NPF-38: Amendment revised the Technical Specifications.

1

Dates of initial notices in Federal Register: November 19, 1986 (51 FR 41860 and 51 FR 41861), December 3, 1986 (51 FR 43683) and December 17, 1986 (51 FR 45208).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 16, 1987.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Louisiana Power and Light Company, Docket No. 50-382, Waterford Steam Electric Station, Unit 3 St. Charles Parish, Louisiana Date of applications for amendment: July 15, 1986.

Brief description of amendment: The amendment revised the Technical Specification by revising the axial shape index allowable ranges; revising the moderator temperature coefficient allowable range; revising the part-length control element assembly insertion limits; and allowing the suspension of the part-length control element assembly insertion limits during certain startup tests.

Date of issuance: January 16, 1987 Effective date: January 16, 1987 Amendment No.: 13 Facility Operating License No.: NPF-38: Amendment revised the Technical Specifications.

Dates of initial notices in Federal Reaister: Auaust 27,1986(51FR30575),

September 24, 1986 (51 FR 33952), and October 8,1986(51FR36093).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 16, 1987.

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No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Omaha Public Power District, Docket No. 50-285. Fort Calhoun Station, Unit No. 1. Washington County, Nebraska Date of application for amendment: October 23, 1986 Brief description of amendment: The amendment deleted the short tenn reporting requirements related to primary coolant specific activity levels ,

and no longer requires plant shutdown if the primary coolant specific activity exceeds the limit of 1.0 microcurie / gram dose equivalent I-131 for an accumulated period over 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in a 12-month period.

Date of issuance: January 20, 1987 Efffective date: January 20, 1987 Amendment No.: 102 Facility Operating License No. DPR-40. Amendment revised the Technical i Specifications.

Date of initial notice in Federal Register: December 17,1986(51FR45191 at45212)

T'he Coninission's related evaluation of the amendment is contained in a Safety Evaluation dated January 20, 1987.

No significant hazards consideration comments received: No.

_ Local Public Document Room location: W. Dale Clark Library, 215 South 15th Street, Omaha, Nebraska 68102 Roche. ster Gas and Electr,1c Corporation, Docket No. 50-244, R. E. Ginna Nuclear P.ower Plant, Wayne County, New York Date of application for amendment: October 15, 1986

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Brief description of amendment: The review and audit section (6.7) of the Technical Specifications ~(TS) requires the licensee to have an independent audit and review group known as the " Nuclear Safety Audit and Review Board (NSARB)". The NSARB provides an independent review and audit of activities dealing with plant operations, engineering design changes, radiological safety, and quality assurance practices. The amendment increased the membership of the plant staff to the NSARB from two to three members and to allow the three members to vote. A restriction was imposed to the quorum requirements for the board to assure that plant personnel would not make up

- a majority of the board members.

Date of issuance: January 28, 1987.

Effective date: January 28, 1987.

Amendment No.: 21.

Facility Operating License No. DPR-18: Amendment revised the Technical Specifications.

Date of initial notice in Federal Reoister: November 19, 1986 (51 FR 41868) l The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 28, 1987.

i No significant hazards consideration comments received: No.

Local Public Document Roon location: Rochester Public Library, 115 South Avenue, Rochester, New York 14610.

N_RC Protect Director: George E. Lear, Director.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-362,

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San Onofre Nuclear Generatina Station, Units 2 and 3, San Diego County, l

. California Dates of applications for amendments: March 2 and April 2, 1984

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Brief description of amendments: The amendments revise Technical Specification 3/4.3.3.8, " Radioactive Effluent Monitoring Instrumentation".

Date of issuance: January 20, 1987 Effective date: January 20, 1987, to be implemented within 30 days of issuance.

Amendment Nos.: 57 and 46 Facility Operatino License Nos. NPF-10 and NPF-15: Amendments revise the __

Technical Specifications.

Date of initial notices in Federal Register: February 27,1985(50FR8007).

The Comission's related evaluation of the amendments is contained in a Safety Evaluation dated January 20, 1987.

No significant hazards consideration comments received

No.

Local Public Document Room location: General Library, University of California at Irvine, Irvine, California 92713.

Vermont Yankee Nuclear Power Corporation. Docket No. 50-271. Vermont Yankee Nuclear Power Station. Vernon. Vermont Date of applications for amendment: November 2, 1984 as supplemented March 4, 1986, and application dated December 29, 1981 Grief description of. amendment: The amendment changes the Technical Specifications to provide trip settings, operability requirements and testing requirements for control and instrumentation circuitry which provides protection in case of degraded grid voltage. The amendment also provides limiting conditions of operation and surveillance requirements for noble gas effluent monitors.

Date of issuance: January 29, 1987

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Effective date: January 29, 1987 Amendment No.: 98 Facility Operating License No. DPR-28: Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: February 27, 1985 (50 FR 8010)

August 23, 1983 (48 FR 38425) i The March 4, 1986 submittal supplemented the November 2, 1984 application.

That supplement provided clarifying information and did not change the detennination of the initial notice.

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 29, 1987.

No significant hazards consideration comments received: No 4

Local Public Document Room location: Brooks Memorial Library, 224 Main Street, Brattleboro, Vennont 05301.

Washinoton Public Power Supply System, Docket No. 50-397, WNP-2, Richland, Washington Dates of amendment request: April 8, 1986, and clarified November 20, 1986 Brief description of amendment: This amendment revises Sections 3.3.7.8 and 4.3.7.8 (Chlorine Detection System), Section 4.7.2 (Control Room Emergency FiltrationSystem)andBasesSection3/4.3.7.8(ChlorineDetectionSystem) of the WNP-2 Technical Specifications by eliminating portions of these sections because chlorine gas is no longer stored in the immediate plant site area. As a result, the threat to control room habitability due to chlorine gas leakage has been eliminated; and the Technical Specification sections are no longer relevant.

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Date of issuance: January 21, 1987 Effective date: January 21, 1987 Amendment No.: 36

. Facility Operating License No. NPF-21: Amendment revises the Technical Specifications.

Date of initial notice in Federal Reafster: September 10, 1986 (51 FR 32281).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated January 21, 1987.

No significant hazards consideration comments received: No Local Public Document Room location: Richland Public Library, Swift and '

Northgate Streets, Richland, Washington 99352 Wisconsin Public Service Corporation, Docket No. 50-305 Kewaunee Nuclear Power Plant, Kewaunee County. Wisconsin ,

Date of application for amendment: August 1, 1986 and as supplemented October 1, 1986.

Brief description of amendment: The amendment corrects errors, replaces obsolete references with current references and reinserts an inadvertently deleted requirement in the Technical Specifications. The Operating License is updated to reflect a previous License Amendment.

Date of issuance: January 21, 1987.

Effective date: January 21, 1987.

Amendment No.: 71.

Facility Operattnq License No. DPR-43. Amendment revised the Technical Specifications.

Date of initial notice in Federal Register: September 10,1986(51FR32282) l l

l l

i

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w

.. The Comission's related evaluation of the amendment is contained in a Q Safety Evaluation dated January 21, 1987.

No significant hazards consideration coments received: No.

a local Public Document Room location: University of Wisconsin Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 54301 NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE AND FINAL DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION AND OPPORTUNITY FOR HEARING (EXIGEfiT OR EMERGENCY CIRCUMSTANCES)

During the period since publication of the last bi-weekly notice, the

Comission has issued the following amendments. The Commission has detennined for each of these amendments that the application for the .

am9ndment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Comission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment.

Because of exigent or emergency circumstances associated with the date

the amendment was needed, there was not time for the Comission to publish,

' for public comment before issuance, its usual 30-day Notice of Consideration

'3 Detennination and Opportunity for Hearing. For exigent circumstances, the Comission has either issued a FEDERAL REGISTER notice providing opportunity for public coment or has used local media to provide notice to the public

.~ in the area surrounding a licensee's , facility of the licensee's application

-w- - , - - . , . . +m ,..w - - - . - . , . . - , _ _ _ _ _ , _ - _ . - - - . - , , , - , , . . , . , . , , _ , ..--.--w ._.,7,-,-pv 7-

- 106 -

and of the Commission's proposed detennination of no significant hazards consider-ation. The Comission has provided a reasonable opportunity for the public to comment, using its best efforts to make available to the public means of comunication for the public to respond quickly, and in the case of telephone comments, the comments have been recorded or transcribed as appropriate and the licensee has been informed of the public coments.

In circumstances where failure to act in a timely way would have resulted, for example, in derating or shutdown of a nuclear power plant or in prevention of either resumption of operation or of increase in power output up to the plant's licensed power level, the Commission may not have had an opportunity to provide for public comment on its no significant hazards determination. In such case, the license amendment has been issued without opportunity for comment. If there has been some time for public comment but less than 30 days, the Comission may provide an opportunity for public comment. If coments have been requested, it is so stated. In either event, the State has been consulted by telephone whenever possible.

Under its regulations, the Commission may issue and make an amendment 1

imediately effective, notwithstanding the pendency before it of a request for a hearing from any person, in advance of the holding and completion of any required hearing, where it has determined that no significant hazards consideration is involved.

The Commission has applied the standards of 10 CFR 50.92 and has made a final determination that the amendment involves no significant hazards consideration. The basis for this determination is contained in the documents related to this action. Accordingly, the amendments have been l issued and made effective as indicated.

l

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N

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Unless otherwise indicated, the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), no environmental f

impact statement or environmental assessment need be prepared for these amendments. If the Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(b) and has made a detennination based on that assessment, it is so indicated.

For further details with respect to the action see (1) the application for amendment, (2) the amendment to Facility Operating License, and (3) the Commission's related letter, Safety Evaluation and/or Environmental Assessment, as indicated. All of these items are available for public i

inspection at the Commission's Public Document Room,1717 H Street, N. W.,

Washington, D. C., and at the local public document room for the particular facility involved.

A copy of items (2) and (3) may be obtained upon request addressed to i

the U. S. Nuclear Regulatory Commission. Washington, D. C. 20555, Attention: Director, Division of Licensing.

l The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendments. By March 13, 1987, the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating license and any person whose interest may be affected by this proceeding and who wishes to participate as a party in the proceeding must file a written petition for leave to intervene.

Requests for a hearing and petitions for leave to intervene shall be filed in accordance with the Comission's " Rules of Practice for Domestic Licensing Proceedings" in 10 CFR Part 2. If a request for a hearing or

- 108 -

petition for leave to intervene is filed by the above date, the Comission or an Atomic Safety and Licensing Board, designated by the Comission or by  ;

the Chairman of the Atomic Safety and Licensing Board Panel, will rule on the request and/or petition and the Secretary or the designated Atomic Safety and Licensing Board will issue a notice of hearing or an appropriate order.

As required by 10 CFR 52.714, a petition for leave to intervene shall set forth with particularity the interest of the petitioner in the proceeding and how that interest may be affected by the results of the proceeding. The petition should specifically explain the reasons why intervention should be permitted with particular reference to the following factors: (1) the nature of the petitioner's right under the Act to be made a party to the proceeding; (2) the nature and extent of the petitioner's property, financial,orotherinterestintheproceeding;and(3)the possible effect of any order which may be entered in the proceeding on the petitioner's interest. The petition should also identify the specific aspect (s) of the subject matter of the proceeding as to which petitioner wishes to intervene. Any person who has filed a petition for leave to intervene or who has been admitted as a party may amend the petition without requesting leave of the Board up to fifteen (15) days prior to the first l

prehearing conference scheduled in the proceeding, but such an amended petition must satisfy the specificity requirements described above.

Not later than fifteen (15) days prior to the first prehearing

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conference scheduled in the proceeding, a petitioner shall file a supplement to the petition to intervene which must include a list of the contentions which are sought to be litigated in the matter, and the bases for each l

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contention set forth with reasonable specificity. Contentions shall be limited to matters within the scope of the amendment under consideration. A petitioner who fails to file such a supplement which satisfies these requirements with respect to at least one contention will not be permitted to participate as a party.

Those permitted to intervene become parties to the proceeding, subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fully in the conduct of the hearing, including the opportunity to present evidence and cross-examine witnesses.

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Since the Commission has made a final determination that the amendment involves no significant hazards consideration, if a hearing is requested, it will not stay the effectiveness of the amendment. Any hearing held would take place while the amendment is in effect.

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Comission, U.S. Nuclear Regulatory Commis-sion, Washington, D.C. 20555, Attention: Docketing and Service Branch, or l

may be delivered to the Commission's Public Document Room, 1717 H Street, N.W., Washington, D.C., by the above date. Where petitions are filed during the last ten (10) days of the notice period, it is requested that the peti-i tioner promptly so inform the Commission by a toll-free telephone call to Western Union at (800) 325-6000 (in Missouri (800) 342-6700). The Western Union operator should be given Datagram Identification Number 3737 and the following message addressed to (Branch Chief): petitioner's name and telephone number; date petition was mailed; plant name; and publication date and page number of this FEDERAL REGISTER notice. A copy of the petition should also be sent to the Executive Legal Director, U.S. Nuclear Regulatory Commission Washington, D.C. 20555, and to the attorney for the licensee.

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Nontimely filings of petitions for leave to intervene, amended petitions, supplemental petitions and/or requests for hearing will not be entertained absent a determination by the Comission, the presiding officer or the Atomic Safety and Licensing Board designated to rule on the petition and/or request, that the petitioner has made a substantial showing of good cause for the granting of a late petition and/or request. That determination will be based upon a balancing of the factors specified in 10 CFR2.714(a)(1)(1)-(v)and2.714(d).

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, City of Dalton, Georgia., Dockets Nos. 50-321 and 50-366 Edwin I. Hatch Nuclear Plant, Units Nos. I and 2 Appling County, Georgia Date of application for amendments: November 20, 1986, as supplemented January 27, 1987 Brief description of amendments: The amendment changes the Technical Specifications to delete the requirement that snubbers are declared inoperable if visible signs of leakage are present.

Date of issuance: January 29, 1987 Effective date: January 29, 1987 MndmentsNos.: 134 and 72 Facility Operatino Licenses Nos. DPR-57 and NPF-5. Amendments revised the Technical Specifications.

l l Public comments requested as to proposed no significant hazards consideration: No.

The Commission's related evaluation of the amendment and final determination l

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of no significant hazards consideration are contained in a Safety Evaluation dated January 29, 1987 Attorney for licensee: Bruce W. Churchill, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N. Street, N.W., Washington D.C. 20037 Local Public Document Room Location: Appling County Public Library, 301 City Hall Drive, Baxley, Georgia.

Georgia Power Company, Oolethorpe Power Corporation Municipal Electric, c Authority of Georcia, City of Dalton,_ Georgia, Docket No. and 50-366, Edwin I. Hatch Nuclear Plant, Unit No. 2. Appling County, Georgia Date of application for amendment: January 27, 1987 B.rief description of amendment: The amendment adds a note to the Technical Specifications in support of a waiver granted orally on January 26, 1987, to restart following a forced outage that occurred earlier the same day. The

, waived requirement was one that prevents Unit 2 from changing modes of 1

operation with the standby service water system inoperable.

Date of issuance: January 30, 1987 Effective date: January 26, 1987 Amendm.ent No.: 73 l Facility Operating License No. NPF-5. Amendment revised the Technical Specifications.

Public comments requested as to proposed no significant hazards consideration: No.

The Commission's related evaluation of the amendment and final detemination of no significant hazards consideration are contained in a Safety Evaluation dated January 30, 1987.

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Attorney for licensee: Bruce W. Churchill, Esquire, Shaw, Pittman, Potts and Trowbridge, 2300 N. Street, N.W., Washington D.C. 20037 Local Public Document Room Location: Appling County Public Library, 301 City Hall Drive, Baxley, Georgia.

Dated at Bethesda, Maryland this 5th day of February,1987.

FOR THE NUCLEAR REGULATORY COMMISSION A u a_- 6 E.'WayneHouston,ActingDirector Division of BWR Licensing Office of Nuclear Reactor Regulation