ML17300A708

Forwards Biweekly Notice,Applications & Amends to OLs Re 870226 NSHC

ML17300A708
Person / Time
Site:	Palo Verde
Issue date:	03/23/1987
From:	Jun Lee Office of Nuclear Reactor Regulation
To:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
References
NUDOCS 8703260338
Download: ML17300A708 (38)

Text

MAR 2 3 1987

~ DISTRIBUTION:

~Docket Fi 1 e PDR LPDR PRC System PD7 Reading JLee (2)

DOCKET NO(S).

50-528/529 Mr. E.

E. Van Brunt, Jr.

Executive Vice President Arizona Nuclear Power Project P.O.

Box 52034 Phoenix, Arizona 85072-2034

SUBJECT:

ARIZONA PUBLIC SERVICE COMPANY ET AL.

PALO VERDE NUCLEAR GENERATING STATION, UNITS 1 AND 2 Notice of Receipt of Application, dated Draft/Final Environmental Statement, dated Notice of Availabi1 ity of Draft/Final Environmental Statement, dated Safety Evaluation Report, or Supplement No.

dated Environmental Assessment and Finding of No Si gnificant Impact, dated Q Notice of Consideration of Issuance of Facility Operating License or Amendment to Facility Operating License, dated g] Bi-Weekly Notice; Applications and Amendments to Operating Licenses Involving No Significant Hazards Considerations, dated 2/26/8~

[see page (e) ]

Exemption, dated Construction Permit No.

CPPR-

, Amendment No.

da'ted Facility Operating License No.,

Amendment No.

Order Extending Construction Completion Date, dated Monthly Operating Report for transmitted by letter dated Annual/Semi-Annual Report-dated transmi tted by letter dated The following documents concerning our review of the subject facility are transmitted for your information ~

Enclosures:

As stated cc:

See Next Page Office of Nuclear Reactor Regulation

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Yan Brunt, Jr.

Arizona Nuclear Power Project Palo Yerde CC:

Arthur C. Gehr, Esq.

Snell 5 Wflmer 3100 Yallcg Center Phoenix, Arizona 85073 Mr. James M. Flenner, Chief Counsel Arizona Corporation Coeafssfon 1200 West Washington Phoenix, Arizona 85007 Chat les R. Kocher, Esq. Assistant Council James A. Boeletto, Esq.

Southern California Edison Company P. 0.

Box 800

Rosemead, Cal ffornfa 91770 Mr. Mark Gfnsbcrg Energy Director Office of Economic Plannfng and Development 1700 West Washington - 5th Floor Phoenix, Arizona 85007 Mr. Wayne Shirley Assistant Attorney General Bataan Memorial Building Santa Fe, New Mexico 87503 Mr. Roy Zfrwerman U.S. Nuclear Regulatory Coamfssion P. 0.

Box 239 Arlington, Arizona 85322 Hs. Patricia Lec Hourfhan 6413 S. 26th Street Phoenix, Arizona 85040 Regional Administrator, Reafon Y

U. S. Nuclear Reaulatory Comnfssfon 1450 Maria Lane Suite 210 Walnut Creek, California 94596 Kenneth Berlin, Esq.

Winston 5 Strawn Suite 500 2550 M Street, NW Washfnoton, DC 20037 Ms. Lynne Bernabei Government Accountability Project of the Institute for Policy Studies 1901 gue Street, NW Washington, DC 20009 Mt. Ron Rayner P. 0.

Box 1509

Goodyear, AZ 85338 Mr. Charles B. Brinkman, Manaaer Washfnaton Nuclear Operations Combustion Engfneerfno,

?nc.

7910 Woodmont Avenue Suite 1310

Bethesda, Maryland 20814

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Chairman Arizona Corporatfon Commfssfon Post Offfce Box 6019 Phoenix, Arizona 85003 Arizona Radiation Regulatory Agency ATTN:

Ms. Clara Palovfc, Lfbrarfan 4814 South 40 Street Phoenix, Arizona 85040 Hr. Charles Tedford. Ofrector Arfzona Radiation Regulatory Agency 4814 South 40 Street Phoenix, Arfzona 85040 Chairman Narfcopa County Board of Supervfsor s 111 South Third Avenue Phoenfx, Arfzona 85003

1

Federal Register'/ Vol. 52, No. 38 / Thursday, February

28. 1987 / Notices 5849 Bi-Weekly Notice; Appiicationa and Amendmenta to OperatIng Ucensee Involving No Significant Hazards Consideratlona APPLICATIONSAND L Background Pursuant to Public Law (Pub. L)97-415, the Nuclear Regulatory Commission (the Commission ls publishing this regular bi.weekly notice. Pub. L 97%15 revised section 189 of the Atomic Energy Act of 1954, as amended (the Act), to require the Commission 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 Commission the authority to issue and make immediately elfective any amendment to an operating license upon a determination by the Commission that such.

amendmcnt involves no significant hazards consideration. notwithstanding the pendency belore the Commission of a request for a hearing from any person.

This bi-weekly notice includes all amendmcnts issued, or proposed to be issued, since the date of publication of the last bi-weekly notice which was published on February 11, 1987 (52 FR 4400).

NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENTTO, FACIUTYOPERATING LICENSE AND PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATIONAND OPPORTUNITY FOR HEARING The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92. this means that operation of the facilityin accordance with the proposed amendments would not (1) involve a significant increase in the probability or consequences ol 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 signilicant reduction In a margin of safety. The basis for this proposed determination for each amendment request is shown below.

The Commission is seeking public comments on this proposed determination. Any comments received within 30 days after the date of publication of this notice willbe considered in making any final determination. The Commission willnot normally make a final determination unless it receives a request for a hearing.

Written comments may be submitted

'y mail to the Rules and Procedures Branch, Division of Rulesand Records, Office of Administration. U.S..Nuclear Regulatory Commission, Washington, DC 20555, and should cite the publication date and page number of this Federal Register 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 5:00 p.m. Copies of written comments received may be examined at the NRC Public Document Room, 1717 H Street.

NW., Washington, DC. The filingof requests for hearing and petitions for leave to intervene is discussed below.

By March 30, 1987. the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facilityoperating license and any person whose interest,may bc 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 ol Practice lor Domestic Licensing Proceedings" in 10 CFR Part 2. Ila 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, willrule on the request and/or petition and the Secretary or the designated Atomic Safety and Licensing Board willissue a notice of hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene shall set for'th 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. or other interest in

'he proceeding; 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

5850'ederal Register / Vol. 52, No. 38 / Thursday, February 28, 1987 / Notices admitted ae ~ party may amend the petition without 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 des'cribed 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 contentione which are sought to be litigated in the matter. and the bases for each contention set forth with reasonable s pecificity. 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 willnot be permitted to participate as a party.

Those permitted to intervene become parties io the proceeding. subject to any limitations in the order granting leave to intervene, and have the opportunity to participate fullyin the conduct ol the hearing, including the opportunit'y to present evidence and crosswxamine witnesses.

. Ifa hearing is requested, the Compission willmake a final deti rmination on the issue of no sig'..iicant hazards consideration. The fmal determination willserve to decide when the hearing is held.

Ii thi final determination is that the amendment request involves no significant hazards consideration. the Commission may issue the amendment and make it immediately effective.

notwithstanding the request for a hearing. Any hearing held would take pla'ce after issuance of the amendment.

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

Normally, the Cominission willnot issue the amendment until the expiration 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 Commission may issue the license amendment before the expiration of the ~ay notice period, provided that ite final determination is that the amendment involves no significant hazards consideration. The final determination willconsider all public and State comments received before action le taken. Should the Commission take this action. it will pubhsh a notice of issuance and provide for opportunity for a hearing after issuance. The Commission expects that the need to take this action willoccur very infrequently.

A request for a hearing or,'a petition for leave to Interv'ene must be filed with the Secretary of the Commission. U.S.

Nuclear Regulatory Commission, Washington, DC 20555, Attention:

Docketing and Service Branch, or may be delivered to the Commission'e Public Document Room, 1717 H Street, NW.,

Washington, DC, 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 Commission by a toll-free telephone call to Western Union at (800) 325-8000 (in Missouri (800) 342-8700).

The Western Union operator should be given Datagram 81dentification Number 3737 and the followingmessage addressed to (Proj ect Dizectorj:

petitioner'e 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 Office of the General Counsel-Bethesda, US. Nuclear Regulatory Commission. Washington, DC 20555. and to the attorney for the licensee.

Nontimely filings ofpetitions for leave to intervene. amended petitions.

supplemental petitions and/or requests for hearing willnot 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 offactors specified in 10 CFR 2.714[a)[1)(i)-(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. NW..

Washington, DC, and at the local public document room for the particular facility involvetL Carolina Power tk Light Company, Dockets Nos. 50-325 and 50-324, Brunswick Steam Electric Plant. Units I and 2, Brunswick County, North Carolina Date ofapplication for amendments:

November 1-3. 1986, as supplemented january 2L 1987.

Descrtpttan ofamendtnent request:

The proposed amendment would change the Techrucai Specificatione (TS) for Brunswick Steam Electric Plant. Units 1 and 2. The proposed change deletes the requirement io perform response time testing on various teinperature switches and thermocouples listed in TS Table 3.3.2-3.

Currently, Surveillance Requirement 4.3.2.3 requires that the isolation system response time of the applicable functions shown in Table 3.3.2-3 shall be demonstrated to be within its limit at least once per 18 months. As part of this requirement. Table 3.3.2-3 includes the testing of: (1) Primary containment isolation within 13 seconds of indication of high main steam line tunnel temperature (Item 1.d): (2) reactor water cleanup system (RWCS) isolation within 13 seconds of indication of high area temperature (item 3.b); (3) RWCS isolation within 13 seconds of indication ofhigh area ventilation temperature gradient (Item 3.c); and (4) high pressure coolant Injection isolation within 13 seconds of indication of high HPCI steain line tunnel temperature (item 4.a.4). Temperature sensors are provided primarily for detection of small line breaks. The Final Safety Analysis Report [FSAR) does not take credit for temperature sensor initiation of main steam line isolation in the event of a main steam line break. Therefore, these sensors are nol subject to the design base accident response time.

The present surveillance program utilizes two distinct testing methods to determine the response time of the temperature sensors for the systems described above. However, the test results are not repeatable and. thus. do not give a reliable indication of the real response time. The Brunswick FSAR safety analysis does not address individual temperature sensor response times or the response times of the logic systems to which the temperature sensors are connected. The isolation times ofprimary containment. RWCS.

and HPCI can be determined by other parameters such as reactor Iow level, high steam flow. or system lov'ressure.

In addition. the testing of the RWCS thermocouples requires removal of the devices. This may result in increased instrument wear causing decreased accuracy and reliability.

Basis forproposed no significant hazards consideration determination:

The Commission has provided standards for determining whether a significant hazards determination exists as stated in 10 CFR 50,92(c). A proposed amendment to an operating license involves no significant hazards considerations ifoperation of the facility ln 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 acrident fr<im

Federal Reg(ster / Vol. 52, No. 38 / Thursday, February 26, 1987 / Notices 5851'ny accident prcviausly evaluated, or (3 involve a significant reduction in a margin of safety.

The licensee has evaluated the proposed amendmcnt against Ihe standards in 10 CFR 50.92 and haa determined the!ollowing:

t. The prnpased amendment does not involve s signihcant mcrease in the probability or consequences of an accident previously evaluated because the FSAR snf<<ty analysis does not address Individual temperature sensor response times or the response times of the logic systems to which iii<< temperature sensors are connected. [The, licensee has verified that credit is'not taken Ior ih<<sc t mperaiure sensor response times m uny acciit<<nt analysis.) The purpose of thes<<siinsors is to initiate isolation of a given systrm in the event oi a breach In the pr<<ysure boundary. The operability status of their safety related function is adequately mnnitor<<d by the remaining surveillance rrquircments of TS 3/4.3.2 li.e.. channel ch<<ck. channel calibration. channel luiiciional test. and logic system functional i<<sl),

=

Z. For the same reasons as given in Item1, the proposed amendment does not create the pi>ssibiiity of s new or different kind of ucudcnt than prev'iously evaluated.

3, The prnposei) amendmcnt does nut iniiiive a significant reduction in a margin of s iieiyThc affected isolation actuation uisirumcntat'ion response times are only mriusur<<it and recorded to enhance overall s)stem reliability and to monitor instrument channel response time trends. Since the i<<mperniure sensors discussed in this request do noi provide repeatabie results when i<<sird, the data collected does not serve any useful purpose. The primary containment.

IIWCS. and HPCI system isolation time is, more accurately tested by using other purometcrs ipressure. Ilow or level) currently iesi<<d I)y surveillance requirements.

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

The NRC staff has reviewed the licensee's no significant hazards consideration'determination ond agrees with the licensee's analysis. Based on this review. the staff therefore proposes to dr!ermine that the requested amcnilmcnt does not involve a sigrificant hazards consideration..

Local Public Document Room lociil>om University of North Carolina at IVt)mtngton, William Madison Randall i.tbrhry. 601 S. College Road,

'tVi!nunl)ton. North Carolina 28403-3297.

rl<<urni y for liccns'cer Thomas A Baxt< r. Fsquire. Shaw. Pittman. Potta and Trowbridge. 2300 N Street NW..

tVust:,nuton. DC 20037.

.'VRC PriiicclDirector. Daniel R.

Ytu)lcr

)

Carolina Power and Light Company, Docket No. 50-281, H. B. Robinson

-. Stcam Electric Plant. Unit No. 2, Darlingtoa County, South Carolina Dole ofamendmenl requesl:

December 18, 1988.

Descripiian ofamendment request:

The proposed amendment would revise Technical Specifications (TS) for the H.

B. Robinson Steam Electric Plant, Unit No. 2. The proposed revision involves revising Technical Specification Heatup

" and Cooldown curves, Figures 3.1-1 and 3.1-2. The Bisting curves expire soon afler the licensee's planned refueling outage in March of 1987 as 10 Effective Full Power Years (EFPY) are reached.

The new curves retain the same numbers which currently apply to the existing curves but with an addition of designators (a).or (b) to differentiate between the new 12.5 and 15 EFPY curves. Appropriate changes to Section 3.1.2 of Ihe TS willalso be made to direct the use of the appropriate set of the curves.

Basis forproposed na significant hazards consideration determination:

The Commission haa provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a, facilityinvolves no significant hazards consideration ifoperation 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 previous)y evaluated: or (3) involve a significant reduction in a margin of safety.

Carolina Power and Light Company haa reviewed their proposed change in

. accordance with10CFR 50.92(c) and has determined that the proposed change does not:

(1) Invo)ve a significant increase in the probability or consequences of any accident previously evaluated because the change orily updates existing operating limIts with new limits which compensate for continuing neutron, fluence. The basis for the analysis remain unchanged; (2) Create the possibility of a new or different kind of accident than previously evaluated because the change does not alter facilityoperation in such a manner or to an extent as to generate a hew or different kind of accident. Slightly more restrictive operating limits are not expected to significantly change the operation of the facility; (3) Invoive a significant reduction in a margin of safety because the same techniques and margins have been applied that were previously used. Only thc flucncc-dependent parameters have been ch:ingcd to update the analysis to compensate for the increased integrated power value. The NRC staff has reviewed tho licensee's determination and agrees with their evaluation in this regard and, therefore. proposi.s ta determine that Ihe proposed change does not involve a significant hazards consideration.

Loca/ Public Document Room lacalion Hortsville Memorial Library, Home ond Fifth Avenues, Hartsvitle.

South Carolina 29535.

Attorneyfor licensees Shaw, Pittman, Potts. and Trowbridge. 2300 N Street NW. ~ Washington. DC. 20037.

KRC Project Directors Lcstcr S.

Rubcnstein.

Carolina Power aad Light Company.

Docket No. 50-281, H. B. Robinson Steam Electric Plant, Unit No.'2, Darlington County, South Carolina Dole ofamcndmenl request: January 12, 1987.

Description ofamendmenl request:

The proposed amendment would revise Technical Specification (TS) Table 4.1-1 for the II.B. Robinson Steam Electric Plant. Unit No. 2. The proposed change, would revise the Technical Specifications to add r'equircmcnts for independent testing of the undcrvo)tage and shunt trip attachments of the reactor trip breakers on a monthly basis, testing of the bypass breakers prior to

'se, and independent testing of the control room manual switch contacts and wiring during et'tch refueling outage.

Basis forproposed na significant hazards consideration delerminalioru The Commission has prbvided standards for determining ss'hcthcr "a

significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation of the facility in accordance with the proposed amendment would not (I) invo)vq a signiflcant increase in thc'probability or consequences of an accident previous)y 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.

Carolina Power tk'Light Company has reviewed their proposed change in accordance with 10 CFR 50.92(c) and has determined that the proposed change does not:

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

Federal Register / Vol. 52, No".38 / Thursday, February 28, 1987 / Notices the proposed change adds new testing requirements to help insure the functioning of the reactor trip system when needed. The change is intended to reduce the frequency of postulated Anticipaled Transients Wilhout Scram (ATWS) events.

(2) Create the possibility of a new or different kind of accident than previously evaluated because the proposed change adds new testing requirements intended to reduce the frequency of postulated ATWS events.

{3) Involve a significant reduction in a margin of safety because the proposed

. change adds additional testing rcquiremcnls which effects an increase in the margin of safety. The NRC staff has reviewed the licensee's determination and agrees with their evaluation in this regard and. therefore.

proposes lo determine that the proposed change does not involve e significant hazards consideration.

.Locol Public Document Room

" locotioru Hartsville Memorial Library.

Home and Fifth Avenues, Hartsville.

South Carolina 29535.

Artoineyfor licensee: Shaw, Pittman, Potts, and Trowbridge. 2300 N Street r'VV.. Wastungton. UC. 20037.

NRC eject DI'rector. Lester S.

Rubcnstcin.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station. Units 1 and 2, York County, South Carolina Dale nfamendment request: july 31, 1985, as supplemented October 10. 1988.

Description ofamendment request:

The proposed amendments would revise Technical Specifics tion {TS)

Surveillance Requirement 4.8.2.1.1a.3) concerning electrolyte leakage. The proposed revision would change the requirement to demonstrate the

~ operability of the 125 volt battery bank 3nd cha gcr uy weiifyu'<8, at Icaat uncs pcr 7 days. that,"thcre is no visible indicalion of elcclrolytc leakage" to the requirement that "there is no visible indication of damaging electrolyte leakage." This change would allow battery operation when there is minor electrolyte leakage.

By letter dated August 15. 1988, the NRC staff requested additional information regarding the proposed smendmenls. By,letter dated October 10, 1986. the licensee modified the july 31,

'1985. request lo alleviate the staffs concerns regarding battery electrolyte leakage and significant discharge of a battery cell.

'hc proposed change is justified because Surveillance Requireinent 4.8.2.1.1a.1) requires that the paramelers in Table 4.8-3. Category A. be met every 7 days. Electrolyte leakage is one of these parameters. The limits on these arameters ensure that the electrolyte evel is sufficient for battery operability.

Thus, some minor electrolyte leakage, that would be unlikely to cause any significant discharge of e battery cell, can be allowed as long as the minimum electrolyte level is maintained.

Basis forproposed no significant hazards consideration determi notion:

The Commission has provided certain examples (51 FR 7744) of actions likely to involve no significant hazards considerations. The request Involved in this case does not match any of those examples. However, the staff has reviewed the licensee's request for the above amendments and.determined that should this request be implemented. it would not (1) involve a significant increase in the probabi'Iity or consequences of an'accident previously evaluated because the operability of the battery bank and charger continues to be ensured via the implementation of

~

existing Surveillance Requirement 4.8.2.1.1a.1). Also, It would not {2) create the possibility of a new or different kind of accidenl from any accident previously evaluated because the proposed change would not affect the applicable accident analyses and would not introduce new modes of operation. Finally, it would not (3) involve a significant reduction in a margin of safety because the proposed change has no adverse impact on safety and does not reduce the safety margin.

Accordingly, the Commission has determined that the above changes involve no significant hazards consideration.

Local Public Document Room location: York County Library. 13S East Black Street. Rock Hill,South Carolina 29730.

AtloI7ieyforlicensee: Mr. Albert Carr.

Duke Power Company. 422 South Church Street. Charlotte, North Carolina 28242.

NRC Proj ect Director. B.J.

Youngblood.

'a Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba Nuclear Station. Units 1 and 2, York County, South Carolina Date ofamendment request. July 31, 1985. as supplemented November 8.

1985, March 7 and October 10. 1986, Description ofamendment request:

The proposed amendments would revise Technical Specifications (TS) 6.5. 8.6, S.S and 6.10. concerning "Administrative Controls." The.proposed amendments would (1) seek to add the Superintendent of Integrated Scheduling to TS 8.5.1.3. 6.5.1.5. 8.6.1b. S.S.2. and 8.8.3c.. (2) seek to add the Superintendent ofStation Services to TS 6.5.1.8 and 8.8.1c., and (3) change the record retention period in TS 8.10.2 for records of quality assurance activities required by the QA Manual.

The effect of the first part would be to add the position of Superintendent of Integrated Scheduling to lhe list containing three other superintendents

'authorized to review and/or approve modifications of safety-related structures, systems or components {TS 8.5.1.3). proposed tests and experiments which affect nuclear safety and are not addressed in lhe FSAR or the Station Technical Specifications (TS 6.5.1.5),

Reportable Events (8.8.1b.), and procedures specified under Specification 8.8.1 and changes thereto (TS 6.8.2 and 8.8.3). as designated by the Station Manager.

By letter dated August 15. 1986. the NRC staff requested addilional information regarding the proposed amendments.

By letter dated October 10, 1988, the licensee stated that each of the superintendents has received Qualified Reviewer Training which includes review and approval of procedures and A..

i AaL iaaoa a.aaua ~ c waaailea va aaaad aaac requirements of 10 CFR 50.59. In addition. each of the superintendents meets the minimum qualifications specified by ANSI N18.1-1971.

The effect of thc second part would be to allow the Superintendent of Station Services to review and approve modifications to the station security program and related implementation procedures. Currently. only the Station Manager has this authority. In the" October 10, 1988. lelter, the licensee stated that the qualifications of the

'uperintendent of Station Services meet the minimum qualifications requirements of ANSI N18.1-1971 for comparable station positions.

Regarding the tliirdpari of the proposed amendments, TS 8.10.2 presently requires that the records of the quality assurance activities be retained for the duration of the Operating License. The proposed change would substitute a new TS 6.10.3 requiring that these recbrds be retained for the period specified by ANSI N45.2.9-1974, "Requirements for Collection. Storage, and Maintenance of Quality Assurance Records of Nuclear Power Plants."

A Federal Register Notice regarding parts (1) and (3) of the proposed amendments was published on )anuary

29. 1986 (51 FR 3714).

Basis forproposed no significant hazards consideration dcterminatioru The Commission has provided cerlain examples (51 FR 7744) of actions likely to involve no significant hazards

Federal Register / Vol. 52, N0. 38 / Thursday, February 26, 1987 / Noticee considerations. The request involved in this case does not match any of those examples. Iiowevcr, the staff has reviewed the licensee's request fol the above amendments and determined that should this request be implemented. It would not (llinvolve a significant increase in the probobilily or consequences of an accident previously evaluated bemuse the qualifications of

. the Superintendents of integrated Scheduling and Station Services meet arceptabie standards for comparable station positions. The proposed change to TS 6.10 would involve only the substitution of ti more specific and more appropriate requirement for QA record retention pursuant to o standard ocrrptcd by the NRC staff. Also, it would not (2) create the possibility of a new or different kind of accident from any occidrnt previously evaluated because the proposed changes would not affect the applicable accident analyses. Finally. it would not (3) involve a significant reduction in a margin of safety because the proposed addition of two superintendents would not represent a loss of review capability.

and would not affect the appiicable acridcnt analyses. The change to TS 6.10 would not shorten the relent!on period for those types of QA records which the Commission has determined should bc'retained for the plant lifetime, ond dors appropriately recognize that some of the QA record types have limited significance and moy be retained for Irsscr periods. Thus, the proposed change has no adverse impact on safety and docs not reduce the safety margin.

Accordingly. the Commission has determined that the above changes involve no significant hazards consideration.

Local Public Document Room loration: York County Library. 138 East Black Street, Rock Hill,South Carolina 29730.

rlttonrcy for licensee: hir. Albert Carr, Duke Power Company, 422 South Church Street. Charlotte. North Carolina 28242.

NRC Proiert Director: B.J.

Youngblood.

Duke Power Corupany, Dockats Nos. 50-269, 50-270 and 50-287. Ocouea Nucleaz Station, Units Nos. 1, 2 and 3, Oconoa County. South Carolina Date ofamendment request: February 12, 1988, as revised on October 10. 1986, and supplemcntcd on October 20, 1986 Description ofamendment request:

The proposed amendments would revise the Slotion's common Technical Specifications (TSs) to govern the operation and maintenance of the recently installed and operational containment hydrogen recombiner system (CHRS). TSs 3.16 and 4.42. on the hydrogen purge system. would be changed to include the CHRS as the primary method to maintain the post-accident amount ofhydrogen In the containment atmosphere below the control limitThe current Oconee TSe, 3.16 and 4.4.3 define the conditions necessary to assure the availability for the reactor building hydrogen purge system (RBHPSl as the means for the containment hydrogen control. The hydrogen purge system willbe available as a backup system.

Recombination of hydrogen and oxygen in the reactor building atmosphere is aa alternate and preferred means ofpost-accident hydrogen controL The CHRS has a greater capacity than the RBHPS. Also, use of the recombiner willnot increase offsite releases of radioactivity. The CHRS consists of two portable hydrogen recombiners, control panel for the recombiners and a portion of the penetration room ventilation system (PRVS). When needed, a recombiner willbe moved to the affected unit, anchored to its foundation. and connected by flexible metal piping to PRVS piping which runs to and from containment penetrations. The control panel willbe locally mounted near the recombiner.

The system functions by drawing air from thc reactor building into a heater in the recombiner. As the air temperature increases, the hydrogen in the air combines with oxygen to form water vapor. The mixture is then returned to the reactor building. The design flowrate is 90 standard cfm, and the design recombination efficiency is 95 percent for hydrogen concentrations greater than 0.5 volume percent.

Basis forptaposed no significant hozan& consideration determination-.

The Commission hos provided guidance concerning the application of the standards in 10 CFR 50.92 by providing certain examples (51 FR 7750). Example (ii)of the types of amendments not likely to involve significant hazards considerations ts an amendment that constitutes an additional limitation.

restriction, or control not presently included In the TSs: for example, a mare stringent surveillance requirement. "

The proposed TSs concern the use of hydrogen recombination as the method for controlling the post-accident reactor building hydrogen concentration.

SpeciAcally, the current TSs 3.16 and 4.4.3 provide requirements for operation and surveillance of the RBHPS es the method for post-accident hydrogen purging. The proposed TSs willpermit the use of CHRS as the primary method to control the post-accident reactor buildinq hydrogen concentrations. The CHRS has a greater capacity than the RPHPS.

Section 15.16 of the Oconee Nuclear Station Final Safety Analysis Report (FSAR) describes the post accident hydrogen control. The FSAR provides detailed evaluation of hydrogen recombination acceptability as a method for controlling the reactor building hydrogen concentrations. The analyses conclude that recombination is tbe preferred technique for containment hydrogen control compared to purging since it does not result in oifsite releases of radioactivity.

The proposed license amendments wiltimprove the margin of safety in that the application ofhydrogen recombination instead of purging as a'n

~

alternate method fol containmcfit hydrogen control provides additional protection since it does not result in offsite releases ofradioactivity.

Thus, the proposed license amendments appear to be encompassed by example (ii)ofamendments not likely to involve a significant hazards corsidera'.ion. On his basis. thc Commission proposes to determine that these amendmcnts do not involve significant hazards consideration.

Local Public Document Room location: Oconee County Library. 501 West Southbruad Street, Walholla, South Carolina 29891.

Attorneyforlicensee.

J. hiichael McGarry, IILBishop, Liberman, Cook.

Purcell and Reynolds, 1200 17th Street.

NW Washington. DC 20038 NRC Project Director: John F. Stolz Duke Power Company. Dockcts Now 50-269, 50-270 and 50-287, Oconee Nuclear Station, Units Nos. 1, 2 and 3. Oconee County, South Carolina Date vfamendmcnt request: August 13, 1986.

Description ofamc<idment request The proposed license amendmcnts would revise Figute? 3-1. "Protective System Maximum Allowable Sc tpoints,"

and Table 2.3-1, "Reactor Protective System Trip Setting Limils,"of the.

Station's common Technical

'SpcciTications (TSs) to increase,the reactor coolant system (RCS) liigh pressure reactor trip setpoint from 2300 psig to 2355 psig.

Basis forproposed no significant hazards consideration determination; Technical justiflication for increasing lhe RCS high pressure reactor trip setpoint is contained in Babcock 8< Wilcox (BhW) Owners Croup Topical Report BAW 1890, "Justificalion for Raising Setpoinl for Reactor Trip on High

Federal Register / Vol. 52, No. 38 / Thursday, February 26, 1987 / Notices Pressure." dated September 1985. In its April 22. 1986. Safely Evaluation of this topical report, the Commission r.onciudcd that it is acceptable to

ese thc high pressure reactor trip s< 'point for B&Wplants from 2300 psig io 235-psig while the power operated relic!

< <lve,PORV) setpoint remains at 2450 psig: that the setpoint change meets the roc'ements of NUREG-0737. Items II.)

ad II.K.3.7, regarding PORV opei... -'nd PORV caused small-break loss o<..

ant accidents (SBLOCA) and the requirements on this matter embodied in IE Bulletin 79-05B; and the topical report may be referenced in licensing submittals by BOW Owners Group men;bere.

The Commission has made a proposed determination that the proposed amendments involve no significant hazards considerations. Under the Commission's regulations in 10 CFR 50.92. this means that operation of the Station in accordance with the proposed amendments 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 total PORV openings per reactor year is negligibly changed by raising the setpoint for reactor trip on high pressure from 2300 psig to 2355 psig. However. as discussed in the Commission's Safety Evaluation of Topical Report BAW-1890. the probability of a SBLOCA caused by a stuckwpen PORV is within the WASH-1400 range of10

• to10

~

per reactor year. In addition, increasing the reactor trip<setpoint in accordance with the proposed amendments wiIInot

,.aller the consequences of a stuck-open RV. On these bases, the Commission finds that the proposed amendments v'ould not increase the probability or the consequences of an accident.

Since the design high pressure reactor trip setpoint is 2355 peig. the original Final Safety Analysis Report analyses remain applicable for the increased ectpoint. Raising the RCS high pressure trip setpoint would not, therefore. create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed setpoint change meets the requirements of NUREG-0737, Items II.K.3.2 and II.K.3.7 regaging PORV openings and PORV caused SBLOCA. as well as thc requirements on this matter embodied in IE Bulletin 79-OSB. In addition. the licensee states that the pn>posed change willresult in improved operational safci1 through a reduction in challenges ro safety systems. On these bases. the Commission find that the proposed change would not involve a significant reduction In a margin of safety.

Based on the above. the Commission proposes to find that the proposed license amendment request does not involve a significant hazards consideration.

Local Public Document Room location: Oconee County Library. 501 West Southbroad Street. Walhalla.

South Carolina 29691.

Attorneyforlicensee: J. Michael McGarry, III,Bishop. Liberman, Cook,,

Purcell and Reynolds. 1200 17th Street NW.. Washington, DC 20036.

NRC Project Director. John F. Stolz.

Duquesne Light Company, Docket No.

50-$34, Beaver Valley Power Station, Unit No. 1, Shippingport, Pennsylvania Date ofomendment request: January 15, 1987.

Description ofamendment requestr The proposed amendment would add a temporary note to Technical Specification 3 7.7.1 to allow the Control Room Emergency Ventilation System (CREVS) to be temporarily removed from service for up to 60 days. This request is being made to allow CREVS modifications. which willincrease the capability of the system, and to remove an air-tight partition of steel plates which currently separates the Beaver Valley Unit 1 and Unit 2 portions of the control room.

Basis forproposed no significant irazards cansideratian determinatiom The proposed amendment is a temporary revision which would provide an alternate means of emergency control room pressurization during that time period when the Beaver Valley Unit 1 CREVS is out of service for equipment modifications and to allow removal of the partition that currently physically separates the Unit 1 and Unit 2 portions of the control room. The Beaver Valley Unit 2 CREVS is sized to provide the required ventilation system capacity for both control room areas. During the temporary period when the Beaver Valley 1 CREVS is out of service, control room habitability willbe maintained by the Unit 2 CREVS, and the required support systems willalso be avaOable (including the applicable portions of the emergency power systems). The probability of occurrence or "consequences of accidents (FSAR Chapter 15 accidents and offsite accidents involving chemical releases) previously evaluated willnot be significantly increased. The implementation of tne requested

'mendment does not create the possibility of a new or different kind of accident from any accident previously evaluated because existing systems are only being modified to increase their capability. and during the modification.

Unit 2 systems willprovide the needed service. Furthermqre. the amendment willnot involve a significant reduction In the margin of safety because equivalent means of control roon<

habitability and pressurization willbe provided. Therefore. the staff proposes to determine that the amendment involves a no significant hazards consideration.

Local Public Document Room location: B.F. Jones Memorial Library, 683 Franklin Avenue, Aliquippa.

Pennsylvania 15001.

Attorney for licensee: Gerald Charnoff, Esquire. Jay E. Silberg.

Esquire, Shaw, Pittman, Potts, and Trowbridge, 2300 N Street NW.,

Washington, DC 20037.

NRC Project Dr'rectorr Lester S.

Rubenstein.

Florida Power and Light Company.

Docket Nos. 50-250 and 50-251, Turkey Point Plant Umts 3 and 4, Dade County, Florida Date of amendment request:

December 19. 1986.

Description ofamendment request:

The proposed change adds a license condition to require implementation of Florida Power 8< Light Company's (FPL)

Plan for the Integrated Scheduling of Plant Modifications for Turkey Point Unite 3 and 4 (the Plan). The Plan describes the responsibilities of FPL and the Nuclear Regulatory Commission (NRC); provides a summary description of the FPL Integrated Schedule Program:

describes the Integrated Schedules; describce the mechanisms for changing the Integrated Schedules: describes the periodic'reporting requirements: and describes the mechanism for changing the Plan.

The Integrated Schedule program will enable FPL to effectively manage implementation of modifications which have been required or proposed by the NRC, as well as other measures which have been identified by FPL and which willensure the continued safe. prudent, reliable, and economic operation of the Turkey Point Plant.

This program was developed to coordinate and schedule all necessary work at the Turkey Point Plant, whether mandated by NRC or identified by FPL and others. The program objectives are to: (1) Conform to regulatory requirements:

(2) provide sufficient lead times for modifications: (3) minimize changes for operators: (4) assure training requirements are fulfilled:(5) effective)y

Federal Register / Vol. 52, No. 38 / Thursday, February 26, 1987 / Notices manage financia and human resources; end (8) specify the framework for changes to developed schedules, The program reflects that fiscal and manpower resources are finite and that a limiton the onsite manpower ls necessary. The program integrates aII'resently planned work at Turkey Point over a nominal five year period to ensure thatindividualtasks are effectively scheduled and coordinated. It i provides a means for new requirements

" to be accommodated taking into account schedule and resource constraints.

The proposed changes to the Turkey Point Technical Specifications are: for Facility Operating License DPR-31. the license willbe amended by adding a new license condition. Condition K, to read as follows:

K. Int<<gru t<<d Schedule

1. The plan for the integrated Scheduling of Plant Mnditicstiuns for Turkey Point Unite 3 and 4 (ihc Plan). submitted on Dcccuibcr19.

1988 is approved.

a. Thc Plan shall be followed by the licensee from cnd after the clfcciivc date of this emcndmcnt.

b. Changes to dates for completion of items idcntiflcd in Schedule B do not require a license amendment. Dates epcciflcd in Schedule A shall bc changed only in eccordencc with applicable NRC proccdurce.

2. This license amendment shall be effective December 31. 1989, subject to renewal upon application by the licensee.

For Facility Operating License DPR-41 ~ the license willbe emended by adding e new license condition.

Condition J. to read as follows:

l, Int<<gtu ted Sch<<dule

1. The plan for the integrated Scheduling of Plant h1odificstione for Turkey Point Units 3 end 4 iihc Plan). submitted on December 19.

i988 is approved.

c. The Plan shell be followed by the licensee from and after the effective date of this amendment.

b. Changes to dales for completion of items identified in Schedule B do noi require a license amendment. Dates specified in Schedule A chait be changed only in eccordancc with applicable NRC proccdurce.

2. This license amendment shall bc cifcciivc December 31. 1989. subject to ccncwat upon application by ihc ticcnecc."

Basis forproposed no signi%'cant hazards consideration determinatiom The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facility involves no signiflcant hazards consideration ifoperalion of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase In the probability or consequences or an accident previously evaluated; or (2) create the possibility of a new or different kind ofaccident from an accident previously evaluated: or (3) involve a signiflcant reduction ln a margin of safety.

The proposed amendments Implement an administrative program to provide for

~chedullng changes and notification of scheduling changes, and does not affect in any way the design or operation of the plant. The program would enhance plant safety bv inore effectively

'ontrolling the number and scheduling ofplant modiflcatlons. thereby ensuring that issues required for safe operation of the plant receive priority and are completed in a timely manner.

In addition, the amendment ls consistent with the NRC guidance in Generic Letter 83-20 dated May 9. 1983 and Generic Letter 85-07 dated May 2, 1985.

Since the proposed amendments will not affect the existing design nor result in any changes to the operational limitation of this plant, the staff proposes to determine that the amendments do not involve a significant increase in the probability or consequences ofan accident from any accident previously"evaluated; do not create the possibility of a new or different kind of accident previously evaluated; and do not involve a significant reduction in a margin of safety. The staff. therefore. proposes to determine that the amendments do not involve a significant hazards consideration.

Local Public Document Room lacatiom Environmental and Urban Affairs Library. Florida International University, Miami. Florida 33199.

Attorneyforlicensee: Harold F. Rels, Esquire, Newman and Holtzer. P.C.. 1615 L Street NW.. Washington, DC 20036 NRC Project Director. Laster S.

Rubens tein.

Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authonty of Georgia, City of Dalton, Georgia, Docket No, 50-321, Edwin L Hatch Nuclear Plant, Unit No. 1, Appllng County, Georgia Date ofamendment request:

September 9. 1988.

Description ofamendment request:

This amendment would modify the Technical Specifications (TS) to clarify the surveillance requirements for refueling interlocks by (a) stating that the reactor mode switch refueling interlocks shall be functionally tested "prior to core alterations" rather than the current "prior to any fuel handling with the head off the reactor vessel" and (b) adding surveillance requirement discussions that specifically address the fuel grapple hoist load setting interlock I

and the auxiliary hoist load setting-interlock.

Basis forproposed no significant hazard! consideration determinatioru The Commission has provided guidance for the application of criteria in 10 CFR 50.92. Providing example of amendments that ere consfdered not likely to Involve a significant hazards consideratiun (51 FR 7751). One such example is (i) a purely administrative change to technical specifications.

The proposed modiTication consists of changes in wording and format that clarify the original wording but do not change the original intent of the surveillance requirements. Therefore since there have been no changes In the requirements the proposed change is an editorial change and is similar to.

example (i).

On this basis the Commission has made a proposed determination that the amendment application does not involve a significant hazards consideration.

Loca/ Public Document Room location: Appling County. Public Library,-

. 301 City Hall Drive. Baxley, Georgia.

Attorneyforlicensee: Bruce W.

ChurchilL Esquire. Shaw. Pittman, Potts and Trowbridge, 2300 N Street NW.,

Washington, DC 20037.

NRC Project Director. Daniel R.

Muller.

Georgia Power Company. Oglcthorpe Power Corporation. Municipal Electric Authorityof Georgia. City of Dalton, Georgia, Dockets Nos. 50-321 and 50-368, Edwin L Hatch Nudear Plant, Units No@ 1 and 2, Appling County, George Dote ofamendment request: August 25, 1988. supplemented January 23. 1987.

Description ofamendment request:

The amendment would modify the Technical Specification (TS) to change the high room temperature setpoint that automatically close" the Reactor Water Cleanup Unit (RWCU) isolation valves in the event of a leak or pipe break'nside the RWCU rooms. It would increase this setpoint from 120' to 150' in order to eliininate some of the unnecessary valve isolations that occur due to high RWCU room temperature that are not related to pipe breaks or leaks in the RWCU piping.

Basis forproposed no significant

'azards consideration determination:

The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation of the facility in accordance with the proposed amendment would not: (1) Involve a

5856 Federal Roglster / Vo). 52, No. 38 / Thursday, February 28, 1987 / Notices 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 signilicant reduction in a margin of safety.

The proposed change involves an increase in the RWCU room temperature setpoint at which the RWCU isolation valves would automatically close ifa pipe brcak or leak occurred in the RWCU room. The RWCU pipe break accident was evaluated m Section 15A.5.5 of the Hotch Unit 2 FSAR. The proposed change does not affect the cause of a pipe brcak and therefore.

does not affect the probability of the pipe break accident. The increase in the temperature setpoint willhave essentially no affect on the consequences of a full pipe break in the RWCU room because the increase in the temperature would be so rapid that by the time it reach 150'F or the time it takes to reach 120'F is insignificant compared to the time it takes to close the isolation valves after the setpoint is reached.

The smallest leak rate requiring automatic actuation of the isolation valves is 25'allons per minute. The difference in consequences of an accident involving this 25 gallon per minute leak with a 150'F as compared to a 120'F setpoint amounts to an increase in total leakage into the RWCU prior to isolation of the RWCU of a few hundred gallons of water. This is insignificant with respect to the 7500 gallons (approximately) water release estimated in the full pipe brcak accident analysis.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because the physical features

,gf the plant are not altered and no new modes of operation are created. The postiilated effects of this change cire bounded by the existing Accident Analysis of the RWCU pipe break as discussed in the Hatch Unit 2 FSAR.

The proposed change does not involve a significant reduction in the margin of safety because (1) the system continues to isolate the postulated leaks and breaks with little water released prior to isolation compared to the design break analysis. as discussed in Section 15.A.5.5 of the Hatch Unit 2 FSAR, (2) the environmental qualification of the affected components remains valid. and (3) allowing the RWCU to operate with a higher RWCU room temperature will prevent some unnecessary forced utages of the RWCU that are currently caused by the existing lower temperature setpoint and the resultant automatic isolation of the RWCU.

On the basis of the above, the Commission has determined that the requested amendments meet the three standards and therefore has made a proposed determination that the amendment application does not involve a significant haxards consideration.

Local Public Document Room location: Appling County Public Library, 301 City Hall Drive, Baxley. Georgia.

Attorneyforlicensee: Bruce W.

ChurchilL Esquire, Shaw. Pittman, Potts and Trowbridge, 2300 N Street NW..

Washington. DC 20037.

NRC Project Director. Daniel R.

Muller.

GPU NucIear Corporation, et alDocket No. 50-289, 'Haec MileIsland Nuciear Station. Unit No. 1, Dauphin County, Pennsylvania.

Dole ofamendment reguest: January 28, 1987.

Description ofamendment reguest:

The TMI-1Fire Detection System in the Auxiliary,Fuel Handling. Intermediate and Control Buildings. end the Fire Suppression System in the Auxiliary Building ae being expanded to satisfy the commitments made concerning 10 CFR Part 50, Apperidix R, arid approved in NRC Safety Evaluations dated June 4, 19M, and December 30. 1988. The amendment would revise the TM-1 Technical SpeciTications (TSs) to include expansions of existing fire detection systems to cover new areas.

Additionally. it would allow converting a manually actuated deluge water spray

'ystem to an automatically actuated sprinkler system in the Auxiliary Building containment penetration area and would allow installing a water curtain system in one location. The amendment would also correct a typographical error in the existing TSs.

Basis forproposed no significant hazards consideration determination: A

p. posed "mendincnt tc an opcratLag license for a facilityinvolves no significant hazards considerations ifthe three standards of 10 CFR 50.92(c) are met. Pursuant to the provisions of 10 CFR 50.91. the licensee has provided an analysis ofno significant hazards considerations using the Commission's

- standards.'Ice Commission's staff agrees with the licensee's analysis. Each standard of10 CFR 50.92(c) is discussed in turn.

Standard 1. Operation of the facilityin accordance with the proposed amendment should not involve a significant increase in the probability of occurrence or consequences ofan accident previously evaluated. The design basis event related to this change is a fire in the affected fire zones. The proposed amendment has no effect on the probability of occurrence of this design basis event. The potential consequences of a fire in the affected fire xones are reduced because the proposed change provides additional assurance that the fire detection instrumentation and suppression provided in each fire zone are operable and therefore capable of promptly detecting. confining and extinguishing the fire.

Standard 2. Operation of the facilityin accordance with the proposed amendment should not create the possibility of a new or different kind of accident from any accident previously evaluated. The design basis event related to this change is a fire in the affected fire zones. The proposed amendment has no effect on the possibility of creating a new or different kind ofaccident from any accident previously evaluated. The proposed amendment provides additional assurance that the fire detection instrumentation and suppression systems installed to satisfy 10 CFR Part 50, Appendix R, requirements are operable and are unrelated to the possibility of creating a new or different king of accident.

Standard 3. Operation of the facilityin accordance with the proposed amendment should not involve a signiTicant re'duction in a margin of safety. The proposed amendment provides for the control of the additional fire detection instrumentation and suppression installed to satisfy 10 CFR Part 50, Appendix R, requirements. The proposed amendment also increases the minimum number ofdetectors required to be operable in the Control Room

~

Cabinets and the Control Room Area in Control Building Elevation 355'.

Therefore, the overall margin of safety for the plant is increased.

The typographical error being corrected concerns the incorrect spelling of a single word and is purely administrative in nature. This change is considered to fall within the scope of example (i) of amendments that are not considered likely to involve significant hazards considerations contained in the Commission's guidelines published within 51 FR 7751.

Accordingly. based on the above discussion. the Commission proposes to determine that the application for amendment does not involve significant hazards considerations.

Local Public Document Room location: Government Publications Section. State Library of Pennsylvania.

Education Building. Commonwealth and Walnut Stree)s. Harrisburg.

Pennsylvania 17128.

Federal Register / Vol. 52No. 38 / Thursday, February 28, 19Sl./ Notices Attorney forlicensee: Ernest L Blake, Jr. ~ Shaw, Pittman, Potts and Trowbridge, 2300 N Street, N.W.,

, Washington, DC 2003?.

NRC Pleat Directarr John P. Stolz.

Indiana and Michigan Electric Company, Docket Nos. 50-$ 15 and 50-318, Donald C. Cook Nuclear Plant. Unit Nos. 1 and

', Berrlen County, Michigan Date af amendment request: January

18. lee.

Description ofamendment request:

The proposed amendment would revise the Technical Specifications for the emergency diesel generators to improve and maintain reliability (Per Generic Letter &t-15 issued july 2.1984), change a number of related Technical Specilications ta improve clarity and correct errors, and revise the emergency battery loads testing. to allow simulated connected loads during tests.

Basis forproposed na significant hazards consideratian determinatiam The Commission's standard for determining whether a significant hazards consideration exists is as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facility

,involves no significant hazards consideration ifoperation 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; (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 sai'ety.

Generic Letter 84-15 on the subject

'Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability."

established new requirements that would reduce the risk ofcore damage

, from station blackout events by. among other things. changes to the technical specifications which support a desired diesel generator reliabilitygoal. The licensee proposes to adopt many of the technical specification changes which were determined by the NRR as risk reduction actions. The proposed changes. therefore. should reduce the probabilities and consequences of accidents previously analyzed and should increase the margin of safety.

The proposed changes willnot place the plant in a new or unanalyzed conditian, therefore. the changes willnot create a new or.different kind of accident from any previously analyzed.

. The licerisee also4!roposes to change a number of related'Technical Specifications to improve clarity and correct errors. Several changes were maite to reflect plant design and to make the iwo Units'echnical Specifications alike. There are also a number of editorial changes and error corrections.

Allof these changes are administrative in nature and do not change the probabilities or consequences of any previously analyzed accidents. The changes reflect plant design similarities and correct errors, therefore. there ls no change to plant operation which would result in a new or d!fferent kind of accidenb The corrections and clariflcations willnot result in a reduction in any margin of safety.

In amendments 88 and 72 for Unit Nos. 1 and 2 respectively, the licensee was granted approval to test the battery capacities with simulated loads using a load bank in place of the actual loads using the static inverters. The surveillance requirement being changed is to determine the condition of the battery; a separate surveillance test is required for determining the performance discharge through actual battery loads. The use of a load bank which simulates actual loads should not affect the test nor would it significantly increase the probabilities or consequences of any previously analyzed accident. Whfle the battery ls connected to the load bank during testing, the battery cannot affect other systems or components which are required to be operable. Therefore. the change would not create a new or different kind of accident from any previously analyzed. The batteries will continue to be capacity tested on the same frequency: only the method of loading the batteries is changed. Since the change willnot impact the batteries in the modes when the batteries are required to be operable, the change will not affect the ability of the batteries to perform their safety function. Therefore.

the proposed change willnot involve a significant reduction in a margin of safety.

On the basis of the above consideration, the staff proposes to find that the changes do not involve a significant hazards consideration.

LocalPublic 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 N Street, NW.,

Washington, DC 20038.

NRC Project Director. B.J.

Youngblood.

lowe EIectric Light and Power Company Docket No. 50-331, Duane Arnold Energy Center, Linn County, lowe Dale ofamendment requesL October

14. 1988.

Description ofamendment request:

The Iowa Electric Light and Power Company (the licensee) proposes to change the Technical Specifications for the Duane Arnold Energy Center to (1) eliminate the requirement far OperationaCommittee review of routine procedure changes that do not affect safety related equipment design bases or the safety objectives of administrative controls by substituting a screening process, and (2) invoke minor administrative revisions to reflect changes in nomenclature.

Bcsis forproposed no significant hazards consideration determinatiom The Commission has provided standards (10 CFR 50.92(c)) for determining whether a significant

'azards consideration exists. A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration if operation of the facilityin 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 cf acc! den! rrom any accident pre~iousiy evaluated: or (3) involve a significant reduction in a margin of safety.

The licensee has "provided an analysis of each of the above criteria for the amendment request as follows:

In reviewing this proposed request for Technical Sptciltcstisn change we have concluded ibst this amendment:

(1) does not invotvt s significant increase in the probability or consequences of sn accident previously evaluated because there willbe no change in the review procedures for these changes which affect ssitiy.related equipment design bases or asftiy objectives of administrative controls. Tht word changes noted above are!o achieve consiattncy in nomenclature snd, therefore. do not invotvt s significant incrtsat in the probability or consequences of sn accident prtviously evaluated.

~

(2) does not crasis tht passibility of s new

, or difftrsnt kind of accident because the uss of s senior Iictnatd operator snd rntmbtr ol the plant supervisory staff or s plant supervisory staff member who poaassata s

atniar optrstor' license io review rouunt procedure changes willprovide s comprehensive review of plant procedures snd their effect on ayattm interrelationships.

For those procedure chsngta which effect asftty-rtisttd equipment design bases or'iht asftiy objtciivta of administrative controls, review snd spprovsi by the Operations

,Coiiunitttt willstill be required. Tht word changes which are needed io schitvs consistency in nomenclature are administrative tn nature snd, thtrsfort. do not crests the possibility of s new or ditftrtnt kind of accident.

(3) dots no! Involve s significant reduction in ~ margin of safety because changes to

5858 Federal Register / Vol. 52, No. 38 / Thursday, February 2B. 1987 / Notices existing procedures which are identified to potentially aileci the safety design bases of equipment or safety objectives willnot only have iin iidditionai level of review (a senior In enseit operator snd s supervisory plant staff member who possesses a senior operator's bcense). but also review by the Operations. Committee. This willresult in an equis alent or more strinttent review process as compared io the existing process.

Based on an evaluation of the above licensee analysis. the staff has made a proposed determination that the proposed amendment involves no sic".:ficnnt'hazards consideration.

veal Public Document Room iucali'on

Cedar Rapids Pub)ic Library.

500 First Street, S.F, Cedar Rapids, Iowa 52401.

Attorney%r licensee: jack Newman, Fsquirc, Kathiecn H. Shca, E:q:.:r:.

Ncwmati and Holtzingcr. 1615 L S'- cct, NW., Washington. DC 20036.

IVllCProject Director. Daniel R.

Muller.

iowa Electric Light and Power Company, Docket No. 50-331, Duane Arnold Energy Center, Linn County, iowa Dole ofamcndmenl request:

November 26, 1986.

Descriplian ofamendmenl requests Thc proposed license amendment would revise Duane Arnold Energy Center (DAEC) Technical Specification Section 3.7/4.7, which requires Standby Gas Treatment System (SGTS) surveillance testing at a flowrate of 4,000 cfm. The proposed changes would allow those SGTS tests thai are required to be performed at the design flowratc to be performed in the flowrate range of design flow minus 10% (3,600 cfm) to design flow (4.000 cfm).

Basis forproposed no signi%'cant hazards considcralion dclcrminalionr The Commission has provided standards (10 CFR 50.92(c)) for dpterinioing whether a significant hazards consideration exists. A proposed amendment to an opcrat)ng license for a facilityinvolves no significant hazards consideration If operation of the facilityin accordance with the proposed amendment would not (1) involve a significant increase ln 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 licensee has provided an analysis of each of the above criteria for the cmendmcnt request as follows:

In reviewing this proposed request for Technical Specification chentte we have concluded that this amendment:

it I diiei noi involve a ~ignilicani increase tn the probabil:iy or consequences of an accident previously evaluated because the Standby Gas Treatinent System willstill perform its design functions and willbc demonstrated to do so using this ilowrate range. These tests willbe performed using the same acceptance criteria for Inplace cold jdioctylphthlate) DOP and halogenated hydrocarbon removal and maintaining a minimum Ila inch went. negative pressure in secondary containment currently in the DAEC Technics) Specifications.

lz) does not create the possibility of a new or different kind of accident because It Involves no modiiications to the plant and affects only the testing of the Standby Gas Treatment System. The changes to the testing willnot affect the perfonnance or the design basis of the Standby Gas Treatment System as described in the updated FSAL (3) does not involve a ~ignificent reduction In a margin of safety because performing the revised Technical SpeciTication Surveillance Requirements at an SGTS train flowneer but below 4.000 cfm le.g., between 3,000 and 4.000 cfm) is consistent with the design basis of the system. The SGTS flowcontrol system will be tested to meet the same iodine retention requirement and maintain How below 4.000 cfm and meet the 1/4 Inch w.g. negative pressure requireinent ofTechnical Speci!ication Surveillance 4.7.C.l.c.

Therefore. this proposed license amendment is judged to involve no significant hazards consideration.

Based on an evaluation of the above licensee analysis. the staff has made a proposed determination that the proposed amendment involves no significant hazards consideration.

Local Public Document Room localionr Cedar Rapids Public Library.

500 First Street, S. E. Cedar Rapids, iowa 52401.

Atlarneyforlicensees Jack Newman, Esquire. Kathleen H. Shee, Esquire, Newman and Holtzingcr, 1815 L Street.

N W, Washington, DC 20036.

NRC Project Director. Daniel R.

Muller.

Louisiana Power and Light Company, Docket No. 50-382. Waterford Steam Elcctnc Station, Unit 3, St. Charles Paris. Louisiana.

Dale ofamendmenl request: January 13, 1987.

Descriplian afamendment raquest:

Louisiana Power and Light Company (LP8L) has proposed the addition of new technical spec) ficat)on (No. 3.3.3.7.3) for the Broad Range Toxic Gne Detection System (BRTGDS) to satisfy Condition 2.C.4, "Broad Range Toxic Gaa Detectors." ofFacility Operating License No. NPF-38 and Supplement No, 6 to the Safety Evaluation Report (NUREG-0787).

Waterford 3 was allowed to operate during the first cycle without a BRTGDs system due to near-term compensatory measures l.e.. periodic surveys of toxic gas inventories. a hot line communication with the St. Charles Parish Emergency Operations Center, a control room operator and plant personnel training program, and procedures with respect to response to toxic gases. Although these measures were considered aBcquatc for short-term operation, an additional level of protection willbe provided by the BRTGDS for operation over the expected plant lifetime.

The proposed BRTGDS would include two redundant photoionization detectors. These detectors would monitor the atmosphere in the reactor auxiliary building outside air intake duct. Whenever the concentration of the detectable gases exceeds 0 preset limit, these detectors willeach induce an electric current in the associated circuit through photoionization, thus generating e signal. This signal automatically switches the control room air conditioning system to the isolation mode ofoperation.

The Waterford Steam Electric Station.

Unit 3 control room operators are presently protected from accidental releases of chlorine and ammonia by separate chlorine and ammonia detectors which monitor the atmosphere in the outside air intake duct. The detectors sound an alarm and iso)ate the control room ifthe concentration of either of these gases exceeds its respective preset )Im)t.

In addition, LPhL is a participant in the St. Charles Parish emergency hot line. which can be expected to alert the control room within five minutes of being notiTied ofany serious release of hazardous chemicals in the area, thus allowing the operators to take immediate protective action. The BRTGDS willthus supplement the present defense. in-depth toxic chemical protective measures and provide additional toxic chemical protection.

This new technical specification will require two independent broad range gas detection systems to be operable with their alarm/trip sctpoints adjusted to actuate at the lowest achievable

)DLH (immediately Dangerous to Life and Health) gas concentration level of detectable toxic gases. They are also required to provide reliable operation.

This proposed technical specification is to be applicable in all modes of operation. Should the above requirements not be met, one of the following actions willbe required. When only one broad range gas detection system is operable, the inoperable system must be restored to operable status within seven days. or within the next six hours the plant must initiate and maintain operation of the control

Federal Register / Uol. 52, No. 38 / Thursday. February 26; 1987 / Notices Sash room ventilation system in the recirculation mode of operation. When both systems are inoperable, the control room ventilation system must be initiated and maintained in the recirculation mode of operation within one hour, The provisions of Speciiication 3.0.4 (Entry into an Operalional Mode) are nol applicable to this proposed technical specification.

The proposed surveillance requirements for this technical specification demand that each broad range gas detection system be demonstrated operable by a performance of a channel check at least onc'e every twelve hours. a channel functional test at least once every 31 days. and a channel calibration at least once every seven days.

LphL has proposed that the Bases (3/

4.3.3.7) for this section be amended to include lhe BRTGDS wilh the other chemical detection systems. It would state that the operability of the BRTGDS shall ensure that sufficient capability is available to promptly detect and initiate protective action in the event of an accidental chemical release. The amended Bases would also provide further definition of the system and indicate that the setpoint is based on testing and operating experience as a result of the Waterford 3 area chemical atmosphere-profile. Due to possible environmental conditions, this setpoint is subjec't to change and therefore would be established and controlled by procedure.

Basis for ProposedNo Significant ilazards Considerations Determination:

The NRC'staff proposes to determine that the proposed change does not involve a significant hazards consideration, because,'s required by the criteria of 10 CFR 50.92(c). operation of the facilityin 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 reduclion in the margin of safety. The basis for this proposed finding is given below.

(1)-The proposed addition has no effect on the assumptions contained in the safely analyses. The technical

'specificalions which preserve the safety analysis assumptions are likewise unaffected by the proposed change. The Waterford 3 progrem for the protection of control room operators is comprised of various in-depth defensive measures.

The control room operators are prolected from accidental releases of chlorine and ammonia by dedicated chlorine and ammonia detectors, which monitor the atmosphere in the outside air intake duct and willsound an alarm and isolate the control room ifthe concentration of either of these gases exceeds its respective preset limit.In addition. LP8 Lis a participant in the St.

Charles Parish emergency hot line.

which can be expected to alert the control room within five minutes of being notified of any serious release of hazardous chemicals ln the area, thus allowing the operators to take immediate protective action. The proposed BRTGDS willsupplement the present defense-in-depth toxic chemical protective measures and willtherefore not increase the probability or consequences of any accident previously analyzed.

(2) The proposed change willnot physically affect existing chlorine and ammonia syslems. Therefore. the proposed change willnot create the possibility of a new or different kind of accidenL (3) The Waterford 3 safety limits and margins are defined and maintained by the Technical Specifications. This proposed technical specification is an additional limitation on plant operation and does not affect the safety limits or margins contained in the other Waterford 3 technical specifications.

The proposed change willtherefore not involve any reduction in a safety margin.

The Coinmission has provided guidance concerning the application of standards for determining whether a significant hazards consideration exists by providing certain examples (51 CFR Part 7751) of amendments that are considered not likely to involve significant hazards considerations.

Example (ii) relates to a change that constitutes an additional limitation.

restriction, or control not presently included in the technical specifications.

In this case, the proposed change in similar to Example (ii) in that the BRTGDS has not previously had a technical specification associated with it.

As the changes requested by the, licensee's January 13, 1987 submittal fit Example (ii) as well as satisfy the criteria of 50.92. the staff proposes to determine that the proposed changes do not involve a significant hazards consideration.

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

Attorney far licensee: Bruce W.

Churchill, Esq.. Shaw, Pittman, Potts and Trowbridge, 2300 N St. NW..

Washington. DC 20037.

NRC Pmgect Dtrectar. George W.

Knighton.

Mississippi Power 5 Light Company, System Energy Resources, Inc., South Mississippi Electric Power Association, Docket No. 50-418, Grind Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi Dote ofamendment request: May 22.

1988 as revised December 9, 1988 and January 29, 1987.

Description ofamendment request:

The proposed amendment would change the Technical Specifications (TSs) for alternating current electrical power systems (TS 3/4.8.1) to reduce excessive testing of the three onsite emergency diesel generators (diesel generators 11.

12 and 13) consistent with the recommendations provided in lhe Commission's Generic Letter 84-se "Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability."

July 2, 1984. The specific changes to the Technical Specifications would include the followingchanges to Limiting Conditions for Operation in TS 3.8.1.1:

(1) Action Statement a. (actions to be taken upon declaring either one of the two required offsite circuits or diesel generator 11 or 12 inoperable) would be divided into two action statements; Action Statement a, for one offsite circuit inoperable and Action Statement

b. for either diesel generator 11 or 12 inoperable. Proposed'Action Statements a and b would require the performance of Surveillance Requirement 4.8.1.1.2.a.4 (verifying operability of the remaining diesel generators) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of declaring the offsite circuit or diesel generator inoperable. The present requirement is to perform the surveillance within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. Also, the time allowed for restoration to operable status (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) is clarilied by specifying that inoperable equipment willbe 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 /> "from time of initial loss."

(2) Action Statement b. (actions to be taken upon declaring one offsite circuit and either diesel generator 11 or 12 inoperable) would be designated Action Statement c and would be changed to require the performance of Surveillance

'equirement 4.8.1.1.2.a.4 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

The present requirement is to perform this surveillance testing within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

Also. the statement regarding time allowed for restoration of all required offsite circuits and diesel generators to operable status (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) would be deleted because it would be redundant

~

'ederal Register / Vol. 52, No. 38 / Thursday. February

26. 1987 / Notices to similar statements in proposed Action Statements a and b.

(3) Action Statement c would be designated Action Statement d and only the references to other action statements would be changed to be consistent with the proposed TS.

(4) Action Statement d (actions to be taken upon declaring two oflsite power circuits inoperable) would be designated Action Statement e and would be changed to require performance of Surveillance Requirement 4.8.1.12.a.4 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The present requirement is to perform this surveillance testing within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. The time allowed for restoration of both offsite power circuits to operable status would be deleted because it would be redundant to a similar statement in proposed Action Statcrnenl a.

(5) Action Statement e (actions to be taken upon declaring diesel generators 1t and 12 inoperable) would be designated Action Statement fand would be changed by deleting the present requirement to perform Surveillance Requirement 4.8.1.1.2.a.4 for diesel generator 13 within two hours from the time of initial loss and at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter. Performance of Surveillance Requirement 4.8.1.1.2.a.4 for diesel generator 13 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the restoration to operability of either diesel generator 11 or 12 would be a condition for continued operation.

(6) Action Statement f (actions to be taken upon declaring diesel generator 13 inoperable) would be designated action statement g and would be changed to require the performance of Surveillance Requirement 4.8.1.1.2.a.4 for diesel generators 11 and 12 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The

.present requirement is to perform this surveillance testing within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and etlcast once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter.

Additional changes would be made to the Surveillance Requirements in TS 4.8.'1.1 hs follows:

(7) Surveillance Requirement 4.8.1.1.2.a.4 (testing required to demonstrate diesel generator starting capability) would be revised by changing the term for initial testing conditions from "ambient" to "standby" and by adding a footnote to specify allowable initial conditions lor testing.

The proposed footnote would allow all diesel generator starts to be preceded by an engine prclube period. The footnote would also allow surveillance testing to be preceded by v4rmup procedures except that a fast cold start from standby conditions would be required once per 184 days. The present requirement is to perform all testing from ambient (standby) conditions.

(8) Surveillance Requirement 4.8.1.1.2.a.5 (testing required to demonstrate diesel generator loading capability) would be revised by adding a footnote to specify allowable loading conditions for testing. The footnote would allow gradual loading during surveillance testing except that fast loading (within 60 seconds) would be required once per 184 days. The present requirement is to perform all testing with fast loading.

(9) Table 4.8.1.1.2-1 (diesel generator test frequency as a function ofvalid teat failures) would be revised to: (a) make the maximum test frequency once per 7 days instead ol the present maximum frequency of once per 3 days: (b) base the determination ofnumber of test failures on a per diesel generator basis instead of the present requirement to base the number of failures on a per nuclear unit basis; (c) clarify the starting date for counting valid tests by adding the OL issuance date; and, (d) clarify the use of the table by deleting the statement "Entry in this test schedule shall be made at the 31 day test frequency."

The TS Bases 3/4.8.1 (alternating current sources) would be changed to reflect the proposed changes to TS 3/

4.8.1 regarding a reduced maximum test frequency and determination ofnumber of test failures on a per diesel generator basis rather than a per nuclear unit basis.

Bosis forproposed no signi%'cant hozards considero lion determinolionr 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 facilityinvolves no significant hazards considerations ifoperation 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 dilferent kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

The licensee has provided an analysis ofsignificant hazards considerations in its request for a license amendment. The licensee bas concluded with appropriate bases. that the proposed amendment satisfies the three standards in 10 CFR 50.92 and. therefore. involves no significant hazards considerations. The NRC staff has made a preliminary review of the licensee's submittal. A summary of staffs review follows.

The objective of diesel generator (DG) periodic surveillance testing is to meet the reliabilitygoals ofRegulatory Guide 1.108 "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants." Such surveillance testing provides assurance of the availability of the diesel generators in mitigating various transients and postulated accidents following a loss of offsite power. The required reliabilityis demonstrated with more frequent testing as the number of test failures increases.

Standard Technical Speciiications require that the DGs be tested in accordance with Regulatory Guide 1.108 where the test interval depends on the demonstrated DG performance and the test interval is established on a per nuclear unit basis.

As described in Generic Letter 84-15.

the staff has for sometime been evaluating the frequency of DG testing and the associated potential lor severe degradation of engine parts. The staff concluded that this test frequency can be reduced to minimize this potential.

Generic Letter 84-15 identified the followingmeans by which periodic surveillance testing of diesel generators can be changed to increase availability of onsite electric power without signilicantly affecting diesel generator reliability:

(a) testing ofremaining operable diesel generators followingfailure of an offsite power circuit or diesel generator can be reduced from a starting test of each diesel generator every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to one test of each diesel generator.

(b) testing of all diesel generators based on the number of failures of all diesel generators for one nuclear unit can be changed to testing of the restored diesel generator based on the number of failures of that diesel generator, (c) the test frequency for an Individual diesel generator based on the number of its failures can be reduced from the present maximum of once every three days to a maximum of once every seven days; and, (d) the number of cold fast starts during surveillance testing can be reduced by using engine prelubrication and warmup procedures recommended by the diesel engine manufacturer to reduce engine stress and wear for most tests.

Changes (1) ~ (2). (4), (5), and (6) to the Technical Specifications would reduce the number of tests ofremaining operable diesel generators following loss of an alternating current power source, thus reducing wear of the diesel generators and consequently increasing their reliability. and availability because of less frequent outages for repairs.

Change (3) is an editorial change to reference redesignated action

Fedora) Rag(stat / Vo), 62, No. 38 / Thursday, February 26, 1987 / Notices statements In the proposed amendment.

'hanges (7) and (8) would mlnlmlze thci number of cold fest starting and loading tests and allow prelubrication and warmup procedures for most surveillance testing, further reducing wear of the diesel generators and further increasing their reliabilityand availability. Change (9) would decrease the number of tests of good diesel generators and decrease the maximum frequency of testing failed diesel generators after their restoration to operable status. However. the reliability goal has not been changed by tha changes to proposed TS Table 4.8.1.1.2-

1. Other changes in the limiting conditions for operation, surveillance requirements and bases are administrative or editorial and do not affect the TS requirements.

Based on its preliminary review, the staff concludes that the proposed changes would not involve a significant increase in the probability or consequences of an accident previously evaluated because the changes will increase diesel generator reliability and availability by eliminating excessive testing which can lead to increased wear and more frequent diesel qenerator failures. The proposed changes would not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed changes do not involve modifications to plant equipment or procedures. The proposed changes do not involve a significant reduction in a margin of safety because the reliability goal of the testing schedule for failed diesel generators after they are repaired would remain the same.

Accordingly, the Commission proposes to determine that the proposed changes to the Technical Specifications regarding diesel generator testing do not involve significant hazards considerations.

Local Public Document Room lacatiam Hinds Junior College.

McLendon Library, Raymond, Mississippi 39154.

Attorney for licensee: Nicholas S.

Reynolds. Esquire. Bishop. Liberman.

Cook, Purcell and Reynolds. 120017th Street NW.. Washinoton. DC 20038.

NRC Project Director: Walter R.

Butler.

.Mississippi Power ai Light Company, System Energy Resources, Inc South Mississippi Electric Power Association.

Docket No. 50-418, Grand GulfNudear Station. Unit 1, Ciaiborne County, Mississippi Date af amendment request: May 28, 1988 as amended November 11. 1986 end February 13. 1987.

Description ofamendment request:

The proposed amendment would change the facilityTechnical Specifications (TSs) as follows: (1) Change the wording in TS 3/4.8.2 and Bases 3/4.6.2.3 to Indicate only one drywell air lock instead of "each" drywell air lock; (2) combine parts 1. 2 and 3 of Action Statement a. of TS 3.8.2.3 into one statement; (3) change the drywell overall air lock leakage test frequency in Surveillance Requirement 4.8.2.3.b,1 from the present frequency of once per 8 months. to each cold shutdown Ifnot performed within the previous 8 months:

(4) change Surveillance Requirement 4.8.2.3.c regarding the verification that only one door in the air lock can be opened at a time from once per 8 months to once per 18 months; and, (5) change Surveillance Requirement 4.6.2.3.d.1 so that the drywell air lock inflatable seal pressure instrumentation channel functional tests willbe run once per 18 months rather than once per 31 days.

Basis forproposed na significant hazards cansi dern tian determination:

The Commission has provided standards for determining whether a significant hazards consideration axis'ts as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facilityinvolves no significant hazards considerations ifoperation of the facility in accordance with a proposed amendment would noL (1) Involve a signiiflicant 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 licensee has provided an analysis of significant hazards considerations in its request for a license amendment. The licensee has concluded with appropriate bases, that the proposed amendment satisfies the three standards in 10 CFR 50.92 and. therefore, involves no significant hazards considerations. The NRC staff has made a preliminary review of the licensee's submittal. A summary of staffs review follows.

Change (1) corrects an error in the TSs; reference to "each" drywell air lock is incorrect because there is only one personnel air lock in the dryweli.

Change (2) clarifies Action Statement a.

because as presently written parts 1, 2 and 3 could be interpreted as separate actions whereas they are intended to be part of the same action. Because changes (1) and (2) correct and clarify the TSs and do not change the TS requirements, these two changes would not Involve a significant increase in the probability or consequences of an accident previously evaluated nor would they Involve a significant reduction In a margin of safety. Because changes (1) and (2) do not Involvti a change In equipment or procedures they would not create the possibility of a new or different kind of accident from any accident previously evaluateiL Changes (3), (4) and (5) involve changes to the survefllance tests for determining operability of the drywell air lock. The present tests require entering the drywell air lock during high power operation because the test intervals are less than intervals for refueling when the plant is in cold shutdown (presently 12 months and in the future. 18 months). During high power operation, the drywall and the drywell air lock are high radiation areas:

therefore, plant procedures restrict entry into the drywell or the drywell air lock when power level is greater than 5'%f rated power. With the present-surveillance requirements. the tests required by TS 4.8.2.3.c (verifying that only one door can be opened at a time) and by TS 4.6.2.3.d.1 (demonstrating operability of the inflatable seal pressure instrumentation'channels for each door) are performed only for '".e outer door when the reactor is operating greater than 5% power. Because testing the inner door would require entering the high radiation field in the drywell air lock, the inner door is dedarcd inoperable and the outer door is locked closed and verified to be locked closed at least once per 31 days as permitted by Action Statement a. of TS 3.8.2.3. The present overall drywell airlock leakage test required by TS 4.8.2.3.b is performed during power operation by declaring the air lock inoperable and entering the air lock briefly to complete the test within the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as permitted by Action

'tatement

b. ofTS 3.6.2.3. Before the air lock is entered to complete the airlock leakage test. the operability of the inner door seal is assured by verifying the inner door seal pressure is greater than 60 psig as required by Action Statement

c. ofTS 3.8.2.3. Change (3) would require the airlock leakage test to be performed during each cold shutdown (induding forced outages) unless it has been performed during the previous 8 months.

This new requirement would eliminate the necessity for licensee personnel to enter a high radiation field to perform the tests and would eliminate the opening of the outer door during high power operation. Assuming no forced outages during an 18-month fuel cyde.

the potential maximum test interval would be increased thus increasing the potential for failure of a leakage barrier in the air lock between tests. However.

the major leakage potential exists for

Federal Register / Vol. 52, No. 38 / Thursday. February 28, 1987 / Notices the inflatable door scale which would remain inAatcd for the proposed interval instead of being deflated and reinflated for the present 6-month test interval.

The greater potential for failure of the door seals during the longer test interval is offset by the lesser potential for failure due to less openings of the outer door for test purposes. Furthermore. the periodic veriiication (once per 7 days) of the airlock door inAatable seal system pressure (TS 4.6.2.3.d.2) willbe retained in the Technical Specifications, thus assuring that the airlock doors are scaled end closed during the interval between tests. Changes (4) and (5) would eliminate the present requirements to test each airlock door during high power operation to verify

~ proper operation of the interlock which prevents opening of both doors at the same time (TS 4.6.2.3.c) and to demonstrate operability of the seal pressure instrumentation channels (TS 4.8.2.3.d.1). In the implementation of the present requirements, however, because of high radiation in the air lock, these tests are not run during power operation (es permitted by Action Statement a.).

Changes (4) and (5). which would permit tests of the door interlocks and seal prcssure inatrumcntatian channels to be made during cold shutdown (once per 18 months) would not therefore result in significant changes to the implementation procedures.

Because changes (3). (4) and (5) would result in less airlock door openings during power operation when the potential accidental release of fission products is greatest and another means is retained in the TS for verifying the door seals are operable between surveillance tests (TS 4.8.2.3.d.2),

changes (3), (4) and (5) would not invalve a significant increase in the consequences of an accident previously evaluated. Because changes (3), (4) and (5) involve only changes to surveillance teats of accident mitigation features they would not involve a significant increase in the probability of an accident previously evaluated nor would these changes create the possibility of a new or different kind of accident from any accident previously evaluated. Because changes (3). (4) and [5) do not Involve a.

change in TS limitingconditions for operation or safety analyses. they would not involve a significant reduction in a margin of safety.

Accordingly. for the reasons cited above. the Commission proposes to determine that these five changes do not involve significant hazards considers tiona.

Local Public Document Room location Hinds Junior College.

McLendon Library. Raymond, Mississippi 39154.

Attorney for licensee: Nicholas S.

Reynolds, Esouire. Bishop, Libennan, Cook, Purcell and Reynolds, 120017th Street NW.. Washington. DC 20038.

NRC Proj ect Director: Walter R.

Butler.

Mississippi Power tk Light Company, System Energy Resources, Inc., South Mississippi Electric Power Association, Docket No. 50-416, Grand GulfNuclear Station. Unit 1, Claiborn'e County, Miaaissip pi Dole ofamendment request: October 17, 1988.

Descriplion ofomendinent request:

The October 17, 1988 application for license amendment requested four changes to the Technical Specifications (TSs). One of the changes, the addition of TSs for smoke detectors in the control rod drive repair room, Ia not addressed in this notice. The otherthree changes to TSs which are addressed herein are: (1)

Change of action Statement b. ofTS 3.4.2.1 by deleting the requirement to close a stuck-open safety relief valve within two minutes and by changing the limitof the suppression pool water temperature from 105 'F to 110'F; (2) addition of a requirement in TS Table 3.8.4.P 1 for a bypass of the thermal overload protection device for the motor operated valve in the reactor core isolation cooling (RCIC) turbine bypass line; and, (31 change of the nomenclature of a valve in TS Table 3.6.6.2-1 for the residual heat removal (RHR) discharge line to the liquid radwaste system.

Basis forproposed no significant hazards consideration determinationl The Commission hes provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92 A proposed amendment to an opera:ing license for a facilityinvolves no significant hazards considerations ifoperation 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 ofaccident from any accident previously evaluated: or (3)

Involve a significant reduction in a margin of safety.

The licensee has provided an analysis of signiAcsnt hazards considerations in its request for a license amendment. The licensee has concluded with appropriate bases. that the proposed amendment satisfies the three standards in 10 CFR 50.92 and. therefore, Involves no significant hazards considerations. The NRC staff has made a preliminary review of the licensee'a submittal. A summary of staff's review follows:

Change (1) would delete the two.

minute time limitfor closing one or more stuck-open safety relief valves and increase the suppression pool temperature limitfrom 105 'F to 110 'F.

The purpose of the time limit for el~sing stuck-open safety relief valves is to assure the valve is closed as soon as possible and thus preserve the maximum capability of the suppression pool to absorb fission product decay heat and reactor coolant and structure sensible heat during a reactor blowdown without exceeding containment structural limits. Necessary and sufficient conditions for preserving adequate capability of the suppression pool ia obtained by specifying minimum pool water level and maximum pool water temperature in TS 3.6.3.1. The specification of a time limitto'lose a stuck-open safety relief valve is therefore not necessary to preserve an acceptable heat absorption capability in the suppression pool. The two.minute time limitwas based on the estimated minimum time for an operator to identify the condition and close the valve with a hand switch in the control room. An April7, 1988 stuck-open safety relief valve event showed that two.minutea was not adequate to determine the condition, operate the valve hand switch and verify that the valve had closed; therefore. the reactor waa manually tripped. Analysis after the trip showed the valve had closed four to five seconds before the trip. Simulated stuck-open relief valves on a simulator have shown that from four to six minutes were required to close the valve when the operators were aware of the condition'o be simulated. Calculations indicate the suppression pool temperature would be less than the limiting temperature for these longer times. The safety ana)ysis of a stuck-open relief valve takes no credit for reactor shutdown within two minutes. The radiological release to the suppression pool following a stuck-opcn safety relief valve ia less than the release from a main steam isolation valve closure. Because the time limit for closing a stuck-open safety relief valve is not necessary in addition to the temperature limit.because the time limit can cause unnecessary reactor trips, and because the safety analyaea do not depend on the time limit, deleting the time limit does not involve a significant increase in the consequences of an accident"previously evaluated. The change in Action Statement b. ofTS 3.4.2.1 to require plant shutdown ifpool water temperature exceeds 110 'F instead of the presently specified 105 'F

Federal Register / Vol. 52, No. 38 / Thursday, February

28. 1987 / Notices

>>auld make this action statement cars.slent wilh Action Statement b.2 of TS 3.8.3.1 which speciflee pool water temperature limits. Because Ihe limiting

'ool temperature would not be exceeded.

Ihe change from 105 'F to 110

'F in the aclion statement does not involve a signiflcant increase in the consequences of an accident previously evaluated. Because change (1) affects accident mitigation only. it does not involve a siqniflcant increase in the probability of an accident previously evaluated. Change (1) does not create the possibility of a new or different kind of accident from any accident previously evaluated because the time limithas not been considered in safely analyses and he proposed temperature limit is consistent with that used in other safety analyses.'Change (1) does not involve a significant reduction in a margin of safely because the proposed temperature limit is the same as that used in the speci."ication of suppression

,ool heat absorption capability.

The Commission has also provided guidance concerning the application of the three standards in 10 CFR 50.92 by providing examples of amendments considered like!y, and not likely. to insolve a signiflicanl hazards consideration. These were published in the Federal Register on March 6. 1988 (51 FR 7751). The KRC staff has made a pre!iininary review of changes (2) and (3) in the licensee's submittal. A discussion of these examples as they re!ate to changes (2) and (3) follows.

One of the examples of actions i'rvaluing no'significant hazards consideration. (ii). is a change that const! tutee an additional limitation, restnction. or control not presently included in the Technical Specifications.

Change (2) is similar to this example bcrause it would add a requirement to coriinuously bypass the thermal overload device for a molor.operated valve in Ihe RCIC system. The bypass willmeet the requirements of Regulatory Guide 1.108. "Thermal Overload Protection for Electric Motor-Operated

~

Va!ves." Addi!ion of the bypass will ircrease the availability of the valve, which is used in the startup of the RCIC tu. bine prior to opening the main steam supply valve.

Another example of an action involving no significant hazards consideration. (i). is a purely administrative change to Technical Specifications: for example. a change in nomenclature. Change (3) is similar to this example because the nomenclature of a valve would be changed. The change wo'uld more clearly identify the valve as one ih the line conn'ecting the

'i IR discharge header to the liquid radwasle system. Several RHR Loop A and Loop B flushfng lines tfe into the discharge header. The nomenclature change would also identify the two ESF power divisions (A and B) that power two solenoid valves that control air supply to the air operated valve.

Accordingly. for the reasons cited above. the Commission proposes to determine that these three changes do not involve signfficant hazards considerations.

Locol Public Document Room locotioni Hinds Junior College.

McLendon Library. Raymond.

htississippi 39154.

Attorney forlicenseei Nicholas, S.

Reynolds. Esquire. Bishop. Liberman.

Cook. Purcell and Reynolds. 120017th Street NW.. Washington, DC 20036.

lVRCProject Director: Walter R.

Butler.

Mississippi Power 5 Light Company.

System Energy Resources, Iuc., South Mississippi Electric Power Association, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County.

Mississippi Date ofamendment request:

November 11, 1986 as revised January

20. 1987, Description ofamendment request:

The licensee'e application dated November 11, 1986 requested two changes to the Technical Speciflications (TSs); (1) Changes to the reactor pressure vessel (RPV) pressure-temperature limitcurves. and (2) changes to the drywell air lock Technical Specifications. This notice addresses only the first item. The proposed amendment would change the Technical Specifications and associated Bases for the RPV pressure and temperature limits (TS 3/4.4.6) to be consistent with the limits provided by the nuclear steam system supplier. the General Electric Company.

Basis forproposed no significont hazards consideration determination:

The Commission has provided standards for determining whether a signiflcant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to an operating license for a facilityinvolves no significant hazards considerations ifoperation 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.

The licensee has provided an analysis of significant hazards considerations ln its November 11, 1988 request for a license amendment. The licensee has concluded. with appropriate bases. thai Ihe proposedemendment meets the three standards ln 10 CFR 50.92 and.

therefore, Involves no significant hazards considerations.

The Commission has also provided guidance concerning the application of these standards by providing examples of amendments considered likely, and

'ot likely, to involve a significant hazards consideration. These were published in the Federal Register on March 8, 1988 (51 FR 7751). The NRC staff has made a preliminary review of the licensee'e submittal. A discussion of these examples as they relate to the proposed amendment follows.

One of the examples of actions involving no significant hazards consideration. (ii), involves a change that constitutes an additional limitation.

restriction. or control not presently included in the Technical Specifications.

The proposed change of the RPV pressure-temperature limitcurves is similar to this example because the proposed curves are more restrictive than the present curves. Curve "A"for hydrostatic or leak testing would not be changed. htore restrictive (higher) temperature limits would be imposed for pressures below 312 psig for Curve "B" which is used for non-nuclear heatup.

cooldown fol!owingshutdown and low power physics tests and for Curve "C" which is used for operation with a'ritical core other. than low power physics tests. The licensee's review of plant operating records showed that the plant hae not operated in the proposed additional restricted area of lhe pressure-temperature limitcurves.

Accordingly. for the reasons cited above, the Commission proposes to determine that the proposed change does not involve significant hazards consideration.

Local Public Document Room locotioni Hinds Junior College.

McLendon Library. Raymond, Mississippi 39154.

Attorneyfor licensee: Nicholas S.

Reynolds, Esquire, Bishop. Liberman, Cook. Purcell and Reynolds. 120017th Slreet NWWashington. DC 20036, hfRC Project Director. Welter R.

Butler Norlheast Nuclear Energy Company (NNECo), Docket No. 50-245, Millstone Nuclear Power Station, Unit No. 1, New London County. Connecticut Date ofamendment request; January

21. 1987.

Federal Register / Vol. 52. No. SS / Thursday, February 2B, 1987 / Notices Descri plion ofamendmcnt request:

The proposed amendment would delete reference to the swilchyard batteriea from technical specifications 3.9245, 4.9.8.2 and Basis 4.9.B.

Basis forproposed no signi%'cant hazards consideration detcrminationi Deletion of the technical specificationa for switchyard batteries willpermit NNECO to shift maintenance responsibility for these batteries from the Millstone Nuclear Power Station.

Unit No. 1 (Millstone 1). Maintenance Department to the New London Substation Maintenance DepartmenL The batteries are part of the components requiring surveillance. calibration and preventive maintenance that make up the switchyard. However, the switchyard breakers. disconnect switches. cables. protective relays, and other components are nol contained in the tcchnical specifications. The proposed changes willprovide consistent treatment of the Millstone 1 preferred offsite power supply.

NNECO has reviewed the proposed changes pursuant to 10 CFR 50.59 and has determined that they do not constitute an unreviewed safety question. The probability ofoccurrence or the consequences of a previously analyzed accident has not been increased and the possibility for a new type of accident has not been created.

Since switchyard batteries do not serve a safety. related function and are not included in the Standard Technical Specifications for General Electric Boiling Water Reactors, NUREG-0123, (STS) the changes proposed herein will establish consistency between the Millstone 1 technical specifications and the STS.

Removal of this nonaafety related provision from the technical speciTications would not result in a dccrcssein safety. because the capability for emergency power is unchanged. and would facilitate focusing attention of maintenance personnel on matters more directly related to nuclear safety.

NNECO baa reviewed the proposed changes. in accordance with 10 CFR 50.92. and has concluded that they do not involve a significant hazards consideration in that these changes would not:

1.!nvolve a significant increase In the probability or consequences of an accident previously evaluated because switchyard batteries do not serve a safety-related function.

2. Create the possibility for a new or different kind of accident from any previously analyzed. The proposed change has no effect on potential failure modes because there are no equipment changes; the only change is the transfer of responaibflity for maintenance of the batteries.

3. Involve a significant reduction in a margin of safety, irithat the reliabilityof station power willnot be degraded. By having the switchyard batteries maintained by tha New London Substation Maintenance Department, the reliability of the offsite power supplies willbe consistent with the entire NNECO system.

Based on the information provided by the licensee. the staff proposes to determine that the licensee amendment request involves no aignlflcant hazards consideration.

LocalPublic Document Boom locotioni Waterford Public Library. 49 Rope Ferry Road, Waterford, Connecticut 06385.

Attorneyforlicensee: Gerald Garfield. Esquire, Day. Berry, 8 Howard, Counselors at Law. City Place. Hartford, Connecticut 06103-3499.

NBCPioj ect Director: Christopher I.

Grimes.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone guc!car Dovrcr Station Utn!t No 2 Nevi London County, Connecticut Dote ofamendment request:

December 2'i, 1986.

Description ofamendment request:

The proposed amendment would change the expiration date for the Millstone Unit 2 Operating License. DPR-65, from December 11. 2010 to July 31. 2015.

Basis forproposed no significont hazards consideration determinationi The currently licensed term for Millstone Unit 2 is 40 years commencing with issuance of the construction permit (December 11, 1970). Accounting for the time that was required forplant construction, this represents an effective operating license tenn of approximately 35 years. ine licensee's application requests a 40-year operating license term for Millstone Unit 2.

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. Although this does not mean that some components willnot wear out during the plant lifetime, design features were incorporated which maximize thc inapectability of structures, systems and equipment. Surveillance and maintenance practices which are implemented in accordance with the ASME code and the facilityTechnical Speciflcations provide assurance that any unexpected degradation in plant

'equipment willbe identified and corrected.

The design of the reactor vessel and its internals considered the effects of 40 years of operation at fu)l power with a plant capacity factor of 80% (32 effective fullpower years). Analyses have demonstrated that expected cumulative neutron fluences wiA not be a limiting consideration. In addition to these calculations. survefllance capsules placed inside the reactor vessel provide a means of monitoring the cumulative effects ofpower operation.

Aging analyses have been performed for all safety-related electrical equipment in accordance with 10 CFR 50.49, "Environmental qualification of electrical equipment important to safety for nuclear power plants," identifying qualiTied lifetimes for this equipment.

These lifetimes willbe incorporated into plant equipment maintenance and replacement practices to ensure that all safety-related electrical equipment remains qualified and available to perform its safety function regardless of the overall age of the plant.

Based upon the above. it is concluded that extension of the operating license for Millstone Unit 2 to allow a 40-year on~otic tt4 t ~

~ enate tnsl v tlu aLs onfntu

~0

~

0

\\

0

~

\\

~

IJ analysis in that all issues associated with plant aging have already been addressed.

Since the proposed amendment involves no change in the Tcchnical Specifications or safety analyses. we conclude that the proposed amendment would not: (I) Involve any significant increase in the probability or consequences of an accident previously evaluated; or (ii) create the possibility of a new or different kind of accident from any accident previously evaluated; or (iii)involve any reduction in the mar(tin of safety.

Based upon the above, the Commission proposes to determine that the proposed amendment. which would provide for a 40.yettr operating life for Millstone Unit 2. involves no signiflcant hazards considerations.

LocolPublic Docuinent Room location: Waterford Pub! ic Library. 49 Rope Ferry Road, Waterford, Connecticut 06385.

Attorneyforlicensee: Gerald Garfield, Esq., Day. Berry and Howard. One Constitution Plaza, Hartford, Connecticut 06103.

NBCeject Director. Ashok C.

Thadani.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit No. 1, Washington County, Nebraska Date ofamendmcnt requesti January

8. 1988 (sic) (1987).

Description ofamendment request:

The proposed amendment would dcleto

Federal Register / VoL 52, No. 38 / Thursday. February 26, 1987 / Notices the hydrogen Auoride detectors from Table 2-11. Toxic Cas Monitors Operating Limits. and from Table 3-3.

Minimum Frequencies for Checks.

Calibrations and Testing of Miscellaneous Instrumentation and Controls. The licensee has concurrently requested that the NRC terminate license SMC-1420 which authorizes possession of natural uranium hexafluoride (UFc) at the Fort Calhoun station. The lipensee's submittal states that the UFi was shipped offsite in April 1986. With no VFs on site. the licensee states that there is no possibility of a toxic gas accident involving hydrogen Auoride. The amendment would also correct a typographical error in Table 3-3.

Basis forproposed no significant hazards consideratioa 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 ifoperation 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 deletion'f the hydrogen Auoride monitors from Tables 2-11 and 3-3 willnot:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluatetL As stated by the licensee, the UFs was shipped offsite and therefore there is no possibility of a toxic gas accident involving hydrogen fluoride.

Accordingly. the deletion of the hydrogen fluoride monitors willnot increase the possibility or consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of accident from'any accident previously evaluated. Because the proposed change willdelete instruments from the Technical Specifications that are designed to monitor hydrogen fluoride and there is no possibility of a toxic gas accident involving hydrogen Auoride. as discussed above. the possibility of a new or different kind of accident from any previously evaluated willnot be created.

(3) Involve a significant reductian in the margin of safety.'fhe proposed change willdelete instruments from the Technical Specifications that willno longer be needed to monitor a hydrogen Aunride toxic gas release because such a release Is not possible. as discussed above; hence, then willnot be a reduction in the margin of safety.

The Commission has provided guidance concerning the application of the standards in 10 CFR 50.92 by providing certain examples (51 FR 7744);

One of these examples. (i), Involving no signiflcant hazards consideration is a purely administrative change to technical speciflcations. The proposed change to correct the typographical error is related to this example Based on the above discussion, the staff proposes to determine that the proposed changes do not involve a significant hazards consideration.

Local Public Document Room locatiom W. Dale Clark Library. 215 South 15th Street. Omaha. Nebraska 68102.

Attorneyforlicensee: LeBoeuf, Lamb, Leiby, and MacRae. 1333 New Hampshire Avenue NWWashington.

DC 20038.

NBCPnjject Director: Ashok C.

ThadanL Portland General Electric Company. ot alDocke1 No. 50-344, Trojan Nuclear Plant, Columbia County, Oregon Date ofamendment requests October 31, 1988.

Description ofamendment request:

The proposed amendment would revise Technical Specification (TS) Section 3/

4.4.9, Figures 3.4-2 and 3.4-3, to reflect heatup and cooldown rate pressure/

temperature limits for a service period of up to 10 Effective Full Power Yean (EFPY) consistent with the May 27. 1963 amended requirements of 10 CFR Part 50, Appendix G. Table 4.4-5 would be revised to reflect compliance with AS'M E185-82 with respect to capsule withdrawals. consistent with the amended requirements of 10 CFR Part 50, Appendix H. The capsule withdrawal schedule would be based on neutron flaence or EFPY, consistent with ASTM E 185-82. The associated bases would also be revised to maintain consistency within the Technical Speciflcations.

Figures 3.4-2 and 3.4-3 of the current TS are based on a service period of up to 15 EFPYs, and do not reflect the amended requirements of 10 CFR Part

50. Appendix G. The capsule withdrawal schedule in Table 4.4-5 is currently based on calendar years, rather than EFPYs as required by the revised 10 CFR Part 50, Appendix K Changes to 10 CFR Part 50 Appendix H require compliance with ASTM E185-82 for,'eactor vessel material irradiation surveillance capsule withdrawaL for capsules withdrawn after July 28. 1983.

The change from calendar years to EFPY is only a minor change in faciIIty operation.

Changes to Section IV.A.2of 10 CFR Part 50 Appendix G require that. when the reactor is not critical and pressure exceeds 20 percent of the preservice system hydrbstatic test pressure, the temperature of the closure flange regions must exceed the reference temperature of the material in these regions by at least 120 'F. To implement this revised regulation, the licensee proposes minor changes to the required pressure-temperature curves for the reactor coolant system. For Trojan, the emended regulation would require that the reactor coolant system be at a temperature higher than 140 'F before exceeding a pressure of 620 psig. To account for possible instrument error, the indicated temperature must be at least 150 'F before exceeding a 4iresiuiry of 560 psig. This represents only a minor change from present requirements.

Basis forproposed no significant hazards cansiderati on determination: 10 CFR 50.92 states that a proposed amendment willnot involve a significant hazards consideration ifthe proposed amendment does not: (i) Involve a signiTicant increase in the probability or consequences of an accident previously evaluated; or (ii) Create the possibility of a new or different kind ofaccident from any accident previously evaluated; or (iii)Involve a significant reduction in a margin of safety. The Commission has also provided guidance concerning the application of these standards by providing certain examples (March 8.

19M, 51 FR 7751). An example of an amendment that is considered not likely to involve a significant hazards considerations is Example (vii)A change to conform a license to changes in the regulations, where the license change results in very minor changes to facilityoperations clearly in keeping with the regulations.

The staff has reviewed the licensee"s no signiTicant hazards analysis and concludes that the proposed changes are within the scope of the Commission's cited example, since the changes are required in order to assure compliance with the regulations, as amended.

Thus. the staff proposes to determine that the requested changes do not involve a signiTicant hazard consideration.

Local Public Document Room location: Multnomah County Library.

801 S. W. 10th Avenue, Portland, Oregon 97205.

Attorneyfor licensee:

J. W. Durham.

Senior Vice President, Portland General Electric Company, 121 S. W. Salmon Street. Portland. Oregon 97204.

Federal Register / Vol. 52, No. 38 / Thursday. February 26, 1967 / Notices NRC Project Director. Steven A.

Varga.

Power Authority of The State of Ne' York. Docket No. 50-286, Indian Point Unit No. 3. Westchester County, New York Dote ofomendmenl request: January

14. 1987.

Description ofomendment request:

The licensee provided the following des."ription:

This proposed revision to the Indian Point 3 Technical Specifications seeks to reduce the elapsed time interval from shutdown requirement for the discharge of more than one region of fuel (72 assemblies). Presently. the Technical SpeciTications require 400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> to have elapsed after shutdown before more than one region of fuel can be discharged from the reactor.

This change willreduce the elapsed time interval prior to discharge to 162 hours0.00188 days <br />0.045 hours <br />2.678571e-4 weeks <br />6.1641e-5 months <br />.

The 400 hour0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> limitation was originally established based on pool decay heat load calculations from a fullcore discharge. not based on any accident analysis. An analysis is provided of the decay heat load after 6 162 hour0.00188 days <br />0.045 hours <br />2.678571e-4 weeks <br />6.1641e-5 months <br /> period which concludes that the maximum bulk pool water temperature would be less than or equal to 198'F and that the spent fuel pool heat removal system ls adequate to transfer the decay heat

'J generated in this case. This is consistent with the NRC Standard Review Plan Section 9.1.3 which states that the temperature of the spent fuel pool water should be kept below the boiling point upon full core discharge.

The analysis assumes 6 component cooling water (CCW) temperature of100

'F. This CCW temperature was determined to correspond to a river water temperature of 87.8'F. A

~easitivtiy study was performed on the analyiis to determine the effect of increasing the assumed river water temperature. The sensitivity study

', results indicate that increasing the assumed river water temperature to 92

'F would incgease the maximum spent fuel pool bu(k temperature to 201.9 'F, which is below the'boiling point.

Bosis forproposed no significonl hozords consideration delerminotion:

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 facilityinvolves no significant hazards considerations ifoperation of the facility in accordance with a proposed amendment would not: (1) involve a significant increase in the prob'ability 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 provided the followinganalysis of this change:

(1) Docs the proposed license cmcndmcnt involve s aignificent increase In the probability or conacqucncca of an accident previously evaluated?

The proposed change docs not incrccac the probability of c previously evaluated accident. The consequences af a fuel handling accident crc described in FSAR Chapter 14.2.1. The assumed fuel decay time after shutdown I~ 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. The consequences of c fuel handling accident assuming c 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay time wiiibound the consequence of such an accident with a greater caauined decay time. As the proposed delay time of 162 hours0.00188 days <br />0.045 hours <br />2.678571e-4 weeks <br />6.1641e-5 months <br /> for the discharge of more than one region of fuel ia grcctcr then 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />. the consequences of the previously cvciuctcd fuel handling accident are not increased.

(2) Docs the proposed license amendment create the possibility of c new or different kind of accident from cny accident previously cvcluctcd7 The proposed change Involves s reduction In the delay time prior to discharge of more than one region of fuel. Reducing thc length of thc delay time willnot create a new or different kind of accident from any accident previously evaluated.

(3) Docs the proposed cmcndmcnt Involve

~ atgniiicant reduction In s margin of acfctyI Upon a complete loaa of the spent fuel cooling system. the adiabatic pool hcstup rate of the pool hca been conservatively calculated to be 15.37 'lhr. For the ccac assuming a river water temperature of 87.6'.

the Instantaneous diachergc of a full cote after s decay time of 162 hours0.00188 days <br />0.045 hours <br />2.678571e-4 weeks <br />6.1641e-5 months <br /> wiiiresult In a steady state bulk pool temperature of 197.7 For this case 0.93 hours0.00108 days <br />0.0258 hours <br />1.537698e-4 weeks <br />3.53865e-5 months <br /> are available to start the redundant spent fuel pooling cooling pump to restore heat removal cnd preclude bulk pool boiling. For the case assuming a river water temperature of 92'. thc Instantaneous discharge of a full core after a decay lime of 162 hours0.00188 days <br />0.045 hours <br />2.678571e-4 weeks <br />6.1641e-5 months <br /> willresult In a steady

~late bulk pool temperature of 201.9'. For this case 0.66 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br /> are svsiicbic to start the redundant spent fuel pool cooling pump cnd preclude bulk pool boiling. Theet time periods are considered adequate to start thc redundant pump.

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

Local Public Document Room locotion: White Plains Public Library.

100 Martine Avenue. White Plains. New York 10601.

Attorneyfor licensee: Mr. Charles M.

Pratt,10 Columbus Circle, New York, New York 10019.

NRC Project Director: Steven A.

Varga.

Public Service Company of Colorado, Docket No. 50-267, Fort St. Vrain Nuclear Generating Station, Plat teville, Colorado Dole ofomendmenl requestr January 15, 1987.

Description ofomendment request:

The licensee has proposed changes to LCO 4.3.1 of the Fort St. Vrain (FSV)

Technical Specifications to require both evaporator-economizer-superheater (EES) sections and both reheater sections be available during operation at power as the minimum number of operable heat exchangers. The FSV Technical Specifications currently require both the reheater section and the EES section of one steam generator and either the reheater section or the EES section of the other steam generator be operable for the removal of decay heal.

For power levels above 39 percent of full power. the plant willonly use the two EES sections Io remove decay heat for the Class 1E and Appendix R design basis shutdowns'. The licensee is retaining the operability requirement,for the reheater sections because they can be used for other accident sequences.

A second proposed change to LCO 4.3.1 is that the EES sections shall be capable ofreceiving water from both the Emergency Condensate Header and the Emergency Feedwater Header instead of the former minimum allowable of only one of these emergency headers.

Basis forproposed no significant hazards considerolion delermi notion:

Recent analyses of Safe Shutdown Cooling with a single EES have shown that maximum fuel temperatures remain well below a level where significant fuel damage could occur, the primary coolant pressure boundary remains intact and, the PCRV safety valves remain seated.

The fuel temperatures are lower than those than would have occurred had decay heat been removed with the rehea ter section. Thus, the consequences of the Safe Shutdown Cooling accident previously evaluated willnot be increased by the LCO 4.3.1 Basis change that omits reliance on the reheaters for Safe Shutdown Cooling. In addition, LCO4.3.1 would require that the EES sections can be supplied by water from either the Emergency Condensate Header or the Emergency Feedwater Header. For these reasons. it is considered that this proposed change does not involve an increase in the probability or consequences of an accident previously evaluated.

The revision to LCO 4.3.1 willnot create a new or different kind of

'ccident. The accident descriptions for Safe Shutdown Cooling with a fire pump

Federal Register / Vol. 52, No. 38 / Thursday, February 2B, 1987 / Notices 58$ F and other Safe Shutdown List equipment in the FSAR (Sections 10.3.9, 14 4.2.1 and 14.4.2.2) willremain essentially as written. FSV willcont'mue to meet the NRCs single failure requirements. FSV contmues to meet the General Design Criteria cominitments in Appendix C of the FSAR without relying on the reheaters for Safe Shutdown Cooling.

Reliance on the EES sections of the steam generators is considered desirable because of their larger heat transfer capability and the fact that their capabilities have been demonstrated in normal plant operation. Thus, the proposed fullyredundant EES sections potentially offer a greater margin of safety. The revision of LCO 4.3.1.

requiring both EES sections to be capable of receiving water from both the Fmergency Feedwater Header and the Emergency Condensatc Header, does not decrease the margin of safety over the former miniinum requirement of the capability of receiving water from only one of these headers.

Based on the above evaluation, il is the staff's initial determination that operation of Fort SL Vrain in accordance with the proposed LCO changes willnot: (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 any margin of safety. He'nce. the staff proposes to determine that this proposed change would involve no significant hazards consideration.

Locol Public Document Room locotiorL Greeley Public Library, City Complex Building, Greelcy. Colorado.

Attorney%r licensea'ryant O'Donnell. Public Service Company of Colorado, P.O. Box 840. Denver.

Colorado 80201&M1 NRC Project Director: Herbert N.

Berkow.

Sacrainento Municipal UtilityDistrict, Docket No. 50-312, Rancho Seco Nuclear Generating Statioa, Sacramento County, California Dote 'ofometidment request: June 18, 1988. supplemented by letter dated January 7, 1987.

Description ofamendment request.

The proposed amendmcnt would delete the surveillance requirements for the Reactor Building Upper Dome Air Circulators (RB-UDAC) from Technical Specification 4.5.2,'1.B.

Basis forproposed n'o significont hozards considerotion determination.

The surveillance requirement for the Reactor Building Upper Dome Air Circulators (RB-UDAC) has been in effect since the issuance ofthe initial operating license for Rancho Seco. Sine then, reviews and data have shown that the sabty function of these circulators, which is to prevent formation of flammable pockets of hydrogen within the Reactor Building (RB) dome during accident conditions, can be adequately accomplished by other existing systems and mechanisms. These existing systems include the RB Emergency Coolers and Containment Spray Systems, as well as natural mechanisms such as convective mixing and molecular diffusion. The RB Emergency Coolers and Containment Spray Systems are safely-related systems with appropriate surveillance requirements i the Technical Specifications to ensure their operability in the event of an accident.

The licensee has reviewed the issue safety-related RB-UDAC against the Commission's Standard Review Plan and the omission of RB-UDACs, or equivalent. at similar reactor pbints, all designed by the same architect-engineer.

The licensee's comparison ofRancho Seco's design with these other plants, which do not have a separate containsnent recirculation system (iw RB-UDAC). revea)s very close similarities in Reactor Buflding design features. This indicates that plants similar to Rancho Seco but of later design satisfy Standard Review Plan 8.2.5. Section I1,3, "Combustible Gas Control in Containment, Acceptance Criteria".

The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c) J. A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation of the facility in accordance with the proposed amendment would noh (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 fram any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

As noted previously, the proposed'hange does not affect ability of the plant to operate safely or perforin necessary safety functions in the event of an accident. The Commission's staff,.

agrees with the licensee's conclusion that plant operation with the surveillance requirement of the RB-UDAC deleted from the Technical Specifications would not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated. The initial design intent of the RB-UDAC locati om Sacramento City-County Library. 828 I Street. Sacramento, California 95814.

Attorneyforlicensee: David S.

Kapian. Sacramento Municipal Utility District. 8201 S Street, P.O.'ox 15830, Sacramento. California 95813.

was to ensure that an accumulation of e

hydrogen would not occur in the RB doree under accident conditions. With respect to this safety-related function of the RB-UDAC. the Rancho Seco Updated Safety Analysis Report (USAR paragraph 6.4.2),states:

The hydrogen willrapidly diffsse by rising into the Reactor Building atmosphere io give an even distribution throughout the Reactor.

Building. in addition. convective mlxiag due to the heating of atr abmre the hot water tn the sump and the forced circulation created by ihe Reactor Building emeittency coolers willenhance thc dispersioa of hydrogen.

Accordingly. operation of the emergency upper dome circutators ie not considered essential to the elimination of regions of high hydrogen concentratirxi.

This conclusion is supported by a research project conducted by the Los of Alamos National Laboratory in 1978.

Their report. NUREG/OR-0304, "Mixin~

of Radiolytic Hydrogen Generated Within a Containment Compartment following a LOCA".provides detailed mathematical and experimental evidence that the molecular diffusion process and thc turbulent mixing caused by small temperature differences between the floor and the ceiling of the containment willbe sufficient to ensure proper mixing of hydrogen in containment.

(2) Create the possibility of a new or, different kind of accident from any accident previously evaluated. The worst accident foreseen, the pos'sibility that the RB-UDAC is unavailable in a future post-accident situation, has already been evaluated as indicated in Item (1). Therefore. the absence of the RB-UDAC or its associated surveillance requirement would not create a new or different kind of accident from any accident p'reviously evaluated.

(3) Involve a significant reduction in a margin of safety. As discussed above, there are various means of achieving the mixing ofpostaccident generated hydrogen in the RB to prevent an explosive or flammable concentration.

Therefore, the possible minor reduction of a margin of safety resulting from the proposed change is considered not

'ignificant.

'Ihus, based an the above criteria, the Commission's staff proposes to determine that thc change involves no

""significant hazards considerations.

Local Public Document Room

Federal Register / Vol. 52, No. 38 / Thursday, February 28, 1'887 / Notices NRC Project Director. john F. Stolz.

South Carolina Electric aod Gss Company, South Carolina Public Service Authority, Docket No. 50-395, VirgilC.

Summer Nuclear Station, Unit 1, FaMield County, South Carolina Dote ofamendment request'ecember

9. 1988.

Description ofamendment request.

South Carolina Electric 8 Gas Company (SCEdG) hereby requests a revision to the VirgilC Summer Nuclear Station Technical Specilications, Section 5.3.1, "Fuel Assemblies." This change would allow SCEkG the flexibilityto reconstitute fuel assemblies in order to reduce coolant activity and utilize the remaining energy in fuel assemblies which contain small numbers of defective fuel rods.

Basis forproposedncp significant hazards consideration determinotiom Section 5.3.1 of the Technical Specifications describes certain design features of the fuel assemblies in the reactor core. The description currently states that each fuel assembly in the core is comprised of 264 fuel rods clad with Zircaloy<. The proposed change would allow substitution of some of the fuel rods with fillerrods consisting of Zircaloyd or stainless steel. The change would also allow fuel rods to be removed from the assembly structure and those positions left vacated. In each reload core the reconstituted assemblies would be specifically evaluated using standard methods currently utilized in accepled reload methodology topicals.

Existing safety criteria and design limits willapply in these evaluations.

Consideration of the effects on peaking factors and core average linear heat rate willbe included in the core physics evaluations.

The Commission has provided certain esrirsples (51 FR 7751) of actions likely to involve no significant hazards considerations. This change is similar to Exemple iiiwhich states: "For a nuclear power reactor. a change resulting from a nuclear reactor core reloading. ifno fuel assemblies signiiicantly different from those found previously acceptable to the NRC-89 for a previous core at the facilityin question are involved. This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed. and that NRC has previously found such methods acceptable."

Accordingly. the Commission proposes to determine that this change does not involve significant hazards considerations.

Loco/Public Document Boom locatioru Fairlield County Library, Garden and Washington Streets.

Winnsboro, South Carolina 29180.

Attorneyforlicensee: Randolph L Mahan. South Carolina Electric and Gas Company, P.O. Box 764, Columbia, South Carolina 29218.

NBCProject Dr'rector. Laster S.

Rub enstein.

South Carolina Electric and Gas Company, South Carolina Public Service Authorit, Docket No. SM95, VlrgflC.

Summer Nuclear Station, Unit 1, Fairfield County, South Carolina Date ofamendment request:

December 11, 1986.

Description ofamendment request; South Carolina Electric 8 Gas Company requests a revision to Technical Specification Section 3.1.2.5. "Borated Water Source-Shutdown." Section 3.1.2.8, "Borated Water Sources Operating." and other 'related Sections 3.5.1, 3.5.5, B3/4.1.2, B3/4.5.4, and B3/

4.8.2.2. The changes revise the Refueling Water Storage Tank (RWST) boron concentration levels from a minimum of 2000 ppm to a minimum of 2300 pprn in Sections 3.1.2.5. 3.1.2.6 and B3/4.12, and from between 2000 ppm and 2100 ppm to between 2300 ppm and 2500 ppm in Section 3.5.5; revise the accumulator boron concentration levels from between 1900 ppm and 2100 ppm to between 2200 ppm and 2500 ppm in Section 3.5.1; and revise the pH value from between 8.5 and 11.0 to 78 and 11.0 for the solution recirculated within containment after a loss ofcoolant accident (LOCA)in Sections B3/4.5.4 and B3/4.8.2.2. The purpose of these changes is to increase the current margin to a post-LOCA shutdown requirem nt in order to have increased assurance of conformance for Cycle 4 and future cycles at VirgilC. Summer Nuclear Station.

Basis forproposed no significant hazards consideration determinati am The purpose of these changes is to increase the current margin to a post-LOCAshutdown requirement (!.e., the reactor willremain subcrltical in the cold condition following a large break LOCA assuming complete mixing of the RWST, reactor coolant system (RCS1 and emergency core coolant system (ECCS) water and other sources of water that may eventually reside in the

'ump post. LOCA with all control rods assumed to be out) in order to have increased assurance of conformance for Cycle 4 and future cycles at VirgilC.

Summer Nuclear Station.

The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). Aproposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifopt.ration of the facility in accordance witlrthe 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 accidents previously evaluated; or (3) involve a significant reduction in margin ofsafety. The licensee has determined that the requested amendment does not involve significant hazards

'considerations for the followingreasons:

(1) The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report is not increased.

a. The only non-LOCA safety analyses affected are those in which the Safety Injection System (SIS) is actuated. For these accidents no adverse impact willoccur, and the conclusions as stated in the FSAR willremain valid.

For most accidents, the higher boron concentrations are beneficial since a more rapid negative reactivity insertion willoccur.

b. In the small break LOCA analysis.

there is no assumption regarding the concentration of boron in the ECCS water and no credit is taken for the negative reactivity produced by soluble boron. Thus, the FSAR conclusions remain valid with the higher allowable boron concentrations.

c. Like small breaks, the large break analysis takes no credit for boron in the ECCS water up to the time of peek clad temperature. The FSAR conclusions remain valid.

d. For maintenance of post-LOCA long term cooling. the higher boron concentrations decrease the maximum allowable time for operator ection (i.e from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> to start hot recirculation and from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> to start hot recirculation and from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> to subsequently alternate hot leg and cold leg recirculation) to prevent boron precipitation. The resulting operator action times are still adequate to assure the required actions can be accomplished to maintain post-LOCA long term cooling.

e. The increased boron concentrations and resulting lower post. LOCA containment spray and recirculating core cooling solution pH are acceptable.

The design basis for pH sensitive issues are maintained.

Federa(

Register / Vol. 52, No. 38 / Thursday. February

28. 198y / Notices 5889 (2) The possibility for an accident or malfunction of a different type than any evaluated previously in the safely analysis report is not created, The proposed changes increase existing boron concentration limits within current operational restraints and without compromising the performance or qualification of safety related equipment. Thus. these changes are considered to be adjustments within the VirgilC. Summer design bases and thus do not create the possibility of a new or difi'erent kind of accident.

(3) The margin of safety as defined in the basis for any Technical Speciflcation is not reduced. The proposed Technical Specification changes maintain the Final Safety Analysis Report designbases and adequate margins of safety. The higher boron concentrations increase the negative reactivity insertion capability of the ECCS providing an increase in the plant's margin of safety.

As discussed in item 1.d above, the higher boron concentrations do impact safety by decreasing the allowable operator action times to prevent boron precipitation in the long term following a LOCA. However, the required operator action scan still be easily accomplished within the minimum acceptable time restraints to prevent boron precipitation.

Therefore, the proposed increases in the Tcchnical Specification boron concentrations do not result in a significant reduction in margins of safety.

The staffhas reviewed the licensee's determination and finds it acceptable.

Accordingly. the Commission proposes to determine that this change does not involve significant hazards considerations.

Loca/ Public Document Room loeatiam Fairfield County Library.

Garden and Washington Streets.

Winnsboro. South Carolina 29180.

Attorneyfor licensee: Randolph R.

Mahan. South Carolina Electric and Gas Company, P.O. Box 764. Columbia, South Carolina 29218.

XRC Project Directors Laster S.

Rubenstein.

South Carolina Electric and Gas Company, South Carolina Public Service Authority, Docket No. 50-395, VirgilC.

Summer Nuclear Station. Unit 1.

Fairfield County. South Carolina Date ofamendment request: January

20. 1987.

Description ofamendment request:

South Carolina Electric 5 Gas Company requests a revision to Technical Speciiication (TS) 4.8.2.3 "Reactor Building Cooling System" and the associated bases to rellect a flow requirement of 2.000 gpm per operating reactor building cooling unit (RBCU) instead of 4.000 gpm per train as currently listed in Technical Speciflcatlons. The proposed change would allow continued operation at lower indicated service water booster pump flowrates due to automatic closure of the outlet valves on the non-operating RBCU's while ensuring that the requlr'ed flow rate to the single operating RBCU in each train ls met.

Basis forproposed na signi%'cant hazards consideration determination:

The existing RBCU service water outlet valves are non:active and therefore cannot be assumed to be capable ofpost accident operation. Since service water (SW) liow presently passes through the operating and non-operating RBCU's.

the TS flow is 200% of the flowrequired for cooling. The modification converts the RBCU service water outlet valves to ective motor operated valves with automatic controls to open the valves on the RBCU selected post accident, and close the valves on the non-selected RBCU.

TS 4.6.2.3 requires that the service water booster pump flow be at least 4.000 gpm to assure that the selected RBCU in each train receives its required 2.000 gptn. Since only 2,000 gpm is required per selected RBCU and the system is designed to have only one RBCU and its associated fan in each train in service for the design basis accident;the required flow through the service water booster pump can be reduced to 2,000 gpm.

The Commission has provided standards for determining whether a significant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation 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 accidents previously evaluated; or (3) involve a significant reduction in margin of safety. The licensee has determined that the requested amendment does not involve significant hazards considerations for the followingreasons:

1. (a) An increase in the probability of the previously evaluated willnot occur.

The valves are being made active to facilitate the removal of heat from the-containment following a Condition IV event by assuring that the entire SW flowprovided for containment cooling is delivered to the active (fans running) reactor building coolers.

(b) There willbe no increase In the consequences of an accident previously evaluated in the FSAR/FPEL Since this modification willincrease the SW lIow margin it willenhance the capability to remove reactor building heat followinga Condition IVevent and assures that any otential release of radioactivity willbe ess than the QAR licensing basis.

(c) The upgrade of these valves to active status willnot adversely affect the functions of equipment important to safety as previously evaluated in the FSAR/FPEIL Separation ofredundant electrical systems ls maintained per FSAR section 8.3.1.4.1. and Electrical Construction Guidelines Dwg. S-200-926. sheet 6. revision 3. paragra ph 2.3.8; and sheet 4, revision 5. paragrsph 2.1.8.

This separation is maintained by physical distance, and by the installation of a single Kaowool wrap, acting as a suitable barrier. Installed conduit is designed to withstand a seismic event, and to avoid main steattr line break 'jet-cones'. The modified valve operators are seismically designed. and valve position status lights on the engineered safety features equipment board are designed as described ln the FSAR.

(d) The use ofactive valves in these positions does not Increase the consequences of a malfunction of equipment important to safety as valves as the single active failure results in the same consequences from the accident.

Allaccidents evaluated in the FSAR/

FPER bound any accidents which these valves may create or contribute to.

2. The proposed change willnot create the possibility of a new or different kind of accident from accidents previously evaluated because the original design was reviewed to10CFR Part 50,

'ppendix A. criterion 38, 39 and 40. and the proposed change is in accordance with that criteria. No malfunctions of equipment important to safety are anticipated which willbe different from those malfunctions already evaluated in the FSAR/FPER. The modification will eot induce water hammer transient loadings in the piping system. Flow limitingorifices in the discharge pipe are revised to maintain the SW systein pressu're in the reactor building and to prevent excessive flow through the reactor building cooling coils. The reduced flow thru the SW booster pump after the modification is greater than the minimum flow at which the pump was tested. The valve operator stroke. time, maintains the margin of safety as delined in TS Table 3.3-5 for service water response limes.

3. The margin of safety as deiined in the bases to TS 3/4.8.2.3 is not reduced by this modification. This modification willpermit a reduction in the

Federal Register / Vo), 52, No. 38 / Thursday, February 26, 1987 / Notices surveillance requirements for SW flow.

The TS surveillance flowof4.000 gpm is the total of: (1) 2.000 gpm to the active funrtioning fun cooler. and (2) 2,000 gpm to the,non.functioning fen cooler. This mo".':cation willallow the isolation of flow io the non.functioning reactor building fan cooler. This willpermit a reduction in the surveillance flow requirements to 2,000 gpm. This modification wiltincrease the margin of flow tv tho reactor building fan cooler operating post accident above the 2,000 gpm required and willprovide cooling capacity greater than that assumed for accident analysis.

The staff has reviewed the licensee's determination and finds it acceptable.

Accordingly. the Commission proposes to determine that this change does not involve significant hazards considerations.

Loca/Public Document Room locatiamFairfield County Library, Garden and Washington Streets, Winnsboro, South Carolina 29180 Attorney for licensee: Randolph R.

Mehan. South Carolina Electric and Gas Company. P.0. Box 764. Columbia. South Carolina 29218.

NRC Project Director. Lester S.

Rubens tein.

Tennessee Valley Authority, Docket Nos. ~27 and 50-328. Sequoyah Nuclear Plant, Units 1 aad 2. Hamilton County'ennessee Dale ofomendmenl request:

December 23. 1988.

Descriplion ofomendmenl request:

The proposed amendments would increase the maximum allowable response time for the alignment of the safety injection system valves between the charging pumps. the volume control tank. and the refueling water storage iank (RWST).

ksis forproposed no signifiant hazards cansideralion detenninatian:

The Commission provided certain examples (51 FR 7744) of actions likely to involve no significant hazards considerations. The request involved in this case does not match any of those.

examples. Therefore. the staff has reviewed the proposed change and the no significant hazards analysis performed by the licensee.

ln its determination the licensee reached the followingconclusions.

(1) The proposed amendment did not involve a significant increase in the probability and consequences of an accident previously evalualed since a Westinghouse analysis of the proposed valve stroke time increase would not adversely impact the current analyaea results contained in the final safety analysis rcport.

(2) There is no possibility of a new or different kind of accident since the change only involves valve regearing.

(3) There is not a significant reduction in margin. This ls because the change does not adversely impact any of the safety margins in the present analysis.

Based on its review of the licensee's determination. the staff. finds that the amendment request meets the standards for concluding that no significant hazard exists. Accordingly, the staff proposes to determine that the proposed changes do not Involve a significant hazards consideration.

Local public Document Room localiom Chattanooga-Hamilton County Bicentennial Library. 1001 Broad Street, Chattanooga, Tennessee 37401.

Altanieyfar licenseei Mr. Lewis E Wallace, Acting General Counsel, Tennessee Valley Authority, 400 Commerce Avenue, 811B33. Knoxville, Tennessee

37902, NRC Pleat Director. B.J.

Youngblood.

Union Electric Company,'Docket No. N-483, Callaway Plant, UnitI, Callaway County, Missouri Date ofamendmenl request. January

29, 1987.

Descri pdon ofamendment request:

The proposed amendment requests that the Action Statement associated with the Ultimate Heat Sink be divided into 2 separate Action Statements to provide clarification and to recognhe that there are 2 separate trains ofcooling available (100 percent redundancy) for the Essential Service Water System.

Currently as stated, the Action Statement requires declaring the Ultimate Heat Sink inoperable ifone train of cooling tower has an inoperable cell. Ifthe Ultimate Heat Sink is inoperable. then both trains of the Essential Service Water are inoperable and require entry into the more restrictive action required by Technical Specifics tion 3.0.3.

aosis forproposed no significont hazards consideralian delenninalion: In accordance with the requirements of 10 CFR 50.92, the licensee has submitted the followingno significant hazards determination:

This amendment request conaiata of a change to Technical Specification 3/4.7.5. The followingdtacuaalon addresses this change and ii~ corresponding significant hazards evaluation.

This change involves the wording of thc Action Statcmcnt for Technics) Specification 3/4.7& tt wiiibe revised io catsbiiah two Action Statcmenta in lieu of the existing onc to clearly delineate that there are two tndcpcndcm trains of cooling available to luatify the operability requirements of the Ultimate Hest Sink (UHS). As cuircntly stated, the UHS ia considered inoperctdc with one cell of one cooling tower train nut.

of.service. tfthe UHS ia inopcmbtc. then ihc entire Essential Service Water Syaicm is Inoperable. and the requirements of Technical Specification 3.0.3 must be mci.

Tht~ change providcamnaiatcncy between the Tcchnical SpccifiQiiona and the FSAR aa to thc purpose for providing two independent traina of cooling for the UHS.

This change does not involve c algnificaiit incrccac in the probabilny or consequences of cn accident previously evaluated. The propoacd change provides ctcrificatioiitn catcbiiah opcrebitiiy of the UHS thus averting the entry into an Action Statement eironcoualy.

Thia change does not create the possibility of a new or different kind of accident from any accident prcviouaty evaluated. This is baaed on the fact that the method and manner of plant operation ia unchanged.

This change docs not involve a significant reduction in c margin of safety. This ta baaed on the fact that no design change ia involved.

but the intent of the Technical Spccilicction ia clarified to meet the aa.butii condition as specified in the FSAR.

Based on the previous discussions. the licensee concluded that the proposed amendment request does not involve a significant increase in the probability or consequences of an accident previously evaluated; nor create the possibility of a new or different kind of accident from any accident previously evaluated: nor involve a significant reduction in the required margin of safety. The staff has reviewed the licensee's no significant hazards considerations determination and agrees with the licensee's analysis.

The staff has therefore made a proposed determination that the licensee's request does not involve a significant hazards consideration.

Local Public Document Room lacalian: Fulton City Library, 709 Market Street

~ Fulton. Missouri 65251 and the Olin Library of Washington University.

Skinker and Lindell Boulevards. St.

Louis, Missouri 63130.

Allorneyforlicensee: Gerald Charnoff. Esq., Shaw, Pittmen. Potts tt Trowbridge. 2300 N Street. NW.,

Washington. DC 20036.

NRC Project Director: B.J.

Youngblood.

Virghla Electric and Power Company, Docket Nos. N-338 and 50-$39, North Anna Power Station. Units No. 1 aod No.

2. Louisa County. Virginia Dote ofomendmenl request:

November 25. 1988.

Description ofAmendment Requestr The proposed changes would modify the NA-1tk2Technical Specifications (TS)-

3/4.12 (Radiological Environmental Monitoring) to reflect established practices, to agree with NRC approved

Federal Reg(ster /-Vol. 52, No, 38 / Thursday, February 28, 1987 / Notices documents. and to conform to regulatory guidance.

The proposed changes to the NA-182 TS willcorrect an administrafive error in Table 4.12-1. paragraph Z.c. This table lists a sample point for offsite monitoring to be 15 to 30 kilometers from the site. This spccification came In existence when the radiological portion of the Environmental TS were incorporated into the NA-1h2 TS (Amcndmcnt 48 for Unit 1 and Amendment 31 for Unil 2).

The original Environmental TS listed eight locations for continuous airborne sampling. There was 1 at the site. 3 on the site perimeter. 3 in nearby communities and 1 at the Orange District Office. These locations are also specified in the Offsite Dose Calculation Manual (ODCM) in accordance with TS requirements. The ODCM lisle site 24, the "Control" monitor at the Orange District Office. as being "22 miles" (35.4 km) from the site. This monitoring site was placed in service In December 1978 as a control location for NA-1h7. and has provided the station with a well established 10 year base ofsampling data. The location is secure from vandalism and assures a stable power source. i In order to correct this discrepancy.

TS Table 4.12-1 paragraph 2.c willbe changed to read "1 sample from a control location1$ -40 kilometers distant and in the least Ircvalent wind direction." This change willbring the NA-1h2 TS into agreement with the NRC approved Offsite Dose Calculation Manual.

Table 4.12-2 of the NA-1hZTS Ilats a reporting level of 30.000 pCi/liter for HI.

This is to be changed to 20.000 pCi/liter in accordance with NURFG-0472, Radiological Efiluent Technical Specifications for PWRs, Revision 3. In addition there is a footnote being added which specifies the sample to be tested.

Table 4.12-3 of the NA-1h2 TS 3/

4.12.1 lists a Lower Limitof Detectability (LLD)for High Resolution Ge(LI)

Gamma Spectroscopy Environmental sample of 10 pCi/liter for I>ii and LLDof 3000 pCi/liter for HI. These are to be changed to 1 pCi/liter for Iisi and 2000 for t4 in accordance with NURFG-0472, Radiological Effluent Technical Specifications for PWRs, Revision 3.,

Finally. a footnote is being deleted fnnn Table 4.12-3 which specifies that. "The ILDfor Gamma isotopic analysis shall be 'used.- A footnole being added lo Taole 4.12-2 and 4.1M willstate that,

-The LLDvalue ie for drinking water samples."

Basis forproposed no significant hazards consideration determiaaliaru The Commission has provided guidance concerning the application ofstandards for determining. whether a proposed action involves a significant hazards consideration by providing certain

'xamples (See 51 FR 7751). Example (I) states: "Apurely admlnistraUve change to technical apecificationa: for example, a change to achieve consistency throughout the technical apecificationa, correction of an error. or a change In nomenclature." The proposed change Is enveloped by example (i) above. since the proposed change would correct an error presently existing in Table 4.1P 1, paragraph 2.C and achieve consistency with the NRC approved ODClvf and NUREG-0472 for Tables 4.1Z.Z and 4.12.3. Accordingly, the Commission proposes to deternine this change involves no significant hazards consideration.

Local Public Document Room location: Board of Supervisors Office, Louisa County Courthouse. Louisa.

Virginia 23093 and the Alderman Library, Manuscripts Department, University of Virginia, Charlottesville, Virginia 22901.

Attorneyforlicensee: Michael W.

Maupin, Esq. Hunton. Williams. Gay and Gibson, P.O. Box 1535, Richmond.

Virginia 23212.

NRC Project Director. Laster S.

Rubenstein.

Washington PubUc Power Supply System, Docket No. ~97, WNP~

Richland, Washington Dates ofamendment requester January 31, April11. July 22, 1988 and January 9. and February 11, 1987.

Description ofamendment request:

This proposed amendment, ifapproved.

willmodify the WNP-2 Technical Specilications surveillance requirement to aUow Appendix J Type B and C testing on an extended schedule of 27 months maximum between tests instead ol the 24 month maximum now required by the Technical Specifications.

Concurrently, the Suppl~eitt Ia seeking exemptions to the requirements of 10 CFR Part 50, Appendix J regulations which require Type C testing during reac! or shutdown for refueling and Type B and C testing at intervals no greater than 2 years.

WNP-2 provides power to the Bonneville Power Administration (BPA) grid which is heavily dependent on hydroelectric power generation, in turn~ -,

impacted by seasonal variation In snow pack and spring runoff in the Pacific Northwest. The spring runoff and subsequent increase in hydroelectric capacity is the most opportune time lor WVP-2 refueling. Hence, the BPA directs the Supply System to refuel on a yearly basis. ideally coinciding with the peak period of hydroelectri capacity.

This peak period can very from year ta year by as much as three months.

As most ol the Type B and C testing ia carried out during refueling outages, strict compliance with the Technical Specificationc.as currently applied requires testing each valve every year during refueling. The Supply System proposes that the B and C testa would be performed on approximately half the valves during each (annual) refueling outage, such that each valve is teated approximately every!wo years (lhe actual interval being no greater than 27 months).

The Supply System has proposed criteria lor the acceptable leakage of individual valves and wiU retest. at the next refueling outage, those valves that fail the leakage criteria.'Specifically. the Supply System's criteria require~lac~i each valve in one ol three categories depending upon its service and previous leakage history:

1. Some valves. such as Main Steam Isolation Valves. willbe tested during each refueling outage regardless of previous leakage history.

2. Valves with leakage rates that, exceed the target leakage rate by a factor of five or more and those valves that were reworked due to excessive leakage during the preceding outage will be tested during each refueling outage.

3. Valves with leakage rates that do not exceed the target leakage rate and did not exceed this leakage during the previous oulage in which they were tested willbe tested every other outage.

Target leakage rate is determined from past performance and is based on ensnring that the overall leakage rate'is less than the Technical Specification limit.

Basis forproposed na significant hazards consideratian determinau'aru The Commission has provided standards for determining whether no significant hazards considerations exist (10 CFR 50.92(c)). A proposed amendment to an operating license for a lacilityinvolves no ~ignificant hazards consideration ifoperation 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 z a new or different kind of accident'from

~ "

an accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The licensee has determined, and the staff agrees. that lhe requested amendment does not: (1) involve a significant increase in the probability or consequences ol an accident previously

Federal Register / Vol. 52, No. 38 / Thursday, February 28, 1987 / Notices evaluated because the proposed change continues a containment isolation valve testing program and operation governed by the overall bounds required by Appendix J of 10 CFR Part 50 and incorporated into the Technical Specifications: or (2) create the possibility of a new or different kind of accident than previously evaluated because changing to a 22 month testing frequency rather than the 24 month frequency in the regulations does not affect the accident scenarios previously evaluated or (3) involve a significant reduction in a margin of safety because, as indicated in (1) above. operation is maintained consistent with the original intent of Appendix J, thus the original safety margin is maintained because those valves that contribute most to the total leakage willbe detected. repaired and retested on a shorter (annual) cycle.

Based on our review of the proposed modification. the staff proposes to determine that the requested change to the WNP-2 Technical Specifications involves no significant hazards considerations.

Local Public Document Room:

Richland Public Library. Swift and Northgate Streets, Richland.

Washington 99352.

Attorney for the Licensee: Nicholas Reynolds. Esquire. Bishop. Libcrman, Cook. Purcell and Reynolds. 1200 Seventeenth Street NW.. Washington, DC 20036.

NRC Project Director. Elinor G.

Adensam.

Washington Public Power Supply System. Docket No. 50-391, WNP-2, Richland, Washington Date ofamendment re riuest.

December 12, 1986.

Description ofomendmenl request: In its December 12. 1986 letter. the licensee rin>>esi~d rhapgoa tn WNP-2Terhnirel Spocification 3/4:/.4, "Snubbers". The Technical Specification currently requires that. Ifa snubber is removed from its associated system while that system is required to be operable. the system Is to be declared inoperable.

requiring entry into the applicable Action Statement. The licensee notes that during Operation Conditions 4 and

5. snubbers are removed periodically for (1) functional testing. and (2) eliminating interferences in the performance of maintenance activities on adjacent equipment. The requestml change would permit the removal of snubber(s) for maintenance or testing (while in Operational Condition 4 or 5) without having to declare the associated system inoperable. provided its (their) removal is substantiated by engineering analysis Additionally. the current Technical Specificatio require's that visual inspection periods of a snubber type occur every 18 months. N 25 percent. if ao inoperable snubbers of each type on any system are discovered per inspection period. This requires that a visual inspection of a snubber type be performed between 13.5 months and 22.5 months following the previous Inspection. At WNP-2, this visual inspection is done during refueling outages, which occur on 12-month cycles. The current visual inspection requirement does not recognize 12-month refueling cycles. Accordingly, the licensee has requested that the visual inspection criteria be revised to 18 months. -50 percent, + 25 percent. This would require a visual inspection of a snubber type between 9 months and 22.5 months followingthe'previous inspection. Without such a change. the licensee could be required to shutdown between refueling outages for the purpose of snubber surveillance. even when the last inspection revealed no failures. The proposed change reduces the minimum interval between inspections; however. the maximum interval between inspections remains 22.5 months. The Basis of the Technical Specification would be revised to be consistent with the proposed Technical Specification change.

Basis forproposed no significant hazords consideration determinationr The Commission has provided standards for determining whether a signiiicant hazards consideration exists (10 CFR 50.92(c)). A proposed amendment to an operating license for a facilityinvolves no significant hazards consideration ifoperation 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 an accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The proposed changes do not involve a significant Increase in the probability or consequences of an accident previously evaluated. The proposed Technical Specification change involves snubber removal In Operational Conditions 4 and 5, "Cold Shutdown" and "Refueling". respectively. The proposed modification does not increase the probability of an accident previously evaluated because those accidents were not initiated by snubber failure in Operational Conditions 4 and 5.

Similarly. the proposed modification does not increase the consequences of an accident previously evaluated because engineering analyses performed by the licensee demonstrate the structural integrity of safety-related systems in which snubbers may be removed during Operational Conditions 4 and 5. It should be noted that during these operational conditions. the reactor is shut down; therefore, stresses and loads on piping systems are much lower than those stresses and loads occurring during power operation. The licensee's analyses also demonstrate that while in Operational Conditions 4 and 5.

structural integrity of safety-related systems is maintained during and following a seismic event. The proposed modiTication to the interval of visual inspections. should no inoperable snubbers be discovered in the previous inspection. is purely an administrative change to accommodate a 12.month operating cycle, and does not increase the probability or consequences of an accident previously evaluated, because the maximum time between visual inspections remains 22.5 months.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. Engineering evaluations will be conducted to demonstrate the capability of the snubbers remaining in place to handle the normal and accident piping loads anticipated in Operational Modes 4 and 5. Accordingly, no new failure mode is created. The change in visual inspection period in no way creates different accidents from those previously evaluated because this change is an administrative change to accommodate a 12-month operating cycle.

The proposed changes do not involve a signifjcant reduction in a margin of safety. Engineering analysis and generic snubber replacement criteria assure that sufficient margin exists to accommodate normal and accident loads anticipated while in Operational Conditions 4 and 5.

Regarding the change to visual inspection period. the margin of safety is actually increased because the proposed change permits the visual inspection to take place as early as 9 months after the previous inspection, instead of the 13.5 months currently required in the Technical Specification. The maximum visual inspection period remains as 22.5 months following the previous inspection.

Based on its review of the proposed modification. the Commission proposes to determine that the proposed change to the WNP-2 Technical Specifications involves no significant hazards consideration.

Local Public Document Room:

Richland Public Library, Swift and

Federal Rag)star / Vol. 52, No. 3$ / Thuraday, Pebrttaty 2S, 1987 / Noticott Northgate Streets, Richland, Washington 99352, Attorney for the Licensee: Nicholas Reynolds, Esquire, Bishop, Llberman, Cook. Purceil and Reynolds, 1200 Seventeenth Street NW., Washington, DC 20038 PiRC Project Director: Hfnor G, Adensam.

PREVIOUSLY P UBILSHEDNOTICES OF CONSIDERATIONOF ISSUANCE OF AMENDMENTSTO OPERATING LICENSES ANDPROPOSED NO SIGNIFICANTHAIEARDS CONSIDERATION DETERMINATION AND OPPORTUNITY FOR HEARING The followingnotices were previously published as separate individual notices. The notice content was the same as above. They were published as individual notices because time did not allow the Commission to wait for this bi-weekly notice. They are repeated here because the bi.weekly notice lists all amendments proposed to bc issued involving no significant hazards consideration.

For details, see the'individual notice in the Federal Register on the day and page cited. Ttus notice does not extend the notice period of the original notice.

Cleveland Electric Illuminating Company, Duquesne LIght Company, Ohio Edison Company, Pennsylvania Power Company, Toledo Mson Company, Docket No. 50-440, Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio, Date ofamendment request: January

23. 198'7.

Briefdescription ofamendment rcquestr The proposed amendment

, would allow sale and leaseback transactions by Ohio Edison Company (OE) relating to its 30.0% ownership interest in Perry Unit 1 and common facilities.

Dote ofpublicotion ofindividual noticein Federal Register. February 11, 1987 (52 FR 4432).

Expiration dcte ofindividual notice.

March 13, 1987.

Local Public Document Room locotion: Perry Public Library, 3753 Main Street, Perry. Ohio 44081.

CPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania Date ofamendment request: January

23. 1987.

'rief description ofamendment: The Emergency Feedwater (EFW) System is beir.g upgraded to safety-related criteria during the present Cycle 8 Refueling Outage in order to comply withTMI-1 restart commitments and NUREG-0737, Action Item II,E1. In accordance with the licensee's application dated January

'3, 1987 (Technical Specification (TS)

Change Request No. 168), this amendment would provide appropriate operability and surveillance requirements for the modified EFW System and associated Heat Sink Protection System (HSPS).

Administrative and editorial changes to affected TS pages were also requestecL including revisions to the EFW and HSPS bases.

Dote ofpublication ofindividual noticein Federal Register. February 4, 1987 (52 FR 3515)-

Expirrrtion dale ofindividualnotice:

March 8. 1967.

Local Public Document Room locotiom Government Publications Section. State Library of Pennsylvania.

Education Building, Commonwealth and Walnut Streets. Harrisburg, Pennsylvania 1712L Toledo Edison Company and The Cleveland Electric Illuminating Companv. Docket No. 50-348, Davis-Besse Nuclear Power Station, Unit No. I, Ottawa County, Ohio Date ofamendment request: January 21, 1987.

Briefdescription ofamendment: The amendment would revise Technical Specification (TS) Sections 3.8.2.3.

4.8.2.3.1. 4.8.2.3.?

and 4.8.2.4.1 and the associated Basis Section 3/4.8. These TS Sections relate to the D.C. electrical power distribution system during plant operation and when shutdown. The proposed revisions would (1) change certain nomenclature used In the TSs to the specific equipment designations used at Davis-Bassa, and (2) revise the Surveillance Reqm<<ements for D.C.

Distribution using the guidance of the model technical specifications for station batteries issued by the NRC on July 18, 1981. The NRC guidance was based upon Regulatory Guide 1.129, February 1978. and IEEE Standard 450-1980. The current Davis-Besse TS Surveillance Requirements were based on earlier revisions of these documents.

Dote ofpublication ofindividual noticein Federal Regbiter. February 2, 1987 (52 FR 3182); corrected Febiuar'y 11, 1987 (52 FR 4438) ~

Expiration date ofindividualnotice:

March 4, 1987.

Local Public Document Room location: University ofToledo Library.

Documents Department, 2801 Bancroft Avenue, Toledo. Ohio 4360L NCflCE OF ISSUANCE OF AMENDMENTTO FACILITY'PERATING LICENSE During the period since publication of the last bleakly notice. the Commission has Issued the following amendments. The Commission has determined for each of these amendments that the application 4

compiles with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission 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.

Notice of Corisideration of Issuance of Amendment to Facility Operating License and Proposed No Significa'nt Hazards Consideration Determination

~

and Opportunity for Hearing in connection with these actions wes published in the Federal Register as indicated. No request tor a hearing or petition for leave to intervene was filed followingthis notice.

Unless otherwise Indicated, the Commission nas determined thui these amendments satisfy the criteria for categorical exclusion in accordance with 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. Ifthe Commission has prepared an environmental assessment under the special circumstances provision ln 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. Allof these items are available for public inspection at the Commission's Public Document Room, 1717 H Street NWWashington. DC, and at the local public document rooms for.the particular facilities involved. A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, Washington, DC 20555. Attention:

Director, Division ofLicensing.

Arizona Public Service Company, et aL Docket Nos. STN 50-528 and STN 50-529, Palo Verde Nuclear Generating Station, Units 1 and 2, Marlcopa County, Arizona Dote ofapplr'cation for amendments:

November 8, 1988.

5874 Federal Register / Vol. 52. No. 38 / Thursday, February

26. 1987 / Notices Briefdescription ofomendmenlst The amendments revised the Technical Specifications to change the organizational structure and to change the version of ANSI Standard N509 to be used for initial testinq of the charcoal filter units in the Essential Filtration Systems.

Dole ofissuonce: February 9. 1987.

Ef'ec.'i ve date: February 9, 1987.

Amendment Nos.: 13 and 7.

Focility Operating License Nos.i NPF-41 ari NPF-5tt Amendments revised the Technical Specifications.

Dole ofinitialnoticein the Federal Register. December 17, 1986 (51 FR 45192). The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 9. 1987.

No significant hazards consideration comments were received: No.

Local Public Document Room location: Phoenix Public Library.

Business. Science and Technology Department. 12 East McDowell Road.

Phoenix. Arixona 85004.

Conunoawealth Edison Company, Docket No. 50-374, La Salle County Station, Unit 2, La Salle County. Illinois Date ofamendment request: October

16. 1986.

Briefdescription ofamendment: This amendment revises the La Salle, Unit 2 Technical SpeciTications to permit replacirig an existing peripheral locking piston control rod drive module with a fine motion control rod drive module during one fuel cycle.

Dole ofissuoncet February 9. 1987.

Effective dote: Upon startup following the first refueling outage.

Amendment No, 30.

Foci%'ty Operating License No. NPF-18: Amendment revises the Technical Speciflcations.

~'Bate ofinitialnolicein Federal Register. November 5, 1986 (51 FR 40277)

The Commission's related evaluation of the amendment is contained In a Safety Evaluation dated February 9. 1987.

No signiiicant hazards consideration coinments received: No.

Local Public Document Room locotion: Public Library of Illinois Valley Community College. Rural Route No. 1, Oglesby. Illinois 61348.

Commonwealth Edison Company',

Docket Nos. 50-254 aad 50-265, Quad Cities Nuclear Power Station, Units 1 and 2, Rock Island County. Illinois Dale o(opplicotion for o'mendmenls:

February 17. 1983. as supplemented and superseded August 23. 1984 and January

20. 1986. respectively.

Briefdescription ofamendmentst Changes were made in the Administrative Control section of the Technical Specifications, primarily to reflect changes In staff organization. and changes in 10 CFR Part 50 regarding minimum staffing and Immediate notification requirements, and the Ucensee Event Reporting system.

Dote ofissuance: February 2, 1987.

Effective dote: February 2, 1987.

Amendment Nos.: 100 and 97.

Foci%'ty Operating License Nos. DPR-29 and DPR-30. Amendments revised the Technical Speciiications.

Dote ofinitiolnoticein Federal Register. November 19, 1986 (51 FR 41848) The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 2. 1987.

No significant hazards consideration comments received: No.

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

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck Plant. Middlesex County, Connecticut Date ofopplicotion foramendment:

October 20. 1981.

Briefdescriplion ofomendment.'his amendment revises Technical Speciflcation 4D by deleting the requirement that surveillance tests of the diesel generators include automatic loading of the charging pumps. No credit for automatic loading of the charging pumps is assumed In any:accident safety analysis.

Date ofissuance: February 9, 1987.

Effecti ve date: February 9, 1987.

Amendment No. 89.

Facility Operating License No. DPR-.

ttt. Amendment revised the Technical Specifications.

Date offnitiolnoticein Federal Register. October 26, 1983 (48 FR 49580).

The Commission's related evaluation of the amendment ls contained in a Safety Evaluation dated February 9.

1987. No signiflcant hazards consideration comments received: No.

Local Public Document Room locoliorc Russell Library. 124 Broad Street. Middletown, Connecticut 06457, Consumers Power Company, Docket No.

50-255, Palisades Plant, Van Bursa County, Michigan Date ofapplication for omendmenl:

July 9. 1984.

Briefdescription ofomendment: The

~ amendment Incorporated Technical Specifications pertaining to the implementation of a post-accident sampling system. The change ls consistent with the requirements of NUREG-0737 TMIAction Plan II,B.3 and the guidance given in Generic Letter 83-37.

Dale ofissuonce: February 3. 1987.

Effeclive dole: February 3, 1987.

Amendmenl No. 100.

Pro vt'st'anal Operating License No.

DPR-20. The amend0ient revised the Technical Specifications.

Date ofinitialnolice in Federal Register. August 22. '1984 (49 FR 33363).

The Commission's related evaluation

'f the amendment ls contained in'a Safety Evaluation dated February 3, 1987. No significant hazards consideration comments received: No.

LocolPublic Document Room locoliorc Van Zoeren Library, Hope College. Holland, Michigan 49423.

Consumers Power Company, Docket No.

50-255, Palisades Plant, Vaa Burea County, Michigaa Dale ofopplicotion foromendmenl:

December 19, 1985.

Briefdescription ofamendment: The amendment deletes out of date footnotes and incorrect references to a motor control center within the Technical Specifications.

Dale ofissuonce: February 10. 1987.

Effective dale: February'10. 1987.

Amendment No. 101.

Provisional operating License No.

DPR-20. The amendment revised the Technical Specifications.

Dale ofinitialnolicein Federal

, Register. December 30. 1986 (51 FR 47072 at 47077). The Commission's related evaluation of the amendment is contained In a Safety Evaluation dated February 10. 1987. No significant hazards consideration comments received: No.

Loco/Public Document Room location: Van Zoeren Library. Hope College, Holland, Michigan 49423.

Consumers Power Company, Docket No.

50-255, Palisades Plant, Vaa Burea County, Michigan Date ofappli colion foramendment:

June 25, 1982.

Briefdescription ofamendment: This amendment revised Section 4.7.1(d) of the Technical Specifications and included an a'ddition to the Bases of Section 4.7 to reflect the manufacturer's true nameplate rating of the emergency diesel generators.

Date ofissuance: February 11. 1987.

Effective date: February 11, 1987.

Amendment No. 102.

Pro visional Operating License No.

DPR-20. The amendment revis~d the Technical Specilications.

Dole ofinitiolnoticein Federal Register. August 23. 1983 (48 FR 38398).

Fecferal Regfeter / Vol. 52, No. 8$ / Thursday, February 28, 1887 / Not(cee The Commission's related evaluation of the amendment ls contained In a Safety Evaluation dated February 11, 1987. No significant hazards consideration comments received: No.

Local Public Document Room locotiom Van Zoeren Library, Hope College. Holland. Michigan 49423; Duquesne Light Company, Docket No.

8M', Beaver Valley Power Station, Unit No. 1, Shipplagport, Pennsylvania Date ofapplication for amendment:

October 21. 1988.

Briefdescription ofamendment: The amendment changes the Technical Specifications for Beaver Valley Unit No. 1 to add requirements for independerit testing of the reactor trip breaker. undervoltage and shunt trip attachments, and bypass breakers.

Review of the design of these components has been documented in the staffs letters of November 8. 1984.

October 9. 1985. and April3. 198L Dote ofissuoncei February 4, 1981.

Effective dote: February 4, 1981.

Amendment No. 107 FocililyOperating License No. DPR-N. Amendment revised the Technical

~

Specifications, Dote ofinitiolnoticein Federal Register. December 17. 1986 (51 FR 45199) The Commission's related

'valuation of the amendinent is contained In a Safety Evaluation dated February 4.,1987.

- No significant hazards consideration comments received: No.

Local Public Document Room locations B. F. Jones Memorial Library.

863 Franklin Avenue, Aliquippa.

Pennsylvania 15001.

Florid Power and Light Company.

Docket Nos. 50-250 and 50-251, Turkey Point Plant Units 3 aad 4, Dade County, Florida Dote ofapplication foramendments:

August 25, 1988. as supplemented November 14. 1986.

Briefdescription ofamendments: The seiendment's modify the technical specifications Li'initingConditions of Operation'(LCO) for a one-time additional 45 day LCO in cori)unction neith the current 3'h day LCO to permit Iinplementation of modifications required to satisfy NUREG-0737, item I.D.3.4. Control Room Habitability.

Date ofissaancei February 2. 1987.

'ffective dale: February 2. 1987.

Amendment Nos. 12l and 115.

Facility Operating Licenses Nos.'PR-3t ond DPRwti Amendments ievised the Technical Specifications.

Date ofinitialnoticein Federal Register. December 30. 1988 (51 FR

<>078). The Commission's related evaluation ofthe amendments fs contained ln a Safety Evaluation dated February 2, 1987.

No significant hazards conslderatloa comments received: No Local Public Document Room locotion: Environmental and Urban Affairs Library, Florida international University, Miami, Florida 33199, Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee Atomic Power Station, Lincoln County, Maine Date ofapplication foramendment:

November 25, as supplemented December 8. 198L Briefdescription ofomendment: This amendment revises the Technical Srecifications to increase the fuel enrichment limitto 3.5 weight percent Uranium-255 for Cycle 10 and beyond.

Dote ofissuance: February 9, 1987.

Effective Dote: February 9, 1987.

Amendment No.: 92.

Facility Operating License No. DPR-38: Amendment revised the Technical Specifications.

Dote ofinilialnolicein Federal Register. December 30, 1M6 (51 FR 47012 at 47081.

The Commission's related evaluation of the amendment ls contained in a Safety Evaluation dated February 9, 1981.

No significant hazards consideration comments received: No. Local Public Document Room location: Wiscasset Public Library. High Street, Wiscasset,

Maine.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear Station, Nemaba County, Nebraska Dote ofamendment n'iiuest; November 6, 19'rief description ofamendment The'inendment changes the Technical Specifications to state that the Main Steam Drain Header Isolation Valves are normally open.

Dote ofissuance: February 3. 1987.

Effeclive dale: February 3, 1987.

Amendment No.: 101.

Facility Operating License No. DPR-

&P. Amendment revised the Technical Specifications.

Dote ofinitialnoticein Federal Register. December 17, 1988 (51 FR 452111).

The Commission's related evaluation of the amendment Is contained In a-Safety Evaluation dated February 3, 1987.

No significant hazards consideration comments recei vedi No, LocolPublic Document Room locotion: Auburn Public Library, 118 15th Street, Auburn. Nebraska 88305.

Niagara Mohawk Power Corporation, Docket No, 44420, Nine Mlle Polat Nuclear Station, Unit No.1, Oswego County, New York Dote ofomendment request: August 22, 1988.

Briefdescription ofamendment: The amendment adds surveillance requirements to Technical Specification Section 4.4.5 to verify the ability of the control room air treatment system to maintain a positive pressure In,the control room to maintain control room habitability during accident conditions.

Date ofissuance: February 10, 1987.

Effective date: February 10. 1981.

Amendment ¹, 91.

FacdRy Operating License Nq, DPR-

83. Amendment revised the Technical Specifications.

Date ofinitialnoticein Federal-Register: October 8, 1988 (51 FR 3809').

The Commission's related evaluation

'f the amendment ls contained in a Safety Evaluation dated February 10,

1987.

No significant hazards consideration comments received: No.

LocolPiiblic Document Room location: State University of New York, Penfield Library, Reference and Documents Department. Oswego. New York 13126.

Northeast Nuclear Energy Company, et al Docket No. 50-338, Millstone Nuclear Power Station Unit No. 2, Town ofWaterford Connecticut Date ofopplicotion foramendment:

October 29.

19'rief description ofamendment: The amendment changes the Technical Specifications (TS) for Millstone Unit 2 related to primary system iodine spiking: $1) The requirement In TS 3.4.8.

"Specific Activity",that operation, outside predetermined primary coolant activity limits. not exceed 10 percent of the unit's total yearly operating time ls deleted; and (2) the current "special report". requirement in TS 6.9.2i associated with primary coolant activity is replaced by an annual report requirement in TS 6.9.1.4.

Dote ofissuance: February 3. 1987.

Effective date: February 3, 1987.

Amendment No.: 11 S.

FocilityOperating License No. DPR-

85. Amendment revised the Technical Speciiiication.

Dote ofiniliolnotice in Federal Register. December 30, 1988 (Sl FR 47072 at 47083).

'The Commission's related evaluation of the amendment Is contained in a Safety Evaluation dated February 3, 1987.

5876 Federal Register / VoL 52. No. 38 / Thursday, February 28, 19N' Notices No significant hazards caciskieraOoo cumsincnts recetvccL No.

Local Public Dacamctit Roam locctiont Waterford pubhc Library, Rope Ferry Road, WaterforcL Connecticut.

Pccific Gas and Electric Company, Docket Nos. 50-275 and 56-323, Diablo Csnyi n Nuclear Power Plant. Unit Nee.

1 and 2, San Luis Obispo County Cahfornia Date ofapplication far omcndmentst Jull 18. 1988 (LAR86-08).

Briefdcscrtption of amendmentst Tcchnical Specification change will implement axial offset control (RAOC) for Diablo. Canycm Unit? after burnup of rele xed 8000 MWD/MTUin firsl cycle.

Date afissuance: January 3L 1987.

E."ective dater January 30, 1987.

Amendment Nas.: 12 and M.

Facility Operating Licenses ¹s.

DPR-tt0 and DPR-ttgt Amendnicnts revised the Technical Specifications.

Date ofinitialaaticein Federal Register. Oclobcr 22, 1988 (51 FR 31518).

The Commission's related evaluation of the amendmcnts is contained in a Safely Evaluation dated January M.

1987.

Nu signifi&snt huziirds uunsiucratiuii comments received: No.

Local Public Document Room location: Califonda Polytechnic State University Library. Documents and M"ps Dcpartincnt. San Luis Obispo, Cab fornta 93407.

Padfic Gas and Electric Company, Docket No.60-133, Humboldt Bay Power Plant. Unit No.3. Eureka, California Date ofapplication foramendment.

May 14. 1986 as revised November 1'1, 1966.

Briefdescriptt'an ofamendment The ecnenchnent estebhshcs physical security requirements for the shutdown Hicnibaid Ba) ~ Jrii ~ 3 'aiC Seiosaa

~ sas'Psaaa reflects the possess-bul-not-operate status of Fe'cility Operating License No.

DPR-7.

Dote ofisintanccc F@rua ry 11, 1967.

E%cti ve Dater When the revised physical securit plan is fully implemented but in no case morc than 90 days from the date of issuance.

Amendment hb 22.

Facility Operating License Na. DPIt-

7. Amendmcnt revised the license.

Dote ofinitialnoticein Federal Register. June 4. 1986 (51 FR 20371).

The November 1I. 1986 letter provided clarifymg information and djd not change the initialdctcrminahun published in the Federal Regular. The Comruission's related evaluation of tbe amendment is ccxitained in a Safety Evaluation dated February 11.1987.

No sitfrdficant hazards'consideration coaunents received: No.

Local Public Document Room location Eared umboldt County Library. 471 I Street (Ccnmty Court House), Eurekas California 95601.

Power Authorityof the State of New York. Docket No. 60-$33/i James A.

FitxPatrick Nuclear Power Plant, Oswego Cocmty, New York Date ofapplicatian faramendment October 8, 1986.

Briefdescription ofamendment: The amendmcnt changes the Technical Specifications to eliminate operabifiity and testing requirements for the recirculation pump discharge bypass valves. The bypass valves and lines are to be removed during the 1987 refueling outage.

Date ofissaancc: Febraary 9. 1981.

Bffective dote.. February 9, 19ty7.

Amendment Na, 104.

Facility Opera titigLicense Na. DPR-6R Amendment revised the Technical Spcctficethms.

Date ofinitialnotice iti Federal Register. November 19, 1986 (51 FR 418811.

The Commission's related evduation ofthe amendment is contained in a Safety Evaluathn dated February 9.

1987; l

No significent haxards;consideration comments received: No.

LocalPublic Document Room lacationt Penfleld IArary. State University College ofOswego, Oswego.

New York Power Authority of the State ofNew York, Docket No. 50-333, James A.

FltxPatrick Nuclear Power Plant, Oswego Ceanty, New Yerk Date ofapplicatian foromendmenta July 11. 1988.

Briefdescription ofamendment. The amendment changes the Administrative Controls section ofthe Technical SpectficaUana to rcQect the management reorganization ofthe Power Authority of the State ofNew York'a headquarters stafL Date oft'ssuance: February '10; 1981.

Effective date: February 10, 2987.

Ameacfmeat Na '106.

Facility Operating Licetee Na DPR-SR Amendment revised the Technical Speciflcatkgis.

Dote ofinitialnoticein Federal Register. October 8, 1988 (5l FR 38102).

The Commission's related evaluation of the amcndmeat hi contained in a Safety Evaluation dated February 10, 1987.

No significant hazards consideration comaicnts received: Na LocolPublic Document Room locati ant Penficld Library, State University College of Oswego, Oswego, New York.

Public Sersdce Hectric 4 Gas Company, Docket No. 50-354 Hope Creek Generating Statimii Salem County, New Jersey Date ofapplication foramendmena September 12. 1988 as supplemented on Scplember 22. and November 10, 1988.

Briefdescription ofamendment: This amendment reviscs the Technical Specifications to be consistent with the Standard Radiological Hfluent Techmcal Specifications for BWRs by permitting continued plant operation

~hould certain gaseous and liquid radioactive effluent monitors be inoperable. provided regular grab samples are collected and analyzed to be below specified limits.

Dote afissuance. February 6. 1987.

Effective date: February 8, 1987.

Amendment Na, 2.

Facility Operatic License Na NPF-57: Amendment revtses the Technical Specifications.

Date ofinitialnoticein the Federal Register. November 19, 1986, (51 FR 41867). Thc Commission's related evaluatkin of the amendment is contained in a Safety Evaluation dated February 8. 1967.

No significant hazards consideration conmumts received: Na LocolPublic Document Room locotiom Pcmuivtlle Public Library. 190 South Broadway. Pennsville. New Jersey 08070.

Rochester Gas and Electric CorporatIoa.

Docket No.60-244, ILF Girna Nuclear Power Pfant, Wayne County, New York Dote ofapplication foramendmcnt:

October 24. 1986.

Briefdescrrption ofamendment: The licensee plans to replace the analog rod position indication (ARPI) system with a Westinghouse microprocessor rod position huflcatton (MRPI) system. The ARPI system is being replaced because the system requires a significant effort to maintain alignments, the aging system is prone to component failures. and spare parts are difficultto obtain. In addition. the replacement of the ARPI system willresolve human engineering discrepancies identified during the detailed cmtrol room design review; a licensing action of NUREG-0737 (Item I.D.1.1).

As a result of the replacement of the ARPI system with the MRPI system, the licensee has proposed changes to the technical specification (TS) that (1) replaces the references to the ARPI

Federal Reg(ster / Vo). 52, No. 38 / Thursday, February 26, 1987 / Notices system with references to the MRPI system. (2) clariflee the requirements related to the Indicated positions versus demand positions, (3) allows the use of the process plant computer system ae a

. backup to the MRPI cathode ray tube (CRT) iflhe CRT should become inoperable, (4) remove the calibration requirement which is no longer necessary when the new system ls operational. and (5) modify the rod movement test.

Date ofissuance: February 10, 1987, Effective date: February 10, 1987.

Amendment No, 22.

Facility Operating License No. DPR-1gr Amendment revised the Technical Specifications.

Date ofinitialnotice in Federal Register: November 19. 1988 (51 FR 41868). The Commission's related evaluation of the amendment is contained in a Safely Evaluation dated February 10, 1987. No significant hazards consideration comments received: No.

Local Public Document Room, lacotion: Rochester Public Library, 115 South Avenue, Rochester, New York 14810.

Sacramento Municipal UtilityDistrict, Docket No. 50-312, Rancho Seco Nuclear Generating Station, Sacramento County, California Date ofopplicatian for amendment:

January 24, 1988, es supplemented September 19, 1988.

Briefdescription ofomendment: This amendment increased the setpoint for

'eactor trip on high pressure from 2300 psig to 2355 psig and Increased the arming threshold for anticipatory reactor trip on turbine trip from 20% to 45% of full power.

Date ol issuance: February 3. 1987.

Effective date: February 3. 1987.

Amendment Not 83.

Facility Opera'ting License No. DPR-

54. Amendment revised the Technical Specifications.

Date ofinitiol notice in Federal Register. November 5, 1986 (51 FR 40281).

, The Commission'e related evaluation of the amendment is contained In a Safety Evaluation dated February 3, 1987.

No signiflcanl hazards consideration comments received: No.

Loco/ Public Document Room locotian.. Sacramento Clty<ounty Library. 828 I Street, Sacramento.

California 95814.

Tennessee Valley Authority. Dockete Nos. 50-259, 50-280 and 50-296, Browne Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, Alabama Dates ofopplicotions for omendments: September 30, 1986.

Briefdescription ofamendments: The amendments change the Technical Specifications to add requirements for Radiological Effluent Technical Specifications (REfS) which comply with Section V of Appendix I to 10 CFR Part 50.

Dote ofissuance: February 5. 1987.

Effective date: February 5, 1987.

Amendments Nas.: 132, 128 and 103.

Facility Operating Licenses Nos.

DPR33. DPR-52 and DPR~'mendments revised the Technical Specifications.

Date ofinitialnoticein Federal Register. December 3. 1986 (51 FR 43685).

The Commission's related evaluation of the amendments is contained in a Safety Evaluation dated February 5.

1987.

No signiTicant hazards consideration comments received: No.

Local Public Document Room locatioru Athens Public Library, South and Forrest, Athens, Alabama 35611.

Tennessee Valley Authority, Docket Noe. 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee Date ofapplication for omendments:

February 1, 1985. as supplemented August 8, 19M.

Briefdescription ofamendments: The amendments change the Technical Specifications to delete a surveillance requirement. add a footnote to allow a reduction in fast start/fast load surveillance testing and change the reporting requirements of Emergency Die e! Generators, and correct an editorial mistake.

Dote ofissuance: February 3, 1987.

Effective date: February 3. 1987.

Amendment Nos.: 52 and 44.

Faci%ty Operating License Nos. DPR-77 and DPR-7R Amendments revised the Technical Specifications.

Dote ofinitialnoticein Federal Register: June 4, 1985 (50 FR 23554) end December 3. 1988 (51 FR 43885).

The Commission's related evaluation of the amendments is contained in'a

'afety Evaluation dated February 3, 1987.

No significant hazards consideration comments received: No.

LocolPublic Document Room Lacatioru Chattanooga-Hamilton County Bicentennial Library. 1001 Broad Street.

Chattanooga, Tennessee 37401.

Tennessee Valley Authority, Docket NoL 50-327 and 50-328, Sequoyah Nuclear Plant, Units 1 and 2, Hamilton County'ennessee Dote ofapplicotion foramendments:

March 4, 1983, as supplemented September 30, 1988.

Briefdescription ofamendments: The amendments change the license conditions related to implementation of Regulatory Guide 1.97, Revision 2 Date ofissuance: February 12, 1987.

Effective dote: February 12, 1987.

Amendment Nos.: 53 and 45.

Facility Operating License Nas. DPR-77 ond DPR-79. Amendments revised Operating License conditions..

Dote ofinitialnoticein Federal Register; October 22, 1988 (51 FR 37521).

The,Commission's related evalua(fon of the amendments is contained in a Safety Evaluation dated February 12, 1987.

No significant hazards consideration comments received: No.

Local Public Document Room Location: Chattanooga-Hamilton County Bicenten~al Libroru 1M1 Broad Stroot Chattanooga, Tennessee 37401.

Toledo Edison Company and The Cleveland Electric Illuminating Company, Docket No. 50-348, Davis-Besse Nuclear Power Station, Unit No. 1, Ottawa County, Ohio Dole ofapplication foramendment:

May 28, 1985.

Briefdescription ofamendment. This amendment reduces the frequency of cold fast starts required for surveillance testing of emergency diesel generators.

A'ortion of the amendment request has been denied by the Commission. A Notice of Denial is being published separately In the Federal Register.

Date ofissuance: February 10, 1987.

Effective dote: February 10. 1987.

Amendment No. 97.

Facility Operating License No. NPF-3.

Amendment revised the Technical Specifications.

'Date ofinitiolnoticein Federal Register. December 4. 1985 (50 FR 49793).

The Commission's related evaluation of the amendment is contained, in a Safety Evaluation dated February 10, 1987.

No signiflcant hazards consideration comments received: No.

Local Public Document Room locotioru University ofToledo Library, Documents Department, 2801 Bancroft Avenue. Toledo. Ohio 43606.

Federal Register / Vol. 52. No. 38 / Thursday, February 28, 2987 / Notices Virghiis Electric and tower Com pany, Docket Nos. 50-280 and 50-281. Sany Power Statai ~, Ualt Nos. 1 and 2, Surry County, Virginia Date ofapplication for amendmcnta June 18. 1986.

Btiufdescription ofamendments:

These amendments revise the audit freqiiency of the Surry Power Station Security Plall from at least ollce pef 24 months to at least once per 12 months.

Dote ofissualice: February 3, 1987.

Effectiva data. February 3, 1987.

Amendment Nus. 212 and 112.

Facility Operatiatt License Nas. DPR-32 and DPR&77 Amendments revised the Technical Specifications.

Dah'finitialnotice m Federal Register. July lb. 1986 (51 FR ~2).

The Commission's related evaluation of the amendinent ts coatained in a Safety Evaluation dated February 3.

1987.

No significant hazards consideration comments received: No Local Public Room location: Swem Library. College of Williamand Mary.

Williamsburg. Virginia 23185.

Wisconsin Public Service Corporation, Docket No. 50-805. Kewaunee NucIear Power Plait, Kewaunee County, Wisconsin Date ofapplication for aniendment:

September 30, 1986.

Briefdescription ofamendment: The amendment removed requirements from the Technical Specifications (TS) that werc duplicated in the Operational Quality Assurance Program. In addition.

a recent change in a taanagement title was made on three pages of the TS.

Date ofissuance: February 3. 1987.

'ffective dole: February 3, 1987.

Amendment No.: 72.

Facility Operatinct License No. DPR-

43. Amendment revised the Technical Sppcificalions.

Dole ofinitiulnoticein Federal Register. December 27. $988 (51 FR 45216).

The Commission's related evaluation of the amendment is contained in a Safety Evaluation dated February 3,

1987, Nn signiiicacit hazards consideration comments received: No.

Local Public Document Room locat:on.. University of Wisconsin Library Learning Center. 2420 Nicolet Drive. Creen Bay. Wisconsin 543trt.

Wisconsin Electric Power Company.

Docket Nos. 50-268 and 50401. Point Beach Nuclear Plant. Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc County. Wiisconsia Date ot oppbcotiua for amendments.

August 29. 1886.

Briefdescription ofamendmentst The amendments revise the Limiting Conditions for Operation (LCO) af the component cooling water system to reflect the addition of a fowth heat exchanger to the system.

'ote ofissuance: February 2. 1987.

Effective dote: 20 days from the date of issuance.

Amerrdment Nos.r105 and 108.

FacilityOperating License Nos. DPR-24 ond DPR-27. Amendments revised the Technical SpeciTications.

Dote ofinitialnoticein Federal Register. October 8, 1986 (51 FR 38108).

The Comrmssion's related evaluation of the ainendments is contained in a Safety Evstuation dated February 2, 1987.

No significant hazards consideration comments received: No.

Local Public Document Room locatiaru Joseph P. Mann Library. 1516 Sixteenth Street. Two Rivers, Wisconsin.

NOTICE OF ISSUANCE OF AMENDMENTTO FACILITY OPERATING LICENSE ANDFINAL DEfEIIMINATIONOF NO SIGNIFICANTHAZARDS CONSIDERATION AND OPPORTUNITY FOR HEARING (EXIGENTOR EMERGENCY CIRCUMSTANCES)

During the period since publicatioa 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 for the aniendment compiles with the standards and requirement ~ of the Atosaic Energy Act of 1954, as amended (the Act. and the Commission's rules and regulations.

The Commission has made appropriate findings as required by the Act and the Conimission's rules and regulations in 10 CFR Chapter L 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 Commission to publish.

for public comment before issuance, its usua130-day Notice of Consideration of Issuance of Amendment and Proposed No Significant Hazards Consideration Determination aad Opportunity for Hearing. For exigent circumstances.

the Commission has either issued a Federal Register notice providing opportunity for public comment or has used local media to provide notice to the public in the area surrounding a licensee's facilityof the licensee's apphcstion and of the Commission's proposed determination of no significant hazards consideration.

The Conuaission has provided a reasonable opportunity for the public to comment. using Rs best efforts to make available to the public means of communication for the public to respond quickly. and in the case of telephone comments, the cocnments have been recorded or transcribed as appropriate and the licensee has been informed of the public comnients.

In circumstances where failure to act In a timely way would have resulted. fur 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. Ifthere has been some time for public comment but less than 30 days, the Commission may provide an opportunity for public comment. Ifcomments 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 immediately 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 Commissioti has applied the standards of 10 CFR 50.92 and has made a fiiialdetermination that the amendment involves no signiTicant hazards consideration. The basis for this determination Is contained in the documents related to this ection.

Accordingly. the amendments have been issued and made eiTective as indicated.

Unless otherwise indicated. the Commission has determined that these amendments satisfy the criteria for categorical exclusion in accordance with 10 CFR 5122. Therefore. pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared for these amendments. Ifthe Commission has prepared an environmental assessment under the special circumstances provision in 10 CFR 51.12(bl and has made a determination based on that assessraent, 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 andi or Environmental Assessment.

as indicated. Allof these

Federal Register / Vol. 52, No, 38 / Thursday, Febrttary 26, 1981 / Notioes 58/9 items are available for public inspection at the Commission's Public Document Room. 1717 H Street NW Washington, DC. and at the local public document room for the particular facilityinvolved, A copy of items (2) and(3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission.

Washington. DC 20555. Attention:

Director, Division of Licensing.

The Commission is also offering an opportunity for a hearing with respect to the issuance of the amendments.

By March 30. 1987. the licensee may file a request for a hearing with respect to issuance of the amendment to the subject facility operating Bcense and any person whose interest may be iffccted 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'ith the Commission's "Rules of Practice for Domestic Licensing I'roceedinqs" in 10 CFR Part 2. Ife request for a hearii i or petition for leave to intervene is filed by the above date. the Commission or an Atomic Safety and Licensing Board. designated by the Commissio'n or by the Chairman of the Atomic Safety and Licensing Board Panel, willrule on the request and/or petition and the Secretary or the designated Atomic Safety and Licensing Board willissue a notice of hearing or an appropriate order.

As required by 10 CfR 2.714, 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!he 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. or other interest in the proceeding; 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 rnatter of the proceeding as to which petitioner wishes to intervene.

Any person who has filed a petition for leave to intervvne or who has been'dmitted as a party may emend the petit:on without requesting leave of the Board up to."fteen (15) days prior to the first prehearinq conference scheduled in the proceeding. but such an amended petition must satisfy the ipecificity requirements described above.

Not later than fifteen (15) days prior to the first prehearinq 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 bo 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 willnot 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 fullyin the conduct of the hearing. Including the opportunity to present evidence and cross-examine witnesses.

Since the Commission has made a final determination that the amendment involves no signiifiicant hazards consideration. ifa hearing is requested.

it willnot stay the effectiveness of the amendment. Any hearing held would take place while the amendment ls In effect.

A request for a hearing or a petition for leave to intervene must be filed with the Secretary of the Commission. U.S.

Nuclear Regula tory Commission, Washington. DC 20555, Attention:

Docketing and Service Branch, or may be delivered to the Commission'a Public

- Document Room. 1717 H Street NW.,

Washington. DC, by the above date.

Where petitions are filed during the last ten (10) days of the notice period, it is requested that the petitioner prompt)y 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 ldenfification Numbar 3737 and the following message addressed to (Project Director)t 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 Office of the General Counsel. Bethesda. US. Nuclear Regulatory Commission, Washington, DC 20555. and to the attorney for the;

'icensee.

Nontimely filings of petitions for leave to intervene. amended petitions.

supplemental petitions and/or requests for hearing willnot be entertained absent a determination by the Commission. the presiding officer or the Atoinic Sa!ety and Licensing Board, that the pettttoa and/or request should be granted based upon a balancing of the factors specified In 10 CFR 2.714(a)(1)(l)-

(v) and 2.714(d).

Florida Power aad Light Company, Docket No. 50-2$ 0, Turkey Point Plant Unit 8, Dade C'a'bnty. Florida Dale ofapplicatiatt foramendment:

February 4, 1987:

Briefdescription ofamendment. The amendmarit extends the surveillance testing requirements ofTechnical Specifications 4.7.1.1 and 4.7.1.2.a relating to the Emergency Containment Filter System to allow the tests to be performed during the Unit 3 refueling outage for Cycle 11.

Date ofissuance: February 12, 1987.

Effeati ve dote: February 12. 1987..

Amendment No '22.

docilityOperating License Na. DPR-31t Amendment revised the Technical Specifications.

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

The Commission's related evaluatioa of the atnendment and final determination of no slgnificant hazards consideration are contained ln a Safety Evaluation dated February 12. 1987.

AttorneyforLicensee: Harold F. Reis.

Esquire. Newman and Holtzer. P.C.. 1615 L Street NW.. Washington, DC 20M6.

Local Public Document Room location: Environmental and Urban Affairs Library. Florida international University, Miami, Florida 33199.

NRC Project Diteator. Laster S.

Rubensteln.

Dated at Bethesda. Maryland this 19th day of February. 1987.

'or the Nuclear Regulatory Commission.

Thomas M. Novak, ActingDirector. Divisiono/PWR Licensing-A.

(FR Doc.87-38M Flied 2-~; L45 am) altlÃlo coos 75$04iw

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