ML20153G913

Informs That Generic Ltrs 85-14,85-17 & 85-18 Applicable to Facility Per Review of All Generic Ltrs Issued from Jul-Dec 1985.Generic Ltr 85-21 Not Yet Issued.Generic Ltrs Encl.No Responses Required

ML20153G913
Person / Time
Site:	Three Mile Island
Issue date:	02/18/1986
From:	Travers W Office of Nuclear Reactor Regulation
To:	Standerfer F GENERAL PUBLIC UTILITIES CORP.
References
CON-NRC-TMI-86-018, CON-NRC-TMI-86-18 GL-85-14, GL-85-17, GL-85-18, GL-85-21, NUDOCS 8602280529
Download: ML20153G913 (2)

Text

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i;P g DISTRIBUTION:

Docket No. 50-320 ly .

NRC PDR Local PDR

!;i DCS TMI HQ r/f TMI Site r/f a February 18, 1986 WDTravers 4 MTMasnik 1 HRC/THI-86-018

}.. RHall Docket No. 50-320 Acting Chief, Tech Support l~- CCowgill

. RCook Mr. F. R. Standerfer LChandler, ELD

!,* Vice President / Director IE

Three fille Island Unit 2 ACRS

C GPU Nuclear Corporation M-Town Office P.O. Box 480 lT Middletown, PA 17057 l -

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Subject:

Generic Letters - July 1,1985 through December 31, 1985

, j' The THI-2 Cleanup Project Directorate has reviewed all generic letters issued from July 1,1985 through Dece:r.ber 31, 1985, copies of which are enclosed. Of

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the nine generic letters issued during this period, we have detemined that three are applicable to the TMI-2 facility: generic letters numbered 85-14,

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'l 85-17, and 85-18. Note that generi: letter 85-21 has not yet been issued.

We have received your response to generic letter 85-18. No other responses are required. We suggest that you review all generic letters for your

,h information. If you have any questions with regard to applicability or l{ compliance, please contact Michael T. Masnik of my staff at (301) 492-7743.

l Sincerely, l CEfeem meep w,

( WWom D. Tmves William D. Travers lc Director

)p TMI-2 Cleanup Project Directorate

Enclosures:

Generic Letters 85-13

( through 85-20 and 85-22 cc: T. F. Demitt R. E. Rogan S. Levin W. H. Linton J. J. Byrne A. W. Miller Service Distribution List (see attached) w/o Encis.

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n:c ronu sie no-so Nacu ono OFFICIAL RECORD COPY c _ _ - _ _ - _ - _ - _ _ _ _ _ _ - _ _ _

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s Dr.sThomas Murl:y Elllis.Siiby. Site Manager Regional Adailistrator. Region I U.S. Department gf [nergy U.S. huclear Regulatory Cosumission P.O. pos 33

. 631 Park Avenue Middletoun. PA 17057 0311 King of Prussia. PA 19406 .

John F. Wolfe. Esq.. Chairman. Division of Three M11e Island Programs Administrative Judge hf-23 3409 Shepherd 5t. U.S. Deper taent of Energy Chevy Chase. MD. 20015 Washington. D.C. 20545 Dr. Oscar H. Paris William Lochstet .

Administrative Judge 104 Davey Laboratory Atomic Safety and Licensing Pennsylvania State University Board Panel University Park. PA 16802 -

U.S. huclear Regulatory Commission Washington D.C. 20555 Randy Myers. Editorial The Patriot

• Dr. FrederictL H. Shon

• 812 Market St.

Administrative Judge . Marrisburg. PA 17105 Atomic Safety and Licensing Board Panet Robert B. Borsue U.S. huclear Regulatory Coavaission Baltock & Wilcom hashington. D.C. 20555

• huclear Power Generation Division Suite 220 .

Karin W. Carter 7g10 Woodmoent Ave.

Assistant Attorney General Bethesda MD. 20814 505 [secutive House P.O. Som 2357 Michael Churchh111. Esq, Harrisburg. PA 17120 PitCOP 1315 Walnut St.. Suite 1632 Dr. Judith M. Johnsrud Philadelphia, PA 19107 (aviroevnental Coalition on

. Nuclear Po er Linda W. Little 433 Orlando Ave. 5000 Hermitage DR.

State College. PA 16801 Raleigh.hC 21612 i

George F. Trowbridge. Esq. Marvin 1. Lewis sha. Pittman. Potts and 6504 3radford Terrace

, Trowbridge Philadelphia. PA 19149 1800 M. St.. kW.

• Washington, D.C. 20036 Jane Lee 183 Valley Rd. '.

7 Atoalc Safety and Licensing Board Panel [tters.PA 1731g U.S. huclear Regulatory Consnission Washington, D.C. 20555 J.B. Liberman. Esquire

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• Berlack lsraels. Liberman Atomic Safety and Licensing Appeal Panel 26 droaduay U.S. haclear Regulatory Comission heu fort, my 10004 Washington. 0.C. 20555 Secretary Walter W. Cohen. Consumer Advocate Department of Justice U.S. hoclear Regulatory Comission Strawberry Square.14th Floor ATTN: Chief. Docketing 8 Service Branch Harrisburg. PA 17127 washington. D.C. 20555 Mr. Larry Hochendoner (d.ard O. 5-art Cauphin County Comissioner Board of Supervisors P.O. Bos 1295 Londonderry Township Harrisburg.~ PA 17108-1295 RFD fl Geyers Church Rd.

Middletown PA 17057 John [. Minnich. Chairpersor. Robert L. Knupp. [ squire Dauphin County Board of Convsissioners Dauphin County Courthouse Assistant Solicitor Knupp and Andrews Front and Market Streets P.O. Bos P Marrisburg, PA 17101 *

• 407 N. Front St.

Harrisburg. PA 17108

, D,auphin County Office of Emergency Preparedness John Levin. (squire

• Court House. Room 7 Front & Market 5treets Pennsylvanf a Public utilities Com.

Parrisburg. PA 17101 P.O. Bch 3265 Harrisburg. PA 17120 U.S. (nvironmental Protection Agency 8egion !!! Office ATTh: II$ Coordinator Curtis 6ullding (51sth Floor) ~

6th & Walnut Streets '

Pnfladelphia, PA 19106 Mr. te in tintner' (secutive Vice President Thomas M. Cerusly. Director General rublic Utilities buclear Corp.

Bureau of Radiation Protection 100 Ir.terpace Park-ey Department of f avironmental Resources Persip;,any. hJ 07054 P.O. 6os 2063 marrisburg PA 17120 Le'. Lennecy Ae treed.

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• August 5, 1985 i

TO ALL REACTOR LICENSEES AND APPLICANTS

.s Gentlemen:

SUBJECT:

TRANSMITTAL OF NUREG-1154 REGARDING THE DAVIS-BESSE LOSS 1

- OF MAIN AND AUXILIARY FEEDWATER EVENT (Generic Letter No. 85-13)

• On June 9, 19.85 Toledo Edison Company's Davis-Besse. Nuclear Power Plant --

"expelFienced a 'lo,ss of all feedwater event while the plant was operating at 90%

power. Shortly after the event, the NRC Executive Director for Operations i directed that an NRC Team be sent to Davis-Besse, in conformance with the staff-proposed Incident Investigation Program, to investigate the circumstances 1

. of this event.

i The NRC Team has now completed its investigation and has documented the factual information and their findings and conclusions associated with the event (see enclosed NUREG-1154, entitled " Loss of Main and Auxiliary Feedwater Event at i the Davis-Besse Plant on June 9,1985"). The report indicates that a total loss of feedwater is a significant event; and that it can have severe consequences if actions to ensure prompt and effective recovery are not taken. The conse-quences and significance of the June 9 event could have been far different had i

l additional equipment failed, had additional errors been made, or had recovery ,

otherwise been delayed. Thus, there are many possibilities 'and differing sequences which could have affected the safety significance of this transient.

In tenns of their principal conclusion, the team concluded that: the underlying cause of the loss of main and auxiliary feedwater event of June 9,1985, was the licensee's lack of attention to detail in the care of plant equipment; the 4

licensee's history of performing troubleshooting, maintenance and testing of I

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equipment, and of evaluating operating experience related to equipment in a superficial manner and, as a result, the root causes of problems are not always

' found and corrected; engineering design and analysis effort to address equipment problems has frequently either not been utilized or has not been effective; and that equipment problems werc riot aggressively addressed and resolved beyond i 4 compliance with NRC regulatory requirements.

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[ You should review the infonnation for applicability to your facility. In l addition, you should ensure that the information in NUREG-1154 is mode available to your plant staff as part of your training program in connection with the Feedback of Operating Experience to Plant Staff (TMI Action Plan

! Item I.C.5).

i I. _ . - . . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ . _ , - ~ . _ . _ _ . . - _ - . _ . _ ~ ~ . - _ _ _ _

This generic letter is provided for information only, and does not involve any reporting requirements. Therefore, no clearance from the-Office of Management and Budget is required. The enclosed report is currently under NRC review.

Any generic requirements stemming from the report will be transmitted at a

. later date following completion of the appropriate procedural steps.

h . Thompson, M. ,

h i ctor D vi ion of Licensing

Enclosure:

NUREG-1154

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\ .... 8 August 1.1985 '

t TO ALL LICENSEES i

SUBJECT:

COPMERCI'At STORAGE AT POWER REACTOR SITES OF WASTE NOT GENERATED BY THE UTILITY (Generic Letter 85-14) '

Gentlemen:

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The Low-Level Radioactive Waste Policy Act of 1980 (P. L.96-573) assigned to '

i ~.__. the states the responsibility to provide for disposal of connercial low-level radioactive waste (LLW) generated within each state.

The Act envisioned.that all states would be within their borders by 1986.capable of providing for disposal of connercial LLW generated

"~~ Based on.the current status of state efforts and -

.the substantial time required to establish new disposal facilities, no new sites will be available for at-least several years. Due to the uncertainty of this situation and statements made by some officials of states within which currently

! . existing operating disposal sites may be sites are located. it appears possible that access to the' restricted.

1 While some licensees have taken steps to temporarily store LLW generated at their sites to alleviate any impact that limiting of access to disposal capacity may have on licensed operations pr 4

used only for interim contingency purpose,s. ovisions It is the for policy storing LLWNR" of the shculd thatbe

. i licenseesextent maxisum should continue to ship waste for disposal at existing sites to the practicable, i ,

. In anticipation of possible curtailment of access to existing disposal facili-

ties, interest generated is being within expressed in some states in commercial storage of LLW the states.

! desirable in states which have not resolved their low-level waste dis l problems,sites.

disposal connercial storage facilities, however, should not become de facto i NRC will require for comercial storage under its jiirisdiction -

j that, in addition to safe siting and operation, connitments and assurances be i made locations. for eventual disposition of all waste stored at connercial storage This includes provisions for repackaging (if necessary), transpor-tation and disposal of the waste, as well as deconnissioning of the facilities.

Some of the concepts for comercial storage involve using nuclear power reactor sites licensee. as connercial storage locations for LLW not generated by the utility As a matter of policy, the NRC is opposed to any activity at a  ;

nuclear reactor site which is not generally supportive of activities authorized

by the operating license or construction permit and which may divert the atten-t tion of licensee constructi6n management of the power reactor. from its primary task of safe operation or the exclusion area of a reactor site, as defined in 10 CFR 100.3Accordingl subject tohRC jurisdiction regardless of whether or not the reac(a), tor iswill be i

located in an Agreement 10CFRL50.15(a 1).

State, pursuant to the regulatory policy expressed in exclusion areas),(the licensing authority is in the Agreement S

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, s In order for NRC to consider any proposal for conenercial storage at a reactor '

site, including conenercial storage in existing low-level waste storage facili-ties. the NRC must be . convinced that no significant environmental impact will '

result and that the consnercial storage activities will be consistent with and i not cocipromise safe operation of the licensee's activities, including diverting reactor management attention from the continued safety of reactor operations.

A Part 30 license is required for the low-level waste storage and a Part 50 license amendment may also be required. The application must include:

By the utility '

A determination by the utility licensee that the proposed low-level waste comercial storage activities do not involve a safety or environmental -

question, and that safe operation of the reactor will not be affected. t In making this detennination, the licensee shall consider:

' ' ' Direct impacts of the connercial storage operation on reactor t Tl 'SpeYations"during normal and accident conditions;

- i Diversion of utility management and personnel attention from '

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' safe reactor operation; Contined effects of onsite and offsite dose during normal and accident conditions; ~ - - '

Influence on effectiveness of reactor emergency plans; Influence on effectiveness of reactor security plans; Financial liability provisions, including impact on indemnity coverage; and Environmental impact of the storage facility, including potential interaction with the generating station.

By the applicant (the utility or another person) j *

• i Infobnation relating to the safety of the consnercial storage operation; Infonnation relating to the environmental impact of the storage operation in sufficient detail to allow staff to establish the need for preparation of an Environmental Impact Statement;

! Financial assurance to provide for the corrnercial storage operation and decomisioning including any necessary repackaging, transportation and 1

disposal of the waste; and Written agreement from the jurisdiction responsible for ultimate disposal, the State, that provisions are sufficient to assure ultimate disposal of the stored waste.

The Office of Nuclear Reactor Regulation (NRR) will conduct an environmental

,' review and review the application to determine whether the low-level waste consnercial storage activities on a reactor site impact the safe operation of 4

the reactor. Following NRR review, the licensing authority for cocinercial

) storage on a reactor site under NRC jurisdiction (all lacatinns in non-Agreement

States and locations within reactor exclusion areas in Agreement States) is the Office of Nuclear Material Safety and Safeguards. The NRC will assess I

environmental impact and will issue an Environmental Impact Statement, if appropriate, in accordance with provisions of 10 CFR 51.20, 51.21 and 51.25 As part of the procedures, the NRC will provide notice in the FEDERAL REGISTER of receipt and availability of any application received for corrrnercial storage activities. Th6 public notice will also indicate the staff's intent regarding preparation of an environmental assessment and its circulation for public review and corment. An Environmental Impact Statement will most likely be needed based on the environmental assessment.

Because the NRC has not yet received or reviewed an application for a centralized cormercial low-level waste storage facility intended to store large amounts of LLW for five or more years, the NRC may consider applying the criteria described above to such cormercial storage facilities whether they be on a reactor site or not.

Interim storage of uttitty ifcensee-generated Li.W will continue to be considered according to the provisions stated in Generic Letter 81-38, dated November 10, 1981.

For additional Reactor information, please contact Frank Miraglia, Office.of Nuclear Regulation U. S. Nuclear Regulatory Comission, Washington, D.C. 20555

[ Telephone: (301)492-7980) or Richard Cunningham, Office of Nuclear Material Safety and Safeguards, U. S. Nuclear Regulatory Comission, Washington, D.C.

20555 [ Telephone: (301)427-4485).

Sincerely, 4.'Willi LL J. Dircks - ,

Executive Director l for Operations l l

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\.....) AUG 6 1985 TO ALL LICENSEES OF OPERATING REACTORS Gentlemen: -

SUBJECT:

INFORMATION RELATING TO THE DEADLINES FOR COMPLIANCE i

WITH 10 CFR 50.49, " ENVIRONMENTAL QUALIFICATION OF ELECTRIC EQUIPMENT IMPORTANT TO SAFETY FOR .NifCLEAR POWER PLANTS" (GENERIC LETTER 85-15)

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The deadline for compliance with 10 CFR 50.'49,' ~" Environmental Qualification of

, Electric Equipment Important to Safety for. Nuclear Power Plants".is specified.

in the rule as the date of the second refueling outage after March 31,1982 or March 31, 1985, whichever was earlier. Some plants have received extensions i

to these deadlines up to November 30, 1985. Where current extensions teminate prior to November 30, 1985, the delegation in 10 CFR 50.49(g) pemits the

' Director of NRR to act on further requests for extensions as long as--the new deadline is not be

, . exceptional cases, the fondComission

~ November itself30,1985. ~ Section may consider 50.49(g) and grant extensions states that "in j

beyond November 30, 1985, for completion of environmental qualification." The purpose of this letter is to advise licensees that it is the Comission's

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intention that extensions, will be granted only in rare circumstances and that '

enforcement action will be taken against }icensees that continue to operate their plants with unqualified equipment 3 beyond November 30, 1985, without -

i extensions approved by the Comission. l It is the Comission's intention that licensees which are not in compliance on November 30, 1985, and which have not.been given extensions either will have to either shut down br, if they have valid staff-approved justifications for contigyed operation, select to operate and face civil penalties of $5,000 per item - per day for each day after November 30, 1985, on which a licensee operates in noncompliance with the rule. For noncompliance identified after November 30, 1985, such fines may be made retroactive to November 30, 1985 for each day a licensee clearly knew, or should I; ave known, that equipment qualification was inconiplete. Some mitigation of any penalty may be considered based upon satisfaction of the following factors: l l

JJ For purposes of enforcement " unqualified equipment" means equipment for which there is not adequate documentation to establish thart l this equipment will perform its intended functions in the relevant environment.

2/ An item is defined as a specific type of electrical equipmet.6,

• designated by manufacturer and model, which is representative of all identical equipment in a plant area exposed to the same l environmental service conditions.

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Did the Itcensee toentify and promptly report the i noncompliance with 10 CFR 50.497 l' 2.

1 Did the ifcensee apply best efforts to complete i environmental qualification within the deadline? i 4

3.

Has the licensee proposed actions which can be expected i to result in full compliance within a reasonable time?

or other means) after NovemberFor equipment which is discovered (th '

to be in noncompilance.with the requirements of 10 CFR 50.4

-report the finding if the condition found meets the reporting criteria of 10 CFR 50.72 (Prompt. Notification) or 10 CFR 50.73 :(Licensee System).

Evaluations of the significance of and corrective action for 411 actual and potential noncompitances should be documented as should the :

circumstances of discovery of the noncompliance or suspected nonco i

! These documents should be retained in appropriate licensee .

files.

If equipment i

addressed in _the plant Technical Specification's~is found to be un j

its intended function during an accident because of equipment qualif problems, the Specifications. licensee is required to follow the provisions of the Techni l

enforcement is appropriate for noncompliance ,1985 identified i~

j Licensees desiring an extension beyond November 30, 1985, must submit an extension request at the theDirector.

earliestIE.possible date to the Comission w to the Director NRR and

• for any extension request beyond Nov1985, will The .

bebasis co,nsidered i

exceptional nature of the case, e.g. ember 30, 1985 must clearly idantify the j i

control, the licensee tiill not be in compliance with the rule on  ;

t the dateuntil operation when compliance compliance will be achieved; and a . justification for coi will be achieved. >

i This of the Office letter does not require of Management and Budget.any response and therefore does not n Coments on burden and duplication may New Executive Office Buf1 ding,20503 Washington, D.C.be direct( ,

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! Jane Axelrad for enforcement questions. questions, the staff contac (301)492-7415 4

and Ms. Axeirad can be reached on (301)492-4909 .

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Hu h L. Thompson r. irector D vi fon of Licensi cc: List of Generic letters i

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/ August 23, 1985 TO ALL LICENSEES OF OPERATING REACTORS AND APPLICANTS FOR AN OPERATING LICENSE (Generic Letter 85-16)

Gentlemen:

SUBJECT:

HIGH BORON CONCENTRATIONS Boron Injection Tanks were originally incorporated into Westinghouse plant designs as a means of mitigating steam line break events. To overcome the reactivity addition resulting from a rapid cooldown, a high concentration of boron in the fom of boric acid was used in the injection tanks. High concentrations of boron result in maintenance and operational burdens to licensees because of the need to prevent boron precipitation.

In the recent past, there have been incidents at operating reactor plants in which boric acid has crystallized in the internals of vital safety related pumps and piping, thereby rendering those systems inoperable. One example is an incident at Indian Point 2 on December 28, 1984 in which the safety injection system was inoperable because all three pumps in the system were frozen with crystallized boric acid.

Over the past several years, the analysis methods for calculating the consequences of a steam line break have improved. These revised calculations demonstrate that the negative reactivity that needs to be added is lower than originally thought and consequently the need for highly concentrated boron injection is reduced or eliminated. Many licensees with Westinghouse plants (e.g. Surry 1&2), have requested that they be allowed to either physically remove the boron injection tank from the safety injection piping, or at least reduce boron concentrations in the tank to the levels safely used in other sections of the safety injection '

pipir.g and refueling water storage tank (e.g., to 2000 ppm). To support their requests, licensees have submitted new analyses of the steam line break event that demonstrated the regulatory criteria (i.e., 10 CFR 100 guidelines dose values) were met. The staff has reviewed these analyses and granted these  ;

- requests.

In light of the safety risks inherent in the present system and these new calculations which show a reduced need for boron injection, the staff encourages you to reevaluate the need for maintaining high concentrations of boron in your boron injection tanks. In the event you perfom a reanalysis of the steam line '

break event or any other event which requires or assumes credit for boron injection, the staff is 41111ng to consider a relaxation of excess' conservatism i in your analyses, provided the relaxation can be justified. As a result, it may be possible to remove the boron injection tanks or reduce the boron concentration.

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o If you recuire any further information regarding this subject, please contact

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. Hu h L Thompson tor Di ion of Licensing i Of ice of Nuclear Reactor Regulation cc: List of Generic letters i

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August 23, 1985 l

1 TOANDALL LICENSEES OF OPERATING REACTORS,(GENERIC LETTER 85-1/JAPPLICAN HOLDERS OF CONSTRUCTION PERMITS Gentlemen:

SUBJECT:

AVAILABILITY OF SUPPLEMENTS 2 and 3 TO NUREG-0933, "A PRIORITIZATION OF GENERIC SAFETY ISSUES" This letter is to inform you that Supplements 2 and 3 to NUREG-0933, "A Prioritization of Generic Safety Issues," were published in January and July 1985, respectively. These supplements present the priority rankings for generic. safety issues related to nuclear power plants. The purpose of these rankings is to assist in the timely and efficient allocation of HRC resources i for the resolution of those safety issues that have a significant aotential for reducing risk. The safety priority rankings of HIGH, MEDIUM, LOW, and ,

DROP have been assigned on the basis of risk significance estimates, the ratio of risk to costs and other impacts estimated to result if resolutions of the safety issues were implemented, and the consideration of uncertainties and other quantitative or qualitative factors, as discussed in the Introduction of NUREG-0933. To the extent practical, estimates are quantitative. Although changes to nuclear power plant designs or operation are assumed, the sole purpose in prioritizing the issues is to estimate costs and other impacts that might result if such changes were implemented.

The priority rankings in the report are based on generic assessments, i.e.,

assessments of plants thought to be typical of a class of plants. Thus the safety significance of an issue might be greater (or less) for a specific plant because of features of the design or operation that are different from those assumed in the generic assessments.

The supplements contain: (1) the issues that have been resolved since the publication of NUREG-0933 Supplement 1 in July 1984; (2) the issues that have been removed from further consideration by the NRC (i.e., those issues with LOW or DROP priority rankings); and (3) the issues that are in the process of resolution (i.e. , those issues nearly-resolved or with HIGH or MEDIUM priority rankings).

NUREG-0933 and its 3 supplements may be purchased by calling (202) 275-2060 or (202) 275-2171 or by writing to the Superintendent of Documents, U.S Government Printing Office, Post Office Box 37082, Washington, D.C. 20013-7082, or the National Technical Information Service, Department of Commerce, 5258 Port Royal Road, Springfield, VA 22161. '

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1 This letter does not contain 'any new requirements or guidance for licensees of operating reactors, applicants for operating licenses, and holders of ,-

construction permit's. No reply is required. If you have any questions on the subject, please contact Ronald Enrit on (301) 492-4%0. .

d WJ yughI. Thompson, . , tor DiWyonofLicensing Ofiice s

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September 27, 1085 3 ,

1 TO ALL POWER REACTOR LICENSEES AND APPLICANTS FOR AN OPERATING LICENSE Gentlemen:

Subject:

OPERATOR LICENSING EXAMINATIONS (Generic Letter 85-18)

To update the schedules you provided in response to Generic Letter 85-04, please provide the best estimate of your needs for operator li. censing examinations for fiscal years FY 1986, FY 1987. FY 1988 and FY 1989 (October 1 to September 30 of each year). You are also requested to provide requalification examination schedules for this same time period. Please identify the dates you have scheduled requalification examinations and anticipated requests for licensing examination site visits and the number of examinations for each v! sit. This information is needed to plan our resource requirements consistent with your operator licensing needs.

Because of continued budget limitations, we will attempt to limit examinations to two visits per site each' year. To meet this goal the regional offices may be

, required to redistribute the requested facility operator'examin tion visits across the entire fiscal year to even out the examination work load and eliminate high demand periods. Therefore, your submittal of this schedule does not guarantee the number or date of examinations requested. However, a best estimate of your needs for examinations will allow us to propose budget modifications, if necessary. Please keep us informed of significant changes in your estimates as they occur, so we can keep our data base current.

Your schedules,. in the enclosed suggested formats, should be returned to Mr. Bruce Boger, Acting Chief, Operator Licensing Branch, AR-5221, Washington, DC 20555, with a courtesy copy to the appropriate Regional Administrator within 30 days of the receipt of this letter and in no event later than November 1, 1985. We appreciate your assistance. If you have any questions concerning this request or your response, please call Mr. Bruce Boger, Acting Chief, Operator Licensing Branch, at (301) 492-4868.

This request was approved by the Office of Management and Budget under clearance number 3150-0131 which expires August 31, 1988. Comments on burden l

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and duplication may be directed to the Office of Management and Budget.

Reports Management, Room 3208, New Executive Office Building, Washington, D.C. 20503.

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Sincerely,

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H kL. Thompson Y m /l/JrW irector D sion of Licensi

Enclosures:

1. Operating Licensing Examination Schedule

2. Requalification Examination Schedule ,

3. List of Generic Letters s

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Enclosure 1 ,

S OPERATOR LICENSING EXAMINATION SCHEDULE Facility NRC Region FY 1986 FY 1987 FY 1988 FY 1989

1. Date Date Date Date

1. R0

1. SR0

1. SR0 Upgrade

1. Instructor Certification

1. SRO Limited to Fuel Handling

2. Date Date Date Date

1. R0

1. SR0

' #SR0 Upgrade

1. Instructor Certification

1. SRO Limited to Fuel Handling

3. Date Date___ Date Date

1. R0

1. SR0

1. 5R0 Upgrade

1. Instructor Certification

1. SR0 Limited to '

Fuel Handling Please indicate initial cold license examinations by placing an asterisk (*)

by the date F1-ase indicate examinations intended to extend an operator's license to a set..ad or subsequent unit with two asterisks (**) (e.g., Unit

( One is in operation and Unit Two is approaching fuel load. Three RO l candidates with no previous license are to be examined on both Units.One and l Two and five R0 candidates with licenses on Unit One are to be examined to extend their licenses to Unit Two. Indicate (2/15/85. RO 3.5**)).

Enclosure 2 REQUALIFIChTIONEXAMINATIONSCHEDULE, Facility NRC Region FY 1986 ,

FY 1987 FY 1988 FY 1989

1. Date Date Date Date

2. Date Date Date Date

3. Date Date Date Date.

4. Date Date Date Date e

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! pity o UNITED STATES E\ > .- ( ,i NUCLEAR REGULATORY COMMISSION

.- WASHING TON, O. C. 20555

-l September 27, 1985

• $.'..~.[

TO ALL LICENSEES AND APPLICANTS FOR OPERATING POWER REACTORS AND HOLDERS OF CONSTRUCTION PERMITS FOR POWER REACTORS Gentlemen:

SUBJECT:

Reporting Requirements on Primary Coolant Iodine Spikes (Generic Letter No. 85- 19 )

Generic Letter No. 83-43 was issued on December 19, 1983, to provide guidance on Technical Specification revisions required as the result of the revisions to 10 CFR 50.72 (Imediate Notification Requirements of Significant Events at Operating Nuclear Power Reactors) and of implementation of 10 CFR 50.73 (Licensee Event Report System). That generic letter discussed changing the requirement from a Licensee Event Report to a Special Report for operating conditions where the specific activity limits of the reactor coolant are exceeded.

As part of our continuing program to delete unnecessary reporting require-ments, we have reviewed the reporting requirements related to primary coolant specific activity levels, specifically primary coolant iodine spikes. We have detemined that the reporting requirements for iodine spiking can be reduced from a short-tem report (Special Report or Licensee Event Report) to an item which is to be included in the Annual Report. The information to be included in the Annual Report is similar to that previously required in the Licensee Event Report but has been changed to more clearly designate the results to be included from the specific activity analysis and to delete the infomation regarding fuel burnup by core region.

In our effort to eliminate unnecessary Technical Specification requirements, we have also detemined that the existing requirements to shut down a plant if coolant iodine activity limits are exceeded for 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in a 12-month period can be eliminated. The quality of nuclear fuel has been greatly improved over the past decade with the result that nomal coolant iodine activity (i.e. in the absence of iodine spiking ) is well below the limit.

Appropriate actions would be initiated long before accumulating 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> above the iodine activity limit. In addition, 10 CFR 50.72(b)(1)(ii) requires the NPC to be immediately notified of fuel cladding failures that exceed expected values or that are caused by unexpected factors. Therefore ,

thir. Technical Specification limit is no longer considered necessary on the basis that proper fuel management by licensees and existing reporting reovirements should preclude ever approachina the limit.

Licensees are expected to continue to monitor iodine activity in the primary coolant and take responsible actions to maintain it at a reasonably low level

fi.e., accumulated time with high iodine activity should not approach l 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br />).

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Enclosed are model Technical Specifications in Standard Technical Specification l (STS) format showing the revisions that may be used in a submittal of proposed Technical Specifications or proposed changes to er.isting Technical Specifications.

These changes will also be incorportted in the next revistor of the STS for all nuclear power reactor vendors. The changes are indicated by a line in the margin of the Action Statement for the Limiting Condition for Operation. A Technical Specification. amendment request should be submitted to the NRC for ,

each facility which currently has Technical Specification reporting requirements upon exceeding coolant iodine activity limits or which has a requirement to shut down after 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> with iodine above the limit. Your request should include appropriate changes to the bases section of your Technical Specifi-cations.

As a matter of information, when Technical Specificction changes or other preposed license amendments and approvals (i.e., proposed facility modifications requiring NRC approval) are required as a result of this or crother generic letter, they are subject to the fee provisions of 10 CFR 170 and a $150 applicction fee should accompany ycur request (see 10 CFR '70.12(c) and 170.21).

If you have any questions relating to this subject, please contact M. Virgilio of my staff on (301 a92-8947).

This request has been approved by OPB Clearance Number 3150-0011, which expires September 30, 1986.

Sincerely, jW-J f Hy(h . Thompson, Mrector, Df ion of Licensing

Enclosure:

Vcdel Technical Specificetions showing Pevisions to STS Reporting Requirements for Primary reclant Specific Activity" List of Generic letters

,i REACTOR COOLANT SYSTEM

• i 3/4.4.9 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.9 -The specific activity of the primary coolant shall be limited to:

. a. Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and

b. Less than or equal to 100/E microcuries/ gram .

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1, 2 and 3*.

a. With the specific activity of the primary coolant greater than 1.0 microcurfe/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 1ess than 500*F within 3.4-1, 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. be in at least HOT STANDBY with T**U

b. With the specific activity of the primary coolant greater than 100/E microcuries/ less than -

500*F within 6gram be in at least HOT STANDBY with T**9 hours.

MODES 1, 2, 3, 4 and 5:

a. With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries/

gram perform the sampling and analysis requirements of item 4 a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.9 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis program of 1 Table 4.4-4. l i

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"With T,yg greater than or equal to 500*F. ,

1 B&W-STS 3/* 4-23 1

ADMINISTRATIVE CONTROLS  ?. g ANNUAL] REPORTS (Add the following to this section) ,

~

The results of specific activity analysis ih which the primary coolant exceeded the limits of Specification 3.4.8 (W and CE plants),

3.4.9 (B&W plants) or 3.4.5 (GE plants). The following infonnation shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first samp,le in which the limit was exceeded; (2) Results of the last isotopic analysis for radiciodine performed prior to ex-ceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and r:ne other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

STS-ALL PLANTS

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/. p%g- UNITED STATES 1

! g NUCLEAR REGULATORY COMMISSION '

l g g m sn m oTom.o.c.acess C . November 8, 1985 s..... .

TO ALL LICENSEES WITH BABC0CK AND WILCOX OPERATING REACTORS Gentlemen:

SUBJECT:

RESOLUTION OF GENERIC ISSUE 69: HIGH PRESSURE 1

INJECTION /MAKE-UP N0ZZLE CRACKING IN BABC0CK AND WILC0X PLANTS (Generic Letter 85-20 )

On January 24, 1982, an unexplained loss of coolant was detected during nonnal plant operation at Crystal River Unit 3. After an orderly shutdown, an inspection of the reactor coolant and associated systems revealed that the high pressure injection / makeup (HPI/MU) check valve, valve-to-safe-end weld, safe-end, and thermal sleeve were cracked. Subsequently, inspections were perfomed at other B&W designed plants. Most of these inspections revealed similar types of cracking in the HPI system of the facilities which indicated that the cracking problem was a generic one. A Safe-End Task Force was formed by the B&W Owners' Group to compile pertinent facts and to make recommendations to solve the cracking problem. The Task Force completed its work in late 1982 and provided its findings and recommendations to the Owners' Group in the

" Babcock and Wilcox 177 Fuel Assembly Owners' Group Safe-End Task Force Report on Gent. ic Investigation of HPI/MU Nozzle Component Cracking."

Th staff reviewed the Task Force recommendations and agreed that the following actions be taken to resolve this issue: l 1

(1) Reroll the upstream end of the thermal sleeve when inspections l indicate that a gap exists or repair and/or replace damaged components; (2) Implement an augmented inservice inspection plan; and (3) Perform a detailed stress analysis of a nozzle with a mo *ied themal sleeve design to justify long term operation.

All participants in the Owners' Group Task Force have performed the recomended repairs to damaged components (Recomendation 1) and have voluntarily implemented a satisfactory augmented inservice inspection program (Recomendation 2).

Performance of a stress analysis (Recommendation 3) is' required to maintain original licensing comitments regarding the stress and fatigue usage allowables required by USAS B31.7 or the ASME code. Analyses for modified nozzles have been perfomed; analyses for nozzles not requiring modification should have been perfomed before licensing. Operating experience for some plants has irAicated that the expected fatigue analyses could be subs.tantially exceeded by the end of life. For example, an incrs. sed number J li2 ac tue s.. i. a..&.o .ould occur due to manual actuation after ' reactor trips to avoid losing pressurizer level. Therefore, the staff has detemined that it is necessary to ensure that  ;

valid stress analyses have been performed.

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i November 8, 1985

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Each licensee should verify that a valid stress analysis has been performed for HP!/MU nozzles in accordance with its licensing commitments to meet the Code requirements. Also, each licensee should verify that the cumulative fatigue usage for these nozzles is within the allowables based on a realistic projection of the thermal cycles expected for the life of the plant. The information including stress analysis results and expected number of thermal cycles should be maintained for future inspection.

This generic letter does not impose any new regulatory requirements or any reporting requirements. Therefore, no clearance from the Office of Management and Budget is required. If you have any questions, the staff contact is John Hannon who can be reached at (301) 492-8543.

Hygh L. Thompson rector Df ion of Licensin cc: List of Generic Letters m

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LIST OF RECENTLY ISSUED GENERIC LETTERS GENERIC LETTER NO. SUBJECT

. DATE 85-04 Operator Licensing Examinations 1/29/85 85-05 Inadvertent Boron Dilution Events 1/31/85 85-06 Quality Assurance Guidance for ATWS

- Equipment that is not Safety-Related 4/16/85 85-07 Implementation of Integrated Schedules 5/02/85 for Plant Modifications }

85-08 10 CFR 20.408 Termination Reports - Format 5/23/85 85-09 Technical Specifications for Generic Letter 83-28, item 4.3 5/23/85 85-10 Technical Specifications for Generic Letter 83-28, Items 4.3 and 4.4 5/23/85 85-11 Completion of Phase II of " Control of 6/28/85 Heavy Loads at Nuclear Power Plants" NUREG-0612 85-12 Implementation of TMI Action I'em II.K.3.5,

" Automatic Trip of Reactor Cooiant Pumps" 6/28/85 85-13 Transmittal of NUREG-1154 Regarding the Davis Besse Loss of Main and Auxiliary .

Feedwater Event 8/5/85 84-14 Commercial Storage at Power Reactor Sites of Low Level Radioactive Waste not Generated by the Utility 8/1/85 85-15 Information Relating to the Deadlines for Compliance with 10 CFR 50.49, " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants 8/6/85 85-16 High Boron Concentrations 8/23/83 85-17 Availability of Supplements 2 & 3 to NUREG-0933, "A Prioritization of Generic Safety Issues 8/23/85 85-18 Operator Licensing Examinations - -9/27/85 85-19 Reporting Requirements on Primary Coolant 9/27/85 Iodine Spikes .

j 85~20 Resolution of Generic Issue 69: High Pressure Injection /Make-up pozzle Cracking in Babcock and Wilcox Plants 11/8/85

.i pa a%

+ o,, UNITED STATES NUCLEAR REGULATORY COMMISSION

$ E WASHeNGTON. D. C. 20655

. e,

+.,...../ December 3, 1985 TO ALL LICENSEES OF OPERATING REACTORS APPLICANTS FOR OPERATING LICENSEES, AND HOLDERS OF CONSTRUCTION PERMITS.

Gentlemen:

SUBJECT:

POTENTIAL FOR LOSS OF POST-LOCA RECIRCULATION CAPABILITY DUE TO INSULATION DEBRIS BLOCKAGE (Generic Letter 85-22 )

This letter is to inform you about a generic safety concern regarding LOCA -

generated debris that could block PWR containment emergency sump screens or BWR RHR suction strainers, thus resulting in a loss of recirculation or containment spray pump net positive suction head (NPSH) margin'.

The potential exists for a primary coolant pipe break to damage thennal insulation on the piping as well as that on nearby components. Insulation debris could be transported to water sources used for long-tenn post-LOCA recirculation and containment sprays (i.e., PWR containment emergency sumps and BWR suction intakes in the suppression pools) and deposited on debris screens or suction strainers. This could reduce the NPSH margin below that required for recirculation pumps to maintain long-tenn cooling.

<# This concern has been addressed as part of the efforts undertaken to resolve USI A-43, " Containment Emergency Sump Performance." The staff's technical findings contain the following main points.

Plant insulation surveys, development of methods for estimating debris generation and transport, debris transport experiments, and information provided as public comments on the findings have shown that debris blockage effects are dependent on the types and quantities of insulation employed, the primary system layout within containment, post-LOCA recirculation patterns and velocities, and the post-LOCA recirculation flow rates. It was concluded that a single generic solution is not possible, but rather that debris blockage effects are governea by plant specific design features and post-loca recirculation flow requirement.

The current 50% screen blockage assumption identified in Regulatory Guide (RG) 1.82, " Sumps for Emergency Core Cooling and Containment Spray Systems," should be replaced with a more comprehensive requirement to assess debris effects on a plant-specific basis. The 50% screen blockage assumption does not require a plant-specific evaluation of the debris-blockage potential and usually will result in a non-conservative analysis for screen blockage effects.

The staff has revised Regulatory Guide (RG) 1.82, Revision 0, " Sumps for Emergency Core Cooling and ContaL. ment Spray Systems" and the Standard Review Plan Section 6.2.2, " Containment Heat Removal Systems" based on the "51:27 5 2

m O 2 .

above technical findings. However, the staff's regulatory analysis (NUREG-0869, Revision 1, "USI A-43 Regulatory Analysis") evaluated (1) containment designs and their survivability should loss of. recirculation occur, (2) alternate means to remove decay heat (3) release consequences (which were based.on pipe break probabilities which did not incorporate insights gained from recent pipe fracture mechanics analyses), and (4) cost estimates for backfits considered (i.e., reinsulating). This regulatory analysis did not support a generic backfit action and resulted in the decision that this revised regulatory guidance will not be applied to any plant now licensed to operate or that is under construction. The revised guidance will be used on Construction Permit Applications, Preliminary Design Approval (PDA) applications, and applications for licenses to manufacture that are docketed af ter six (6) months following issuance of RG 1.82, Revision 1, and Final Design Approval (FDA) applications, for' standardized designs which are intended for referencing in future Construction Pennit Applications, that have not received approval at six (6) months following issuance of the RG 1.82, Revision 1.

Although the staff has concluded that no new requirements need be imposed on licensees and construction pennit holders as a result of our concluding analyses dealing with the resolution of USI A-43, we do reconinend that RG 1.82, Revision 1 be used as guidance for the conduct of 10 CFR 50.59 reviews dealing with the changeout and/or modification of thermal insulation installed on primary coolant system piping and components. RG 1.82, Revision 1 provides guidance for estimating potential debris blockage effects. If, as a result of NRC staff review of licensee actions associated with the changeout or modification of thermal insulation, the staff decides that Standard Review Plan Section b.2.2, Revision 4 and/or RG 1.82, Revision 1 should be (or should have been) applied to the rework by the licensee, and the staff seeks to impose these criteria, then the NRC will treat such an action as a plant-specific backfit pursuant to 10 CFR 50.109. It is expected that those plants with small debris screen areas (less than 100 ft2), high ECCS recirculation pumping requirements (greater than 8000 gpm), and small NPSH margins (less than I to 2 ft of water) would benefit the most from this type of assessment in the event of a future insulation change. RG 1.82, Revision 0 with its 50% blockage criteria does not adequately address this issue and is inconsistent with the technical findings developed for the resolution of USI A-43.

This information letter along with enclosed copies of NUREG-0897, Revision 1 RG 1.82, Revision 1 and SRP Section 6.2.2, Revision 4 should be directed to the appropriate groups within your w janization who are responsible for conducting 10 CFR 50.59 reviews.

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l No written response or specific action is required by this letter.

Therefore, no clearance from the Office of Management and Budget is required. If you have any questions on this matter, please contact your project manager..

Hu fi L. Thompso Jr 31 rector Dv ion of Licensip,

Enclosure:

NUREG-0897, Revision 1 RG 1.82, Revision 1 SRP Section 6.2.2, Revision 4 D

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