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SAFETY EVALUATION REPORT 3
GENERIC LETTER 83-28, ITEM 2.1 (PART 1) '
EQUIPMENT CLASSIFICATI0h (RTS COMPONEhT5)
PILGRIM NUCLEAR POWER STATION~
DOCKET N05. 50-293 INTR 000CTION AND
SUMMARY
On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident was terminated manually by the operator about 30 seconds after the initiation of the automatic trip signal.
The failure of the circuit breakers was determined to be related to the sticking _
of the undervoltage trip attachment. Prior to this incident, on February 22, '
1983, at Unit 1 of the Salem Nuclear Power Plant, an automatic trip signal was generated based on steam generator low-low level during plant start-up. In this case, the reactor was tripped manually by the operator almost coin-cidentally with the automatic trip. l Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO), directed the staff to investigate and report on the generic ;
i implications of these occurrences at Unit 1 of the Salem Nuclear Power Plant.
The results of the staff's inquiry into the generic implications of the Salem unit incidents are reported in NUREG-1000, " Generic Implications of the ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28 dated July 8,19831 )
all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two ATWS events.
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This report is an evaluation of the response submitted by Boston Edison Company, the licensee for the Pilgrim Nuclear Power Statfor, for Item 2.1 (Part 1) of Generic Letter 83-28. The actual documents reviewed as part of this evaluation are listed in the references at the end of the report.
Item 2.1 (Part 1) requires the 1icensee to confinn that all Reactor Trip System components are identified, classified and treated as safety-related as indicated in the following statement:
Licensees and applicants shall confirm that all components whose functioning is required to trip the reactor are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, in-cluding maintenance, work orders, and parts replacement.
EVALUATION The 1fcensee for the Pilgrim Nuclear Power Station responded to the requirements '
of Item 2.1 (Part 1) with submittals dated November 7,19832 and June 28, 19853 .
The licensee stated in these submittals that all components th.at are required to l perform the reactor trip function were reviewed to verify that these components are classified as safety-related equipment in the plant "Q-list." The licensee further confinned that documents used to control activities associated with this equipment are identified as "Q" which designates the use of safety-related procedures.
l CDNCLUSION Eased on our review of these responses, we find the licensee's statements confirm that a program exists for identifying, classifying and treating components that are required for performance of the reactor trip function as safety related. This program meets the requirements of Item 2.1 (Part 1) c' the Generic Letter 83-28, and is therefore acceptable.
REFERENCES
- 1. NRC Letter, D. G. Eisenhut to all Licensees of Operating Reactors, ,
Applicants for Operating License, and Holders of Construction Permits,
" Required Actions Based on Generic Implications of Salem ATWS Events (Generic Letter 83-28)," July 8,1983.
- 2. Letter, W. D. Harrington, Boston Edison Co., to D. B. Vassallo, NRC, November 7, 1983.
- 3. Letter, W. D. Harrington, Boston Edison Co., to D. B. Yassallo, NRC, June 28, 1985.
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- . ' , ", , ' , * " , , * " " " ' " ~ ' December 30, ;H3 8ECo 83- 308 Mr. Domenic B. Vassallo, Chief i Operr. ting Reactors Branch #2 Division of Licensing Office of Nuclear Reactor Regulation <
U.S. Nuclear Regulatory Commission Washington, D. C. 20555 License No. OPR-35 Docket No. 50-293 Incorporation of Vendor Manual
, Validation into a Nuc~ea- Operation Procedure
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Dear Sir:
In responc,ng to Generic. Letter 83-28 :n wovember 7,1983, Boston Edison stated in Section 3.1.j'that the vende- manual validation process, which had been undertaken as part of our Procedure ';)pdate Program (PUP), was Wing incorporated into a Nuclear Organization Crocedure (NOP). The NOP was in review at the time of the response and emplacement was expected by January 1, 1984. ,
We regret that preparation for the curren refueling has made this date unattainable. As discussed between Mr. V. Rooney (NRC) and Mr. P. M. Kahler (Boston Edison) in a telephone conversation on December 27, 1984, Boston Edison now expects completion of this tenn on February 15, 1984.
W<< appreciate your attention to this issue. Should you require further information concerning it, please conta:t us.
Very truly yours, W
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atFtetNCES: 1) W. SICo letter me.83-131 D. Marrington #ated 5/18/8).
to 0. 8. vassalto
- 2) SECo Letter No.83-121 dated 5/18/83 W. O, Marrington to D. 8. vassallo
- 3) Meeting between BECo and the meC on May 22, 1984
Dear 51r:
On may 22, 1984, loston (dison Coacany met with members of your staf f (Ref e*ence 3) to discuss 80ston (dison Co's proposed method of resolution for enn of the deficiencies contained in the Technical Evaluation Report 81sc us sic'et (Ita) written ipy f ranklin Research Center under contract to the met.
also took place at the meeting regarding 1) loston (dise.'s approach in resp.n45ng to 10CFR50.ag section (t)(1), (b)(2) & (b)(3), 2) the Pilgrin maintenance and Surveillance Program to address equipment qualification, 3) Nus;tC64 Osteautroer Besten Edison's position on tu Info. Notices $2-52 & 83 72. and 4) .sa ndr w =
Justification f or continued Operation.
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The purpose of this letter is to provide you with 1) documentation of the discussions held at the May 22 aceting, 2) final resolution of l l deficiencies for all TER equipment itees including the updated resolution of r~
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itees d ich were identified as
- Evaluation in Progress' et the ltime , -'
, of Ma the ita. (nclosure 1 to this letter contains the suesnary of the proposedfor those ' eoip.ent ites' ',
J M I 'L res.t. tion for each of the deficiencies in the it8.
for dich the documentation for environmental aus11fication is not yet l [ * '
comeleted, a justification f or continued operation (JCo) is provided as enc *osure 2 to this letter. E- .
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-3ECo Letter No.84-099 L,1y 9, 1984 m svos coisos coway At the May 22nd, 1984 meeting, a number of specific issues related to i TER resolution were discussed and their conclusions have been incor into the final resolution.
as 'Out of Scope' to 50.49 requirements will have traceable documentati support such a conclusion.
letter.
However, it is available for your audit. Other issues such as the :
qualification concerns with Rockbestos Cable and Terminal 81ocks final in
. instrumentation circuits in the drywell are addressed as part of. the resolution f or these items in Enclosure 1.
Generic deficiencies listed in Section 5 of the TER deal with 1)
' instrumentation accuracy requirements in instrument qualification evaluation' and 2) "Nhy Pilgrim MSLB curve ends at 2000: seconds, while the c continuing to rise.' The results is addressed as part of the instrument qualification evaluation.
of this evaluation are documented as a line item on Pilgria equipment.
qualification evaluation sheets (EQES=SCEW) which Pressure - Temperature (p-t) Profiles f or both inside and outs containment.
a postulated high energy line break and are used as the basis for 8Eco'sT equipment qualification evaluation. seconds as shown on MSLB c drywell design temperature limit ofMSLB THE 281*F.
curve evaluations (previously inside dryw of LOCA and plotted in M632 SH.16 apply.
submitted) should be used f or information only.
As agreed in the meeting items to be environmentally qualified that have been added to the ' Master List of Electrical Equipment
- and not f acto in the TER resolution process, will be submitted with resolutions and applicable .lCO's in our next submittal on August 3,1984.
The method of identification of electrical equipment within the scope of 10CFR50.49 paragraph (b)(1), (b)(2), & (b)(3) is describe to this letter.
will be performed.
The concerns raised in IE Notices 82-52 and 83-72 and disc May process.
22nd meeting have been evaluated and incorporate E.Q. Notice No.1 deals with limitorque motor '
is applicable to PNPS. operators which were tested Enclosure 1 totothis a much more s Under the motor operators at PNPS will ever be subjected. letter item 11, only 1.E. Notice 82-03 is applicable at PNPS.In I.E. Notice 83-12f our current evaluation.Even though equipment addressed in E.Q. Notices 21 & 22 does 1 apply to PNPS.
exist at PNPS, tr.e f ailure parameters E.Q. Noticedescribed in these notices 24 is being addressed by are much t
v conservative for PNPS conditions.recomended inspections and replac parts.
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4 na tener No.84-099 July 5. ".984 .
Coote= coisas cowey On Maintenance and Surveillance Practices, your staff was informed at the May 22nd meeting by BEco that the qualification of equipment will be assured from the time its qualification is established. New equijpent to be added in the plant will be evaluated for E.Q. requirements prior te. Its procurement and hence assuring its qualification. Trending of the-equipment for possib'le degradation of operational characteristics is currently addressed yonderathterfaces p by plant Failure & Malhnction Report Frocess, -
by ths.exhting KCo prograss andAcast.ralized approad Prograe (VETIP sahene yldeterface Technical 4Hf4tGationand the NS$$ Vender; and the'*6hed)et**ilheM q As discussed at the May 22nd meeting it is requested that supplemental SER's be issued to indicate Boston Edison's Equipment Qualification Program as cescribed in this letter meet the require.wnts of 10CFR50.49 and that the deficiencies noted in the SER date April 13, 1983 are considered resolved.
We would be pleased to answer any questions you may have regarding the enclosed information.
Very truly yours.
W. D. Harrington WDH/TAV/mre Enclosure 1: TER Resolution Matrix Enclosure 2: Justification For Continued Operation Enc usure 3: Methodology to identify equipment within the scope of 10CFR50.59 (b)(1), (b)(2) &
(b)(3)
Enclosure 4: Pressure - Temperature Profiles M632 SH) - 16
.a7g "%g'1, UNITED STATES
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July 13,1984 Docket No. 50-293 Mr. William D. Harrington Senior Vice President, Nuclear Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199
Dear Mr. Harrington:
SUBJECT:
LONG TERM PROGRAM The Commission has issued the enclosed __ Amendment Mn 75 to Facility -
Operating License No. DPR-35 for the Pilgrim Nuclear Power Station. This amendment (Enclosure 1) incorporates a license condition in response to your application ~ dated July 5,1983, as revised by your submittal dated May 7, 1984 The license condition requires the Boston Edison Company (BECo) to follow its " Plan for the Long Tem Program - Pilgrim Nuclear Power Station" (the Plan) and the terms therein for revising the schedules for specific modifications.
Our evaluation of the Plan is provided in Enclosure 2 and a copy of the Plan (as revised) is enclosed as Attachment 1 of Enclosure 2. Based on our review, we conclude that your revised Plan is acceptable.
The Consnission has recently issued Orders to the licensees of operating reactors, confirming their schedules for compliance with the emergency response capability requirements specified in NUREG-0737, Supplement 1. A similar Order is not being sent to you because your schedule comitments for complying with those requirements are included in the Plan. The procedures in the Plan, which you are required to follow in order to modify its scheduled completion dates, previde assurance that significant actions will be completed in a timely fashion.
Boston Edison Company is to be coninended for developing this integrated program for scheduling safety modifications at the Pilgrim Nuclear Power Station. We anticipate that the resulting improvement in control and management of available resources will facilitate more systematic and timely implementation of such modifications. After you have had soma g L L..L be' E/r
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Mr. William D. Harrington '
experience with this Plan, we would appreciate receiving your comments, including any suggestions for improvements.
Sincerely,.
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- \,, 11 l Dar'reH 'G. Iisenhut,. Director.
Division of Licensing Office of Nuclear Reactor Regulation.
Enclosures:
- 1. Amendment No. 75 to License No. DPR-35
- 2. Program Evaluation cc w/ enclosures:
See next page l
l
Mr. William D. Harrington ' ' ' '
Boston Edison Company Pilgrim Nuclear Power Station cc:
Mr. Charles J. Mathis, Station Mgr. Thomas A. Murley Boston Edison Company Regional Administrator RFD #1, Rocky Hill Road Region I Office Plymouth, Massachusetts 02360 U. S. Nuclear Regulatorf Commission 631 Park Avenue Resident Inspector's Office King of Prussia, Pennsylvania 19406 U. S. Nuclear Regulatory Commission Post Office Box 867 Mr. A. Victor Morisi Plymouth, Massachusetts 02360 Boston Edison Company 25 Braintree Hill Park Mr. David F. Tarantino Rockdale Street Chairman, Board of Selectman Braintree, Massachusetts 0218t 11 Lincoln Street Plymouth, Massachusetts 02360 Water Quality and Environmental Commissioner Department of Environmental Quality Engineering 100 Cambridge Street Boston, Massachusetts 02202 Office of the Attorney General '
1 Ashburton Place 19th Floor Boston, Massachusetts 02108 U. S. Environmental Pratection Agency Region I Office Regional Radiation Representative !
JFK Federal Building Boston, Massachusetts 02203 Mr. Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 600 Washington Street, Room 770 Boston, Massachusetts 02111
_ . . _ _ , _ _ . . _ . . - _ _ - - - - - - - U
, ENCLOSURE 1
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BOSTON EDISON COMPANY DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 75 License No. DPR-35
- 1. Thp Nuclear Regulatory Commission (the Conrission) has found that:
A. The application for amendment by Boston Edison Company (the licensee) dated July 5,1983 (as revised icy 7,1984) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Consission's rules and regulations set forth in 10 CFR Chapter I:
B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Ccenission; C. There is reasonable assurance (1) that the activities authorized l by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Connission's regulations; i
D. The issuance of this amendment will not be inimical to the cormon defense and security or to the health and safety of the public; and i E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Connission's regulations and all applicable requirements have been satisfied.
- 2. Accordingly, the license is amended by adding a new paragraph 3.H to read as follows:
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3.H Long Tenn Program
- 1. The " Plan for the Long Term Program for Pilgrim Mclear Power Station" (the Plan) submitted on May 7,19Ed, is approved.
a) .The Plan shall be followed by the licensee frza and after the effective date of this amendment.
b) Changes to dates for completion of items iden:ified in Schedule B of the Plan do not require a license amendment. Dates specified in Schedule A sha 1 be changed cnly in accordance with applicable NR" procedure.
- 3. This license amendment is effective as of the date of its 'ssuance.
FOR THE NUCLEAR REGULATORY CMMISSION
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Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing Date of Issuance: July 13,1984 l
ENCLOSURE 2
- 'f UCQ UNITED STATES
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INTEGRATED SCHEDULING PROGRAM _
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 75 TO FACILITY LICENSE NO. DPR-35 BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293 1.0 Introduction By letter dated April 14, 1983, Boston Edison Company (BECo, licensee) submitted its Long Term Program (the Program) for modifications of the Pilgrim Nuclear Power Station (PNPS) for NRC review and approval. Sub-sequently, the licensee requested in a July 5,1983 license amendment application that a condition be added to the PNPS operating license requiring BEco to follow the " Plan for the Long Term Program for Pilgrim Nuclear Power Station." Following discussions with the NRC staff, BECo revised its application by letter dated May 7,1984 The Program was developed by Boston Edison to coordinate and schedule major necessary work at PNPS, whether mandated by NRC or identified by BECo. The Program integrates all presently planned work at FNPS over a nominal three year period to enable effective scheduling and coordination of individual tasks.
The " Plan for the Long Term Program - Pilgrim Nuclear Power Station" (Attachment 1) is the implementation vehicle for the licensee's Long Term Program. The Plan describes how tne Program functions, the mecha-nisms for changing and updating it, and the interaction of the NRC and i BEco under the provisions of the Progran and its associated schedules.
The staff issued a notice of the proposed license amendment in the Federal Register of September 21, 1983 In that notica, the staff proposed a determination of no significant hazards consideration.
The licensee's subsequent submittal of May 7,1984 revised the Plan to update the schedules semiannually instead of quarterly. The submittal also incorporated editorial changes which recognize that a licensee-proposed change in Schedule E would be extended if NRC recuests discussion of the proposed change. Schedules A and B were also updated to reflect the expected accomplishment of additional modifications during the extended refueling outage d6 rather than during a mid-cycle #7 outage. These are administrative changes considered by the staff to be within the scope of the initial notice. j l
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2.0 Program Description The program developed by Boston Edison is based on a ccanputer generated listing of over 500 items of prioritized work. Certain of these items, were organized into Schedules A and B using critical path methodology and considering site manpower and engineering support requirements for a three year period.
> BECo's program integrates the engineering, procurement and installation of planned NRC-reguired modifications with Boston Edison's own require-ments for plant modifications, maintenance, refueling, and operations.
In developing its Program, the licensee prioritized work items, considered the impact of work-area manpower densities during modifi-cations and refueling activities, and identified the desirability of a mid-Cycle modification outage to complete certain work items.
Boston Edison stated that itnplementation of its program would be facilitated by working towards completion schedules which are cycle (or outage) dependent versus fixed calendar dates.
Although not specifically accounting for future new requirements (other than those currently envisaged in its proposed program),
BEco's Program is structured so that additional required plant modifications can be integrated into the overall program to identify the impact of such new requirements on the overall scFedule.
The Plan submitted by the licensee identifies two categories of modifications. Schedule A identifies schedules for modifications established by existing Rule or Order. Schedule B identifies schedules for completion of:
- 1) Regulatory items (of either a generic or plant specific nature) identified by NRC which would result in a) plant modifications, b) procedure revisions, or c) changes to facility staffing requirements and which have an implementation date comimitted to by Boston Edison;
- 2) Items perceived by BECo as prospective NRC requirements; and,
- 3) All major Pilgrim tasks resulting from mandates of agencies other than NRC and BEco initiated system upgrades for availability improvement.
3.0 Evaluation 3.1 Implementa tion The licensee's July 5,1983 submittal (as revised) incorporates an application for acerdment to incorporate a license condition reouiring that Boston Edison follow the Plan and permitting the licensee to make
changes to the Plan and its schedules for certain categories of items in accordance with the provisions of the Plan. We have reviewed the licensee's Plan and have determined that:
- 1) Changes to schedules for completion of modifications imposed by Rule or Order (Schedule ' A' completion dates) will continue to be sought through the exemption or Order-date extension process (For example, Boston Edison's existing request for exemption from certain schedular requirements of 10 CFR 50.48 regarding fire protection.).
- 2) Schedules for completion of other modifications (Schedule 'B' completion dates) are identified and provisions are made in the Plan to require BECo to provide the NRC with prior written notification of changes to Schedule B completion dates to enable further explanation or discussion of such changes.
- 3) Provisions are made in the Plan for incorporating new or anticipated regulatory items into Schedules A and B as these requirements are identified by NRC and/or formalized by Rule or Order.
The licensee identified each planned NRC-required modification as an individual line item in its Schedules. Semi-annual reports of utility progress towards implementation of NRC-identified modifications are proposed by the licensee.
The licensee's proposal to incorporate a condition into the PNPS operating license which requires BECo to follow the Plan provides an appropriate mechanism to assure that NRC is informed as to whether required safety modifications are performed in a timely manner. At the same time, the Plan provides a suitable mechanism for changes to l completion dates (due to unforseen circumstances) for modifications !
not imposed by Rule or Order and for keeping the NRC informed of such changes for its consideration. Thus, the degree of flexibility needed :
to assure effective program implementation is provided while at the same time assuring that NRC's responsibilities are not compromised.
The Plan and the proposed license condition submitted by the licensee are functionally identical to those approved by the staff in Amendment i No. 91 to the Duane Arnold Energy Center (DAEC) operating license. A l copy of this amendment was transmit';ed to all power reactor licensees !
by Generic Letter 83-20 on May 9,1983. This letter identified the i
approach addressed by Amendment No. 91 as one which is acceptable to i the NRC. Thus, we find that 1) the Plan proposed by Boston Edison is equivalent to a previously approved Plan for implementation of an '
integrated scheduling Program, and 2) the license condition proposed i by BECo is equivalent to the previcusly approved license condition for the DAEC on this subject.
l l
3.2 Proposed Schedules Attachments 2 and 3 provide Boston Edison's proptsed schedules for completion of all presently known BECo-planned ard NRC-required modifications over a three year period.
For requirements imposed by Rule or Order (other than the schedule for completion of hydrostatic testing of Class II and III piping) the utility proposes completion by required dates. The utility has requested an extension of time to perform hydrostatic testing of certain Class 2 and Class 3 piping systems. Afte- staff action on this matter, the schedule for completion of these tests will be revised as necessary.
With respect to NUREG-0737 Supplement 1 items, the utility revised its initially proposed schedules. In negotiating the dates for completion of Supplement 1 items, the licensee committed to :roviding the neces-sary reports, plans and analyses and to provide its final schedule for full implementation on a schedule which meets our guidelines.
Consequently, we find it acceptable.
Certain schedules for completion of modifications to the Pilgrim facility are keyed to completion of required NRC staff reviews that would result in subsequent approvals. For example, the schedule for certain modifications required by 10 CFR 50.48 is determined by the date of completion of the staff review. The licersee has proposed completion cf this issue during refueling outage %. 7, which is con-sistent with the provisions of the rule for determining required completion schedules.
Boston Edison has proposed completion of other NRC roodifications not required by Rule or Order on a schedule consistent with its previous commitments. Significant regulatory items in this category scheduled for completion during the December 1983 refueling outage include purge and vent valve modifications and RPS power supply modifications. As agreed to by tre licensee, the implementation schedule for the " Scram Discharge Volume" has been moved from Schedule B o Schedule A since this item was the subject of a Comission Order dated June 24, 1983.
With this change, we find the licensee's schedule acceptable.
Based upon our review of the information contained in BECo's submittal, we find the dates proposed by the licersee for completion of modifications acceptable.
4.0 Sumary Based on the considerations aodressed herein, we f'nd that:
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- 1) The proposal by Boston Edison that its Plan be implemented :y a license condition requiring the utility to follow the Plan is acceptable.
- 2) The licensee's proposal that changes to implementation dates imposed by existing Rule or Order will continue to be sough:
through the exemption or order date extension process is ac:eptable.
- 3) Schedules for new requirements should be established for the F%PS on a plant specific basis.
- 4) Based upon our review, the completion dates proposed by the licensee in its submittal appear reasonable.
- 5) The license condition and the Man submitted by Boston Edisen are equivalent to that already apprcved by Amendment No. 91 to me Duane Arnold Energy Center.
5.0 Environmental Consideration This amendment involves changes in the installation or use of facility components located within the restricted area. The staff has determined that the amendment involves no significant increase in the amounts of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupation radiation exposure. The Comission has previously issued a proposed findirg that this amendment involves no significant hazards consideration and there has been no public coment on such finding. Accordingly, t,is amendment meets the eligibility criteria for categorical exclusicn s.et forth in 10 CFR Sec 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be i prepared in connection with the issuance of this amendment.
6.0 Conclusion We have concluded, based on the considerations discussed Isove, taat:
(1) there is reasonable assurance that the healtf and saf ety of t9e public will not be endangered by operation in the propossd manner, and (2) such activities will be conducted in compliance wit > the Comdssion's regulations and the issuance of this amendment will not be inimic:1 to the common defense and security or to the health and iifety of the public.
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l Principal Contributor: Kenneth T. Eccleston and Paul H. Leech Attachments:
- 1. Long Term Program
- 2. Schedule A <
- 3. Schedule B Dated: July 13,1984
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i t_--_________ . _ _ _ _ ___
- ATTACHMENT 1 to-ENCLOSURE 2' Plan for the Long Term P'rogram - Pilgrim Nuclear Power Statis I. Introduction Boston Edison Company (BECo) has developed a comprehensive program which will enable the Company to effectively manage implementation of certain modifications which have been required, or proposed by, the NRC, as well as other measures to enhance plant safety and reliability which have . been identified by the Company. A description of the program, identified as "Long Term Program - Pilgrim Nuclear Power Station (PNPS)," was submitted to the NRC on April 14, 1983 by BEco Letter No. 83-91.
This program was developed to coordinate and schedule major necessary work at PNPS, whether mandated by NRC or identified by BECo and others. The program objectives are to (1) conform to regulatory requirements; (2) provide sufficient lead times for modifications; (3) minimize changes for operators; manage (4) assure financial training requirements are fulfilled; (5) effectively and human changes to developed schedules.
resources; and (6) specify the framework for This Long Term Program (the Program) reflects that fiscal and manpower resources are finite and that a limit on the onsite manpower is necessary, i
The Program integrates all presently planned work at PNPS over a nominal three year period to ensure that individual tasks are effectively scheduled and coordinated. It provides a means for new requirements to be accommodated l
taking into account schedule and resource constraints.
The purpose of this docurent is to describe the plan used to isolement the i Program.
i It des:ribes how the Program functions, mechanisms for changing and updating it, and the interactions of NRC and licensee staffs under the Program, and its associated schedules.
II. Sumary of Long Term Program Development The Program is based .on a computer generated listing of over 500 items of prioritized work. The listing takes into account projections for bud and site manpower and engineering support requirements for three years,gets on an item-by-item basis covering major plant modification activities. It represents the PNPS work list and commitment list which is regularly modi-fied and updated to meet changing conditions, including new NRC regulatory requirements. The final product of this Program is the development of schedules as discussed below.
III. Scheduling upon completion of the complete work listing, Boston Edison deternined that detailed and integrated schedules were required for the major work items.
Upon completing the comprehensive listing of major work items, the tasks were organized into Schedules A and B using critical path methodology (CPM) for selected work items. CPM schedules identify critical paths in the work
effort for each task, which,in turn, enables prompt adaptation of schedules to meet contingencies such as strikes, delays in procurement or installation or modification of fuel cycle schedules. Both schedules are briefly des-cribed below:
Schedule A -
All items whkh have implementation dates mandated by
. NRC rules, orders, or license conditions. l Sche 62 B - Regulatory items (of either a generic or plant speci-fic nature) identified by NRC which have implementa-tion dates committed to by Boston Edison and which would result in either a) plant modifications, b) procedure revisions, or c) changes in facility staff-ing requirements; or items perceived by Boston Edison as prospective NRC requirements; or major PNPS tasks resulting from mandates of agencies other than NRC and BEco-initiated system upgrades for availability improvement.
Schedule A dates may be modified only with tne prior approval of NRC, in accordance with existing NRC procedures. Changes in Schedule B dates require written notification to NRC as described in Section V.
Schedules A and B, taken together, provide a basis for assessing the overall effects of changes to schedules and a departure point for discussion between NRC and the licensee regarding such changes, as discussed below.
IV. Schedule Modifications .
An important aspect of Borton Edison's planning effort is the recognition that the attached schedules will need to be modified at times to reflect changes in regulatory requirements, to accommodate tnose activities that Boston Edison finds necessary to improve plant efficiency and reliability, and to take into account delays resulting from events beyond BECo's control.
It is important that the procedure used by Boston Edison for changing the schedules be documented.*/ In addition, the NRC must play a role in the 1 oversight of the scheduling process (and must, in fact, judge the accept-ability of proposed date changes in Schedule A). Accordingly, it is impor-tant that the NRC's role, and the Interaction between the NRC and BECo be clearly defined, as discussed below.
V. Boston' Edison Company Res' possibilities The integrated schedule requires that BECo monitor the progress of the work undertaken, manage its activities to maintain the schedule, and act promptly to take necessary actions when a schedule change is needed.
- /
~~ Schedules A and B will contain sufficient detail to identify those items with completion dates keyed to fuel cycle cutages. In such cases, a change in outage period shall not be considered a schecule change.
I I
)
A. Periodic Updating i
BECo will update Schedules A and B semiannually and su:mit the revised schedules to NRC, beginning six months following NRC c:ncurrence in the Plan. In addition to updating the schedules, BEco will:
Sumarize progress in implementing NRC requirements concerning plant modifications.
Identify changes since the last report.
- Sumarize the reasons for schedule changes associated with regulatory requirements.
B. Changes to Schedules Changes to the schedules may arise from a variety of reasons, such as new work activities; modifications in the scope of scheduled work; problems in delivery, procurement, etc.; changes in NPC rules and regulations; or other NRC or BECo actions.
Where it is necessary to add a new work item or to charge the schedule for an item, the following general guidance will be utilized to the extent appropriate:
Assess the priority of the work item and its safety significance.
Schedule the new or changed item to avoid reschedul'ng other items, if it can be reasonably achieved.
Alter Schedule B items before Schedule A items.
- Select a schedule for the new or changed item which will help in maintaining an optimum integrated program of work.
As noted above, no changes will be made in Schedule A without prior NRC approval. Should a change become necessary, it wi'l only be proposed after Boston Edison has determined that rescheduling of non-NRC required work items either will not significantly assist in maintaining Schedule A without change; or that the safety, cost or schedule penalties from rescheduling non-NRC required wrt significantly outweigh the change in a Schedule A comp'etion date.
Boston Edison will inform the NRC Project Manager when serious consideration is given to requesting a change in Sched.le A. When BEco.detemines that a change in Schedule A is necessary, it will submit a written request for NRC approval in accordance w-ith applicable procedures.
Boston Edison will notify NRC in writing at least 30 days before adopting a planned delay for an item in Schedule B. Such notification :
will also include the reasons for the delay and describe any compensatory actions indicated.
The revised date proposed by BECo will go into effect unless NRr, in writing, requests further explanation or discussion concerning such change. IF NRC makes such a request, it will be made within 15 days of receipt of BECo's written notification. In.this event, discussions will be initiated to promptly develop a schedule date which is mutually acceptable to Boston Edisen and the NRC Project Manager while considering overall program impact. The written notification by NRC will serve to extend the schedule date for the period of time required for such discussions. If a new date is established in these discussions such date will supersede the date set forth in Schedule B.
The new date will be incorporated in a revised Schedule B in the next schedule update submitted to NRC. If a new date cannot be established in these discussions, BECo changes in scheduled dates will be effective unless subsequently modified by NRC Order.
In the event of unplanned delays or circumstances beyond BEco control, BECo shall promptly notify the NRC Project Manager of the new date and incorporate it in a revised Schedule B in the next schedule update submitted to NRC.
VI. NRC Review As pointed out in Section V.B above, changes to the schedules are inevitable. Action required by NRC is discussed below: '
A. Boston Edison Originated Changes
- 1. Upon receipt from BECo of a reauest for modification of Schedule A, NRC will act promptly (consistent with resource availability and prior-ity of other work) to consider and decide on the request in accordance '
with applicable procedures.
- 2. If the request for a modification of Schedule A is denied, NRC shall promptly inform Boston Edison and provide the reasons for denial.
- 3. NRC consideration of BECo changes in non-Schedule A items is covered by V.B.
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B. NRC Originated Changes (Schedule A)
It is recognized that formal NRC regulatory actions may: (1) impose a new regulatory requirement with a fixed date or (2) establish a firm date for a previously identified regulatory requireneet. In taking any such action the NRC, to the extent consistent with its overall regula- i tory responsibilities and, unless public health, safety, or interest !
require otherwise, will take into account the impact of such action on 1 BECo's ability to complete effectively the items on Sc.hedules A and B, and, in consultation with BEco, will try to minimize such impact.
Although any formal regulatory action taken by the MRC will be effective in accordance with its terms without inclusion in Schedule A, the NRC and BEco recognize the desirability of incorporating such action into Schedule A, particularly in order to incorporate at the same time an other appropriate changes in the total integrated schedule program. y Accordingly, ence such formal regulatory action is taken (or earlier, if practicable), the NRC will provide BECo a reasonable opportunity to propose overall changes in the total integrated schedule program which would most effectively acconinodate such requirements. Any resulting changes in items in Schedule A will be approved by kRC in accordance with established procedures, and will thereupon be reflected in a revised schedule A submitted by BECo. BECo will infons the NRC of any resulting changes in Schedule B in accordance with Section V. above.
C. New NRC Issues (Schedule B)
The NRC may, from time to time, identify new regulatory issues which may result in a) plant modifications, b) procedure twision or development, or c) changes in facility staffing requirements. For issues as to which NRC requests scheduling information, these issues may be included in Schedule B in accordance with the date commitment developed in discussions between BECo and the NRC staff. As for the case of NRC-originated changes to Schedule A items, the NRC will provide BECo a reasonable opportunity to propose overall changes in the total integrated schedule program which would mest effectively accommodate such issues. Any resulting changes in integrated program schedules will thereupon be reflected in a revised Schedule B submitted by BEco.
1 Vli. Modifications to the Plan The licensee and the NRC recognize that the Plan itself may require future modifications. Accordingly, BECo will draft proposed modifications and submit a license amendment application for approval of the ;roposed changes. The changes will be made effective upon amendmert issuance by the NRC.
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UNITED STATES I a NUCLEAR REGULATORY COMMISSION /
$ ,E . WASHINGTON, D C,20555 g ,
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April 23,1985 l
Docket No. 50-293 Mr. William D. Harrington Senior Vice President, Nuclear Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199
Dear Mr. Harrington:
SUBJECT:
RE0 VEST FOR ADDITIONAL INFORMATION FOLLOWING PRELIMINARY STAFF REVIEW OF LICENSEE RESPONSES TO GENERIC LETTER 83-28 'N' I Re: Pilgrim Nuclear Power Station \
The staff has completed a preliminary review to assess the completeness and adequacy of licensee responses to Generic Letter 83-28 Items 2.1, 2.2, 3.1.3, 3.2.3, 4.4 and 4.5. For the Pilgrim Station, your responses were found to be incomplete for Items 2.1, 2.2.2, 3.1.3, 3.2.3 and 4.5.3. Brief descriptions of the deficiencies are given as guidelines for corrective action in the enclosed Request for Additional Information. Efforts by Owners Groups, INPO and NSSS vendors have been or are being made to produce generic responses that may be useful in meeting the requirements of Generic Letter 83-28 Items 2.1, 2.2.2 and 4.5.3. You may wish to contact your Owners Group or INPO regarding the applicability of such generic responses to your plant.
In order to preserve our present review schedule, the staff needs to have the supplementary infortnation within 60 days of receipt of this letter for Items 2.1, 2.2.2, 3.1.3, and 3.2.3, and within 90 days for Item 4.5.3. If you intend to fonnally endorse the BWR Owners Group response to Item 4.5.3 (NEDC-30844), the staff should be advised within 60 days. Your plant-specific response to Item 4.5.3 should then be provided within 90 days afte.- NRC i completes its review of NEDC-30844 and issues its evaluation. We request your cooperation in meeting this schedule. - .'..
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Mr. William D. Harrington This Reauest for Additional Information was approved by the Office of Management and Budget under clearance number 3150-0011 which expires April 30, 1985. Conments on burden and duplication may be directed to the Office of Managewent and Budget, Reports Management Room 3208, New Executive Office Building, Washington, D. C. 20503.
Sincerely, I
l Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Enclosures:
- 1. Request for Additional Information
- 2. Ltr dtd March 20, 1985, H. Thompson (NRC) to S. Rosen (INPO) cc w/ enclosures:
See next page I
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1
Mr. William D. Harringten Boston Edison Company Pilgrim Nuclear Power Statio CC' Pr. Charles J. Mathis, Station Mgr. Thomas A. Murley Boston Edison Coripany Regional Administrator RFD #1, Rocky Hill Road Region 1 Office Plymouth, Massachusetts 02360 U. S. Nuclear Reculatory Commission 631 Park Avenue Resident Inspector's Office King of Prussia, Pennsylvania 19a06
- 11. S. Nuclear Regulatory Commission Post Office Box 867 Mr. A. Victor Morisi Plymnuth, Massachusetts 0??60 Boston Edison Company 25 Braintree Hill Park Mr. David F. Tarantino Rockdale Street Chairman, Board of Selectman Braintree, Massachusetts 02184 11 Lincoln Street Plymouth, Massachusetts .02360 l
Office of the Commissioner Massachusetts Departnent of Environmental Quality Engineering One Winter Street Boston, Massachusetts 07108 I
Office of the Attorney General i 1 Ashburton Place '
19th Floor l Boston, Massachusetts 02108 l
l Mr. Robert M. Hallisey, Director l Radiation Control Program Massachusetts Departnent of Public Health i
150 Trenont Street l Boston, Massachusetts 02111 1
1 L_z-_-_-_----_---_ 1
ENCLOSURE 1 REQUEST FOR ADDITIONAL INFORMATION PILGRIM NUCLEAR POWER STAT *0N Item 2.1 (Part 1) - Incomplete Licensee needs to submit a statement that components used for reactor trip have been reviewed and are identified as safety-related on documents and in information handling systems.
Item 2.1 (Part 2) - Incomplete Licensee needs to describe its program for establishing and maintaining an interface with vendors of components used for reactor trip. Information submitted shall describe how the program assures that vendor technical infonnation is kept current, complete, and controlled throughout the life of the plant and how the program will be implemented at Pilgrim.
. Item 2.2.2 - Incomplete Same as for item 2.1 (Part 2) except that it applies to all other safety-related components.
Item 3.1.3 - Incomplete Licensee needs to state if he has found any post-maintenance testing requirements for RTS compor.ents that may degrade safety. If any such requirements are identified, the licensee shall describe actions to be i taken including submitting needed Technical Specification changes.
l' l
Item 3.2.3 - Incomplete j Same needs as for Item 3.1.3 except that it applies to all other safety-related compo;'ents.
Item 4.5.3 - Incomplete '
The staff finds that modifications are not required to permit on-line
. testing of the backup scram valves. However, the staff concludes that testing of the backup scram valves (including initiating circuitry) at a :
refueling outage frequency, in lieu of on-line testing, is appropriate and should be included in the Technical Specification surveillance l requirements. The licensee needs to address this conclusion.
Regarding the scram pilot valves (including all initiating circuitry), the l
licensee needs to provide the results of a review o' existing or proposed ,
intervals for on-line testing considering the concerns of sub-items 4.5.3.1 to 4.5.3.5 of the generic letter. The responsa should show how these intervals result in high reactor trip system availability and present proposed Technical Specification changes for staff review.
i 2
The staff has just received the BWR Owners Group response to item 4.5.3 (NEDC-30844). If the licensee intends to fortnally endorse the Owne-s Group response, the licensee should delay his plant-specific response to Item 4.5.3 until after the staff completes its review of the Owners Grom response.
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- ---- J--_ _ _ _ _ _ , _ _ _ _
ENCLOSURE 2 e
f "'%,,
- , UNITED STATES 3
- f. ',i NUCLEAR REGULATORY COMMISSION
, I wasec row. o. c. rous March 20,1985 Wr. Steven L. Rosen Vice President and Director Analysis and Engineering Division Institute of Nuclear Pover Operations 1100 Circle 75 Parkway, Suite 1500 Atlanta, Georgia 30339
Dear Fr. Rosen:
The Section staff has completed their review of the NUTAC Report on Generic Letter 8 2.2.2 The staff has found that although it does present ideas that may benefit the exchance of infomation among utilities principally through enhanced
- participation in the Nuclear Plant Reliability Data System (NPRDS) and the Significant Event Evaluation and Infomation Network (SEE-IN), it fails to address of safety-related equipment and the utilities as expressed part) and Item 2.2.2 of Generic Letter 83-28.
presented in the enclosure. The results of our review are Please contact us within 60 days of the date of this letter with a proposal on how you intend to address the concerns that have been identified. The response should indicate whether the staff's concerns will be resolved generically by
. revisions to the NUTAC Report, or addressed on a plant specific basis. We also results of the staff's evaluation. request that you infom the utilities referencing th Accordingly, the staff will be advisino those utilities to supplement their responses to describe how current procidures will (second be modified and2.2.?
part) and Item newofones initiated Generic to meet each element of item 2.1 Letter 83-28.
The staff is available to discuss the results of this evaluation with you and representatives of NUTAC. Should you have any questions, the staff enntact is John Hannon. Mr. Hannon can be reached on (301) 492-8543.
Sincerely, ugh L. ps n r., Director Division of Licensing i
Enclosure:
As stated 1 4 ]WF
- 1 ENCLOSURE REr!EW 0F NUTAc REPORT ON VETIP Our review has raised the concerns enumerated below. .
3.
The report's primary esqphasis appears to be on infonnation exchange between utilities. Also a key item in the Vendor Equipment Technical Infonnation Program (VETIP) is stated to be the development by each utility of an active faternal program to contribute infonnation to SEE-IN and NPRDS.
These do not address the concerns identified in Itsers 2.1 (second part) add 2.2.2 which address the need for.the utilities to establish direct and peHodic contact with all vendors of safety-related equipment to assure themselves that they are in possession of the latest and most up-to-date epipment technical infonnation (ETI).
2.
Page 7 states that VETIP does not rely on action by other than RSSS vendors. Action by all vendors of safety-related equipment te supply and update maintenance and test information is needed to satisfy the requirements of ! tests 2.1 (second parti and 2.2.2.
The intent of these-items is for the utility, as the user. G take the initiative in con-tacting the vendors for ETI and to screan received vendor inf.ormation for usable test and maintenance information. The program should contain reconenendations to address this intent.
_ _ _ _ _ _ _ _ . - I
3.
The second paragraph on page 8 leaves out an important step.
T,he utility as the discoverer and analyzer of the failure needs to consult with the manufacturer of the equipment to arrive at a solution to the problem t caused the failure.
The VETIP progras, by not including the vendor /manu-facturer, fails to adequately address the concerns ofr Items 21 and 2.2.2 of the generic letter.
The program should be revised to include I the vendor / manufacturer as a direct participant in the corrective proc 4
Also on page 8 reference is made to *INp0 supplier participant practic but these are not identified or described. We cannot, therefore, assess their impact on the vendor interface concerns.
These practices should be ~
described.
5.
Section 3.1.1 discusses the NPRDS and its enhancemen ;
on pages 9 and 10 it apparently fails to include infomation on tests and (
i recommended maintenance in its base of infomation. This is not l responsive to the concerns of the generic letter and the program should ' .
be supplemented to include this information.
6.
The SEE-IN program as outlined in Section 3.1.2 apparently ignores ve of equipment other than RSSS. It does sepply copies of documents generated to "affected vendors" but it is not clear that vendor participe-tion is encouraged.
This could constitute a serious lack in the program and could make it fail to meet the positions of the generic letter item 2.2.2.
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3-7 Section 3.1.3 attempts to address the utility-vendor interface, however, 3
we note the following deficiencies which should be corrected. !
(1)
The majority of the attention is given solely to the RSSS vendor interface. Interfaces with other suppliers of safety-related equipment are virtually excluded. This appears to be a serious deficiency with regard to meeting the requirements of item 2.2.2.
(2)
This sectior mentions bulletins or advisortes issued by the MSSS supplier but neglects to propose means for verifying '
that individual utilities receive this information in a time- i ly manner.
8.
Section 3.2 discusses enhancements to existing programs which include i only NPRDS and SEE-IN. A clear exposition of exactly how taplementation of the proposed enhancements will improve the quality and availability of equipment technical information should be included in the NUTAC/VETIP final report. Typical concerns to be discussed include:
. l 4
I 1
(1)
Reporting procedures for NPROS are stated to be contained in
' - l INPO Reports83-020 and 84-011. Are improvements to these pro-cedures being considered or 1sqplemented as part of this program? f If so, describe them and how t, hey will aid in accomplishing the program's objectives.
(
t (2) (a) For NPRDS - s' tate if all stilities have agreed to use these
- uniform procedures." k' (b) For SEE-IN - state if guidelines containing specific minimum
~
requirements for obtaining uniform input to the program have been prepared and if agreement by the utilities to meet these i
minimum requirements has been obtained.
(3) Are the results of studies and analysis of the failure data con-tained in the NPRDS database made available to manufacturers of -
l affected safety-related equipment on a regular basis to aid in improvement of equipment design and reliability?
(4)
Also the failure reporting guidamce given on page 17 should be j
revised to state that utilities should supply information when l
inadequate, missing, or faulty maintenance and test practices are identified as a contributory factor in a failure. t 1
. \
o
9.
On page 20 reference is made to'the NS$$ vendor technical bulletins as a pathway that would have ensured that maintenance and test information reached the utility. This ignores the fact that from 1974 thru 1982'
{
a vendor technical bulletin program existed yet the utility did not re-ceive maintenance information revisions for the RTB's during that period and apparently never asked for them. The main thrvst of items 2.1
~
(second part) and 2.2.2 is concerned with the establishment of contin-uing interfaces 'between al'1 vendors of safety related equipment and utility to assure that incidents like this do not reoccur.
~
- 10. Section 4.1.1.1: *
(a) N555 Vendor Contact - Needed improvements are not identified here.
The statement is too general. The types of necessary direct vendor contact are not discussed or referenced. Specific improvements to
, this program should be described such as periodic contact with vendor, verification of receipt of bulletins, periodic verification of up-to-date status of utility's ET1, etc. '
(b) For other vencors - The utility's program should seek assistance and ETI on a regular periodic basis and receipt of such informa-tion should be documented and verified. The NPCS and SEE-IN programs will not resolve this concern.
-6 (c)
The administrative procedural changes outlined on page 22 should also apply to test procedures.
(d)
The administrative program requirements presented on page.
22 sho 1d also apply to test procedures.
, 11.
The VETIP Stock Diagram presented in figure 1 should be clarified by (
indicating the directions ,of faformation flow. Clarify the purpose of the lowest box of the figure and clarify the role of the owner's Groups
. in the VETIP.
12.
For the SEE.!N Functions listed in Appendix C. clarify whose function
'it is to perform the action analysis, who determines the s10nificant events to evaluate, and how these functions are related to the vender interface program.
1
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- ., UNITED STATES h j >m
'G.-* ,
[ p, NUCLEAR REGULATORY COMMISSION
- j WASHINGTON, D C 20555 o., s
~
..... June 17, 1985 l
Docket No. 50-293 1
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l l Mr. William D. Harrington Senior Vice President, Nuclear l Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Dear Mr. Harrington
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION RELATIVE TO b- -- --
GL83-28, ITEM 1.1 (POST-TRIP REVIEW) t .
By letter dated November 7,1983, you responded to Generic Letter 83-28
. with regard to required actions based on generic implications of the Salem
. ATVS events. We have reviewed your response with respect to Item 1.1 i
! (post-trip review) and find that it does not fully meet our guidelines in the following areas:
1 C. The methods and criteria for comparing the event with expected plant performance.
D. The criteria for the need of independent assessment of the event. l 1
i T-E. A systematic safety assessment program to assess unscheduled j
k The guidelines for the above areas and further details relative to our
, request for information are provided in the enclosure. Please respond 7 within 60 days of receiving this letter. .
The reporting and/or recordkeeping requirements of this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511. l l
Sincerely, Domenic B. Vassallo, Chief 2 Operating Reactor Branch #2 ~
Division of Licensing . l
Enclosure:
As stated cc w/ enclosure: '
See next page e f % 4 f(o 4 r k y -
- _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ - - - - _ - _ - - _ - - - - -- -- nw
I I
Mr. William D. Harrington Boston Edison Company )
Pilgrim Nuclear Power Station ,
1 ~
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CC: "
Mr. Charles 61, Mathis, Station Mgr. Thomas A. ftriey Boston Edison Company Regional Administrator RFD #1, F.ocky Hill Road Region 1-Office Plymouth, Massachusetts 02360 U. S. Nuclear Regulatory Commission 631 Park Avenue Resident Inspector's Office King of Prussia, Pennsylvania 10406-1 U. S. Nuclear Regulatory Comission Post Office Rox 867 Mr. A. Victor Morisi Plyinnuth, Massachusetts 0?360 Boston Edison Company ,
i L. .. ... .
25 Braintree Hill Park __
I A Mr. David F. Tarantino Rockdale Street Chairman, Board of Selectman Braintree, P' massachusetts 021'84 11 Lincoln Street-
, Plymouth, Massachusetts 02360 ]'
2 Office of the Corsnissioner
. Massachusetts Department of Environmental' Quality Engineering One Winter Street Boston, Massachusetts 07108 l
. Office 'of the Attorney General a 1 Ashburton Place j p._ 19th Floor 5
,. Boston, Massachusetts 0?)08 p'. Mr. Robert M. Hallisey, Director Radiation Control Program C . Massachusetts Departnent of -
"o Public Health
- 150 Tremont Street Boston, Aa'ssachusetts 02111 e
l ENCLOSURE REOUEST FOR ADDITIONAL INFORMATION _
PILGRIM NUCLEAR POWER STATION- i
~r SALEM ATWS ITEM 1.1 - POST-TRIP REVIEW I' I. Peview Guidelines The following review ouidelines were developed after initial evaluati n of various utility responses to Item 1.1 of Generic Letter 83-28 and in:ceporate -
- 'the best features of these submittals. As such, these review guidelfres in effect represent a " good practices" approach to post-trip review:.
A. The Boston Edison Company (licensee) should have systematic safe::
assessment procedures established that will ensure that the follow ng
.- restart criteria are met before restart is authorized.
5
~
- The post-trip review team has determined the root cause and sequence of events resulting in the plant trip.
- Near term corrective actions have been t'aken to remedy the cause of the trip.
The post-trip review team has performed an analysis and l Ti- determined that the major safety systems responded to .the
'. event within specified limits of the primary system paramete-s..
- The post-trip review has not resulted in the discovery of a pa potential safety concern (e.g., the root cause of the ever:
- 7 , . occurs with a frequency significantly larger than expecte(). ,
- If any of the above restart criteria are' not met, then an 1 independent assessment of the event is performed by the P; ant Operations Review Cam-ittee (PORC), or another designated group with similar autnority and experience.
B. The responsibilities and authorities of the personnel who will peeform the review and analysis should be well defined.
The post-trip review team leader should be a member of plant management at the shif t supervisor level *or above and should hold or should have held a Senior Operator license on the plant. The term leader should be charged with overail responsibility for directing the post-trip rev'iew, incluc,r:
data gathering and data assessment and he/she should have t~e necessary authcrity te obtain all personnel and data need?f for the post-trip review.
i
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ __ i
2 A second pe-son on the review tean shouN ce a Shht Technical Ace'ser (STAi or should he'd a relevan+. e-- neeridrf degree with s:echl transient analysis trainino.
A team leader and the STA (Engineer) shcu'd be responsible to concur on e. decision / recommendation to restart the plart.
A nonconcurrence from either of these persons should be sufficient to prevent restart until the trip has been reviewed by the Plant Operations Review Committee (PDFC) or equivalent n ganization.
C.
The licensee should indicate that the plant response to the trip i
y _, event will be was evaluated and a determination made as to whether thei plant response within acceptable limits.
t include: The evaluation should, l ,
A verification of the proper operation of plant systems and equipment by comparison of the pertirant data E during the post-trip review to the applicable data l* provided in the Final Safety Analysis Report (FSAR).
An analysis of the sequence of events to verify the proper functioning of safety-relsted and other important equipeent.
Where possible, comparisons with previous similar events should be made, a~ D.
- The licersee should have procedures to ensure tnat 11 physical evidence necessary for an independent assessmert is preserved.
E.
! 3.a Each licensee should provide in its submittal, copies of the plant precedures which contain the information recuired in Items A through D. As a minimum, these should include the followino: -
The restart criteria for determining the acceptability of The qualifications, responsibilities and authorities of key personnel involved in the post-trip review process The nethods and criterie fot determining vbether ths:
plant variables and system responses were within the-limits as described in the FSAR -
The review c-iteria for Jetemining the need for en independent
~
II. Additional Information Requested We have reviewed the licensee's November 7,1984 submittal egainst the above guidelines and we found the information responsive tdfareas I.A.
and I.B. However, the following information is needed for completion of our review in areas I.C., I.D., and I.E.:
C. The methods and criteria for comparing the event information with known or expected plant behavior should be addressed. We recommend that the pertinent data obtained during the post-trip review be compared to the applicable data in the FSAR. Where possible, comparisons with previous similar events should be made.
The licensee has established procedures to ensure that all D ~.
,L- . __ _ _ . . . physical evidence necessary for an independent assessment is ~ __
S-preserved. The licensee has also indicated that if the cause of '
the trip is bnknown, an independent assessment conducted by the operations Review Committee is required for the event. We find
, this insufficient and recommend that an _ independent assessment should be performed if any of the restart criteria are not net.
E. The licensee should develop, and provide for our review, a ~*
- , systematic safety assessnent program to assess unscheduled reactor trips. -
4 :
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E* 5 O@] I June 28. 1985 /
NM SEco 85-119 30 k.b.%..L.dd s-O~ 0,; , - ..
Mr. Domenic 8. Vassallo. Chief Operating Reactors Branch #2 Division of Licensing l
ygC m gk Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission CMDOMNT # _
washington. D. C. 20555 -
' ", g-License DPR-35 Docket 50-293 .,
s.
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Subject:
NRC Request for Additional Inf ormation Following Preliminary --
Staf f Review of Licensee Response to Generic Letter 83-28 dated
~
April 23. 1985
References:
(1) Boston Edison Company Response to Generic Letter 83-28 SECa ' tr. #83-215. detec November 7.1983 (2) Bosto. Edison Company Response to Generic Letter 83-28 l
- ' ' # 'j.'
Section 2.2.2. Vendor Interf ace. BEco Letter #64-061 asAME dated O fil 21. 1984 'f Dear Sir; 356 ! I Boston Edison Company (BECo) received the subject letter on May 2,1985, b
requesting additional information on Generic Letter 83-28 within 60 days f rom "
A/ST Jj the receipt of NRC letter. SECo previously submitted responses to GL 83-28 by {i Reference (1) and (2). The attachment provides our response to the subject NRC request on GL 83-28. As stated in the attachment, response to item 4.5.3 ~< Md87 q will be provided by August 1.1985, following SECo evaluation of BWROG Response on item 4.5.3. Should you desire any further information on our q j,_,
response, pleasa contact us. _
very truly yours, ,
I M f-
"~2 -
I I D J Attachment l
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4Ei. EASED JUL 17 885 i <
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ATTAC *ENT BECo RESPONSE TO NRC REQUEST TOR ADDITIONAL INFORMATION TO GENERIC LE~TER 83-28 A. NRC REQUEST 4-Item 2.1(Parti) - Incomplete I
Licensee needs to submit a statement that components used for reactor tric have been reviewed and are identifie3 as safety-related on documents and in information handling systems.
BECo RESPONSE As stated in Reference (1) Item 2.2.1.h,componentswithinsystems classified as safety related are themselves considered safety-related and are Mcluded in our Q-list. This Q-list includes Safety-related components of RPS. ,
. B. NRC REQUEST
. Item 2.1 (Part 2) - Incomplete Licensee needs to describe its program for establishing and maintaining ar interface with vendors of components used for reactor trip. Information submitted shall describe how the program assures that vendor technical information is kept current, complete, and controlled throughout the life of the plant and how the program will be implemented at Pilgrim.
Item 2.2.2 - Incomplete Same as for Item 2.1 (Part 2) except that it applies to all other safety-related components.
BECO RESPONSE SECo receives and reviews correspondence from the NSSS Vendor in accordance with our currect establishe:: practice on GE correspondence for applicability to safety-related equipment repairs, maintenance and operations. With regard to all other safety-related components, the Vendor Equipment Technical Information Program (VETIP) as defined in the March 1984 NUTAC document is considered a valid response to Section 2.2.2 of the Generic Letter 83-28. BECo is in the process of addressing those elements of VETIP which would supplement our ongoing Operations Experience Review and Vendor Manual Control Programs (See Ref. 2) for an effective and efficient Vendor Interface Program.
C. NRC REQUEST Item 3.1.3 - Incomplete ,-
Licensee needs to state if he has found any post-mainteriance testing requirements for RTS components that may degrade safety. If any such requirements are identified, the licensee shall describe actions to be taken including submitting needed Technical Specification changes.
L_______- --
Item 3.2.3 - Inc r:'ete Same needs as for Item 3.1.3 except that it applies to a'1 other safety-related cyrc.onents.
BECO RESPONSE To date BECo has not found any post maintenance testin; requirements that degrade safety. Recently, BECo has completed a HPCI reliability study.
The results ;* this study are being evaluated to deternine changes in Technical Specifications. BECo is also investigating : e emergency diesel generator performance to optiml;' the performance. If trese studies identify any post-maintenance testing that may degrade sa'ety, BECo will propose to amend technical specifications.
D. NRC REQUEST Item 4.5.3 - Incomplete The staff finds that modifications are not required to :ermit on-line testing of the backup scram valves. However, the staff concludes that testing of the backup scram valves (including initiattrq circuitry) at a refueling outage frequency, in lieu of on-line testing. is appropriate and should be included in the Technical Specification survefitance requiremen t s . The licensee needs to address this conclashon.
~
Regarding the scram pilot valves (including all initiating circuitry), the licensee needs to provide the results of a review of ents ing or proposed intervals for on-line testing considering the concerns of sub-items 4.5.3.1 to 4.5.3.5 of the generic letter. The response smould show how these intervals result in high reactor trip system avaitamility and present proposed Technical Specification changes for staff review.
The staff has just received the BWR Owners Group resporse to Item 4.5.3 (NEDC-30844). If the licensee intends to formally endcrse the Owners Group response, t,e licensee should delay his plant-spe:ific response te Item 4.5.3 until af ter the staff completes its review cf the Owners Group response.
BECo RESPONSE BECo is currently evaluating BWROG Response to Item 4.5.3 (NEDC-30844) and will advise you regarding its applicability to PNPS by August 1,1985.
Also, our response to the remaining items of 4.5.3 will be submitted at that time.
i 4
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3 6L BV Rb S p tv3 3 0. TON Ect.ON C OM P ANY
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. . . . . . .--. . . . . . I RfCO:' TC* j @. oJ
-*a August 23. 1985 !
BECo 85-156
' Q: n ~ p",' . . e ,' ' 3 FEii'r - Mg Mr. Domenic B. Vassalto. Chief Operating Reactors Braaca #2 C1 vision of Licenstrg NII TM h G Office of Nuclear React::rr Regulation i C- .W '- O 9E .3. ~1 U.S. Nuclear Regulat:ry ommission I Mashington, D. C. 2C555 gg ] g License DPR-35 ,, .,
Docket 50-293 - .I *, _ _ _
Subject:
NRC Re:uest for Additional Information following Preltaleary #
Staff Restem of Licensee Response to Generic Letter 83-28, dated e
^
April 23, 1985
/UM1b72
' '~ ~
References:
(1) Scston Edison Company Response to Generic Letter 83-28.
Regarding NRC Recuest for Additional Information. SECo -
Le:ter #85-119. dated June 28, 1985 (2) Boston Edison Company Response to Generic Letter 83-28, Section 4.5: Functional Testing of Backup Ecram Valves, BE;o Ltr. s34-093, dated June 28, 1984 N vC u Aw m,3 m t.,s.
Dear Sir:
"A"E "#
Boston Edison Company (Deco) received the subject letter on May 2,1985, requesting additional Information on Generic Letter 83-28. BEco previously 588 b l
submitted a response to the NRC request on GL 83-28 by Reference (1) stating that a response to Ites 4.5.3 mill be provided by August 1.1985, following h/37' BEco evaluation of 84 T. Response on Ites 4.5.3.
J On July 18, 1985, in a telephcae conversation with the NRC Project Manager, Mr. Paul Leech and the staff reviewer, Mr. Don Lasher, BECo requested the dM((
rationale or NRC position regarding the testing of backup scram valves. At u that time NRC Project Manager Indicated that BEco response to Item 4.5.3 of GL i 83-28 can be delayed to address the Staff position on backup scrams. ,
Subsequently. BEco has received and evaluated the NRC position on bachT, scrans along with the 8WROG Report NEDC-30844. Our response to the NRC request on Item 4.5.3 is as follows:
C C kf0L,EST !
Ites 4.5.3 - Incomplete i
The staff finds that modifications are not reautred to permit on-Ilne testing ~1 of the tackup scram valves. However, the staff concludes that testing of the '
backup scras valves (including initiating circuitry) at a refueling outage l OCUWENY ~
- RELEASED l
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FOR USE becstanc~ ~
-g Fogs 44+q.- b-rp j o
p 50STON EDISON COMPANY H . Domenic B. Vassallo, Chief .
August 23, 1985 Page 2 -
1 frequency, in lieu of on-line testing, is appropriate and should be 1 :'uded in the Technical Specification surveillance requirements. The licensee .ee35 to address this conclusion.
Regarding the scram pilot valves (including all initiating circuitry). :ne licensee needs to provide the results of a review of existing or prop:ses intervals for on-line testing considering the concerns O' sub-items 4.5.3.1 and 4.5.3.5 of the generic letter. The response should show how these intervals result in high reactor trip system availability and present croposec Technical Specification changes for staff review.
The staff has just received the BNR Owners Group response to Item 4.5.3 (NEDC-30844). If the licensee intends to formally endorse the Owners Group response, the licensee should celay his plant-specific response to Item 4.5.3 until after the staff completes its review of the Owners Group response.
BECo RESPONSE BECo performs on-line functional testing of the reactor protection system, including the independent testing of the diverse trip features, in at:ordance with the Technical specifications Section 3/4.3.B.
With regard to the RTS reliability, BECo hereby endorses the BHR Owne s Group Report NEDC-30844, as applicable to Pilgrim Nuclear Power Station. Ttis report demonstrates a high reliability of the RTS in a generic basis :aking into account the scram system and on-line testing freque,ncies. The o -line testing frequencies modeled for the RTS reliability are representative of PNPS frequencies included in the PNPS Technical Specifications. Therefore. SECo has concluded that nc changes to the existing Technical Specifications are recommended at the present time.
Regarding the backup scram testing (See Ref. 2), BECo's position is as follows:
- 1. The addition of backup scram valves to the BWR scram system was irade by GE on the basis that such an addition was desirable though not essen:ial.
The backup scram valves are not required to meet any transients, anc credit for them is not taken in Pilgrim't FSAR. The system is interded to provide an alternate source of rod insertion in cases where individLal rods fall to insert. The probability of enough rods independently failing in quantitles sufficient to prevent shutdown is negligible; there' ore the
, {
use of backup scram valves is not considered essential to.the safe ;
shutdown of the plant, and provides no increase in the safety mar;ir:. j
- 2. The BHROG Report NEDC-30844 demonstrates a high reliability of the :.TS without taking credit for backup scram testing.
30570N EDI DN COMPANY l'
Mr. Domeni: B. Vassallo, Chief August 23, 1985 Page 3 i
- 3. BECo does not routinely conduct a specific sne'l'ance test for the backup scram valves, nor do we feel that spe:'ft: :esting is required.
However, valve 00eracility is indirectly demccs: rated by PNPS Procedure No. 2.1.5, " Controlled Shutcown from Power" 7-is procedure reautres that the scram pilot valve air header alarm clear :efore certain other activities proceed. This demonstrates bacKuc scram valve operaollity because the alarm sensing circuit is configurec such that no alarm will be received if the bacKuo scram valves fail to ennaust. If the valves then fall to return to their normal position, the air header will not repressurize and the alarm will not clear. T91s verification of valve operability is performed in the interest of equipment reliability and not plant safety.
- 4. BECo has implemented the preventive maintenan:e recommendations of GE SIL-128 to the scram pilot valves and backup scram valves.
- 5. Also, we believe scram diversity is enhanced at Pilgrim by the Alternate Rod Insertion (ARI) function of the Recircula:lon Pump Trip System (RPT).
This system is not redundant to the Reactor Procection System (RPS), but does snare some common parameters with the RPS logic. ARI/RPT is intended to scram the reactor for certain undesirable transients which were not covered by the RPS. The RPT/ARI system is surveilled in accordance with Table 4.2-G and 3/4.2G of Pilgrim's Technical S;:ecifications.
- 6. BECo has reviewed the NRC Staff position, "Ge9eric Safety Evaluation '
Report en BWR Backup Scram Testing". He concur with the staff that, as a non-safety related design feature, the backup scram is not a diverse scram system in the context of a diverse protective system function as addressed by GOC-22. " Protection System Independence", nor would it be required to satisfy GDC-21, " Protection System Reliability and Testability", with respect to inservice testability. BECo review Of the staff position concerning the safety implication of testing incicates that the position does not provide any evidence which would sucport the testing of backup scrams or their inclusion into the Technical Specifications.
Therefore, BECo has concluded that backup scram valve testing is not required at the present time, and BECo plans no further action concerning such testing.
This lotter completes our response to Item 4.5.3 of GL 83-28.
Should you desire any further information on our response please contact us.
Very truly yours.
WGL/kmc
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,I UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 September 11,1985 M L" 5118o Docket No. 50-293 Mr. William D. Harringten Senior Vice Presidert. Nuclear Boston Edison Comps.sy 800 Boylston Street Boston, Massachusetts 02199
Dear Mr. Harrington:
SUBJECT:
IMPLICATIONS OF SALEM ATWS EVENTS -
ACTION ITEM 1.1 - POST TRIP REVIEW Re: Pilgrim Nuclear Power Station i
We have completed our review of your November 7,1983 and August 13, 1985 submittals relative to Action Item 1.1 of the Salem ATWS Events which you provided in response to our Generic Letter No. 83-28.
Based on our review, we conclude that the Post Trip Review Program and Procedures for Pilgrin Station are acceptable. J l
Our Safety Evaluation supporting our conclusions is enclosed. j i
Sincerely, l 1
Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
See next page ;
i
& [qb9..l_6f & N 01 'Y l
b
Mr. William D. Harrington Boston Edison Company Pilgrim Nuclear Power Station CC:
Mr. Charles' J. Mathis, Station Mgr, Boston Edison Company RFD #1, Rocky Hill Road Plymouth, Massachusetts 02360 Resident inspector's Office U. S. Nuclear Regulatory Commission Post Office Box 867 Plymouth, Massachusetts 02360 Mr. David F. Tarantino Chairman, Board of Selectman 11 Lincoln Street Plymouth, Massachusetts 02360 Office of the Commissioner Massachusetts Department of Environmental Quality Engineering One Winter Street Boston, Massachusetts 02108 Office of the Attorney General 1 Ashburton Place 19th Floor ,
Boston, Massachusetts 02108 Mr. Robert M. Hallisey, Director j Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street -
Boston, Massachusetts 02111 i
Pegional Administrator, Region I U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Mr. A. Victor Morisi l' Boston Edison Company 25 Braintree Hill Park Rockdale Street Braintree, Massachusetts 02184 1
l
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UNITED STATES
! , g NUCLEAR REGULATORY COMMISSION l WASmNGTON D. C. 20555 k....+,/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION IMPLICATIONS OF SALEM ATWS EVENTS, ITEM 1.1 DOST-TRIP REVIEW BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293
1.0 INTRODUCTION
On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant (SNPP) failed to open upon an automatic reactor ;
trip signal fran the reactor protection system. This incident occurred during the plant startup and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined.to be related to the sticking of the under voltage trip attachsnent. On February ,
22, 1983, during startup of SNPP, Unit 1, an automatic trip signal occurred Os the result of steam generator low-low level. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip. Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (ED0) directed the staff to investigate and report on the generic implications of these occurrences. The results of the staff's inquiry into these incidents are reported in NUREG-1000
" Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Connission regnested (by Generic Letter 83-26 dated July 8,1983) all licensees of operating reactors, applicants for an operating license, and holders of construction pennits to respond to certain generic concerns. These concerns are categorized into four aress:
(1) Post-Trip Review, (2) Equipment Classification and Vendor Interface, (3) Post-Maintenance Testing, and (4) Reactor Trip Systern Reliability Improvements.
The first action item, Post-Trip Review, consists of Action Item 1.1,
" Program Description and Prc;edure," and Action Item 1.2, " Data and Infonnation Capability." This Safety Evaluation (SE) addresses Action Item 1.1 only.
i 2.0 EVALUATION The following review guidelines were developed after initial evaluation of various utility responses to Item 1.1 of Generic Letter 83-28 and incorporate the best features of these submittals. As such, these review guidelines, in effect, represent a " good practices" approach to post-trip 1 review. We have reviewed the licensee's response to : tem 1.1 against these !
guidelines:
35 .L & gf 427 kM
2 A. The licensee or applicant should have systematic safety assessment procedures established that will ensure that the following restart criteria are met before restart is authorized.
o The post-trip review team has detemined the root cause and sequence of events resulting in the plant trip.
o Near term corrective actions have been taken to remedy the cause of the trip.
o The post-trip review team has performed an analysis and determined that the major safety systems responded to the event within specified limits of the primary system parameters.
o The post-trip review has not resulted in the discovery of a potential safety concern (e.g., the root cause of the event occurs with a frequency significantly larger than expected).
o If any of the above restart criteria are not met, then an independent assessment of the event is performed by the Plant Operations Review Connittee (PORC), or another designated group with similar authority and experience.
B. The responsibilities and authorities of the personnel who will perfom the review and analysis should be well defined.
o The post-trip review team leader should be a member of plant management at the shift supervisor level or above and should hold or should have held a Senior Reactor Operator (SRO) license for the plant. The team leader should be charged with overall responsibility for directing the post-trip review, including data gathering and data assessment and he/she should have the necessary authority to obtain all personnel and data needed for the post-trip review.
o A second person on the review team should be a Shift Technical Advisor (STA) or should hold a relevant engineering degree with special transient analysis training.
o The team leader and STA (Engineer) should be responsible for I concurring on a decision / recommendation to restart the plant. A nonconcurrence from either of these persons should be sufficient to prevent restart until the trip has been reviewed by the P3RC or equivalent organization.
C. The licensee or applic6nt should indicate that the plant response to the trip event will be evaluated and a determination made as to whether the plant response was within acceptable limits, The j
evaluation should include:
o A verification of the proper operation cf plant systems and equipment by comparison of the pertinent data obtained during the post-trip review to the applicable data provided in the Final Safety Analysis Report (FSAR).
o An analysis of the sequence of events to verify the proper functioning of safety related and other important equipment.
Where possible, comparisons with previous similar events should be made.
D. The licensee or applicant should have procedures to ensure that all physical evidence necessary for an independent assessment is preserved.
E. Each licensee or applicant should provide, in its submittal, copies of -
the plant procedures which contain the infomation required in Items A through D. As .a minimum, these should include the following:
o The criteria for determining the acceptability of restart, o The qualifications, responsibilities and authorities of key j personnel involved in the post-trip review process. 1
)
o The methods and criteria for determining whether the plant variables and system responses were within the limits as described in the FSAR.
o The criteria for detennining the need for an independent review.
I By letters dated November 7,1983 and August 13, 1985, the licensee for Pilgrim Station provided information regarding its Post-Trip Review Program and Procedu % We hose evaluated the licensee's program and procedures against the soo've review guidelines. A brief description of the licensee's response and the staff's evaluation of the response against each of the review guidelines is provided below:
A. The licensee has established the criteria for determining the acceptability of restart. We find that the licensee's criteria conform to the guidelines described in Sectier. A above and, therefore, are acceptable.
B. The qualifications, responsibilities and authorities of the personnel 4 who will perform the review and analysis have been clearly described.
We have. reviewed the licensee's chain of corrtand for responsibility for post-trip review and evaluation and we find it acceptable.
C. The licensee has described the methods and criteria for comparing the event information with known or expected plant behavior. Based on our review, we find the licensee's methods to be acceptable.
_ - - - _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ - L
I i,...
4 D. The licensee has established the criteria .for determining the-need for an independent assessment conducted by the Operation Review Committee. In addition, the licensee has established procedures to ertsure that all physical evidence necessary for an independent assessment is preserved. We find these actions conform with the guidelines described in Sections A and D above.
E. The licensee has provided a systematic safety assessment program to assess unscheduled reactor trips. We have reviewed this program and find it acceptable.
3.0 CONCLUSION
S Based on our review, we conclude that the licensee's Post-Trip Review Program and Procedures for Pilgrim Station are acceptable.
Principal Contributor: D. ' Sh um l
l Dated: September 11, 1985 l
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2 March 24, 1986 BECo 86-034 CONTROL COPY RECORD TYPE: M.05" Mr. John A. Zwolinskt. Director 8WR Project Directorate #1 01 vision of Licensing QA N0fl QA O office of nuclear Reactor Regulation N1 D Y o'ss KEYWORDS:%Tm Ram-Tan Mb 04 License DPR-35 i Doctet 50-293 CONTROLLED DISTRIBUTl0er Long Tere Program - Seelannual Update
Dear str:
COMPOKI.1T #:
In accordance with Section V.A of the " Plan for the the Long Tere Program -
Pilgrie helear Power Station', Boston Edison Company herein provides the seelannual update to the Long Tere Plan. This consists of tfie following: QLI Attachment 1: Schedule A and 8 and Additional ! tees RMG CONTROL f" Attachment 2: Comalteent Description Attachment 3: Progress Since June 1985 Update Attachment 4: Summary of Changes and Reasons Associated With Regulatory Requirements We have provided a list of additional 1 tees that could have a potential influence on Schedule B ltees. This additional 11st 15 outslee the regulatory scope of the Long Tore Program and thus exempt from the license conditions imposed on Schedule B ltees. This list is included to show parallel work AN A Itees which are integrated into the total "living schedule *. and to justify i the schedules proposed for NRC desired ltees. NA*AE *f ;
He trust the above information and attachments are informative and adequate Sa.s ( }
for your review. However, thould you require additional reformation or .~
clarification. please do not hesitate'to contact us.
very trui,yoers, Md {*7- _
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(Attachment 2) ma ge 1 of 6 Commitment Description i SCHEDULE A Item Description Reference Document (s)
Appendix R Installing an underground ductline NRC Letter dated 1/17/83 for rerouting cables out of areas where redundant cables could be lost due to an exposure fire.
Installing raceways inside the process NRC Letter dated 1/17/83 building to reroute the cables.
Installing cables in the installed duct- NRC Letter dated 1/17/83 line and rerouted raceways.
Modification to the alternate shutdown NRC Letter dated 1/17/83 panels for the Diesel Generator.
Relocation of alternate shutdown panel NRC Letter dated 1/17/83 for ADS system to provide se::aration between redundant ADS systers.
Alternate Shutdown System Addition - 1) NRC Letter dated 1/17/83 Torus Water Level and Temperature Monitoring
- 2) BECo Ltr. #83-70 dated March 11, 1983 Installing sprinkler lines in Reactor NRC SER dated 12/18/84 Building Elevation 23'. 1) between Fire Zones 1.9 & 1.0, 2) between Fire Zones 1.11 & l .12.
ATHS Rule Compliance New rule issued by NRC requires Standby 1) 10CFR50.62 Liquid Control System modifications by the second refueling outage from June 26, 2) BECo Ltr. #85-140 dated 1984 (i.e. RF0 #8 for BECo). August 7, 1985 BECo has evaluated several options to 3) NRC Ltr. from Mr. H. R.
comply with the Standby Liquid Control Denton to BWROG Chairman System 86 GPM control capacity requirements. dated August 19, 1985 Based on an NRC clarification to the BWROG concerning existing technical specification 4) BECo Ltr. #85-178 dated adequacy, SEco submitted a plan and schedule October 7, 1985 -
on October 7,1985, stating that the Standby ATHS Implementation Liquid Control System modification is Schedule targeted for completion during RF0 #8, any future changes to the ATHS modification schedule, if necessary, will be made through our LTP Semi-annual submittals.
f.
, (Attachment 2 continued) Page 2 of 6 Commitment Description j
l SCHEDULE B i Item Description Reference Document (s)
IE Bulletin 80-11 Per additional BECo engineering and plant 1) IE Bulletin #80-11 reviews, a limited number of safety-related 2) BECo Ltr. #84-55 dated walls and masonry blockouts located in April 13, 1984 reinforced concrete walls are being 4) BECo Ltr. #85-038 dated mod i fied. The schedule shown is February 21, 1985 contingent on receipt of SER documenting 5) BECo Ltr. #85-134 dated NRC approval of BECo analytical July 26, 1985 me thodology. (Per reference 6) 6) BECo Ltr. #85-230 dated December 31, 1985 IE Bulletin 84-02 Per BECo review of bulletin, hundreds of 1) IE Bulletin #84-02 HFA relays were replaced during RF0 #6. 2) BECo Ltr. #84-110 dated ,
Only 6 " Category 4" relays remain to be July 17, 1984 l replaced. They are scheduled to be '
replaced during RF0 #7.
Hydrogen Water Chemistry BECo conducted initial feasibility studies 1) BECo Ltr. #84-146 dated and is continuing the studies with a September 11, 1984 Test Program. The test program will 2) NRC Ltr. #1.84.357 dated provide detailed hydrogen / oxygen 12/4/84 control data as well as interim materials protection. The primary hydrogen control system remains targeted for completion during RF0 #7. Should the primary system schedule slip, the test system will be available.
BECo is currently evaluating and expanding the scope for this program to include overall materials protection activities; 1.e., improved water chemistry monitoring and control hardware.
CROR BECo will install new operator workstations 1) NUREG 0737: Item I.D.1 in the control room, as the first of these 2) BECo Ltr. #84-159 dated improvement actions developed under Control September 24, 1984 Room Design Review. The workstations will 3) HRC Ltr. #85-157 dated accommodate the new plant computer system May 16, 1985 as well as resolve human engineering 4) NRC Ltr. #86-002 dated discrepancies. A meeting was held January 6, 1986 August 21, 1985 to resolve NRC SER consnents on BECo's submittal. BECo is ,
developing a response to Reference 4.
1 l
, (Attachment 2 continued) Page 3 of 6 Commitment Description SCHEDULE B (continued)
Item Description Reference Occament(s)
_ Regulatory Guide 1.97 Per CECO evaluation of compliance require- 1) NRC Generic Ltr. #82-33 ments, modifications will be made by end 2) BECo Ltr. #84-187 dated l of RF0 #8. BECo is preparing a response to November 1, 1984 the NRC Request for Additional Information 3) NRC Ltr. #85-372 dated (NRC Letter NO.85-372) on a schedule December 12, 1985 to be negotiated with the NRC Project Manager.
SPDS A Safety Parameter Display System to provide 1) NUREG 0737: Item I.D.2 l a concise display of critical plant operating 2) Generic Ltr. #82-33 variables will be implemented at PNPS. 3) BECo Ltr. #84-133 dated The plant computer is being replaced, and August 10, 1984 SPDS has been tied in with this, although 4) NRC Ltr. (SER) #85-100 the computer upgrade is strictly voluntary dated March 21, 1985 and is not an NRC commitinent. The 5) BECo Ltr. #85-107 dated December 1986 completion date has been June 13. 1985 extended to March 1987 due to extended pre-operational testing requirements.
Analog Trip System
'BECo is replacing the mechanical swt tches 1) BEco Ltrs.84-200 & 85-099 in the RPS with an analog system. dated 11/28/84 & 5/28/85
- 2) NRC Ltr. dated 6/11/85 Reactor Water Level BECo is relocating the reference leg 1) NRC Generic Ltr. #84-23 outside the drywell. dated 10/26/84
- 2) BECo Ltrs.84-200 & 85-099 dated 11/28/84 & 5/28/85
- 3) NRC Ltr. dated 6/11/85 Hydrostatic fests Complete third phase of testing. 1) BECo Ltr. #83-163 dated June 27,1983
- 2) BECo Ltr. #84-066 dated May 7, 1984 Augmented Inspection Program of Piping Susceptible to IGSCC An augmented inspection program of 1) NRC Generic Ltr. #84-11 l piping susceptible to IGSCC will. be dated 4/19/84 l developed. The draf t copies of the report 2) BECo Ltr. #84-074 dated will be submitted to the NRC for comments. 8/4/84 This final report will be submitted to the NRC for their review 30 days before RF0 #7.
L __ __ -
, (Attachment 2 continued) Page 4 of 6 Commitment Description SCHEDULE B (continued)
Item Description Reference Document (s) lE Bulletin 85-03 This I&E Bulletin requests licensees to develop and implement a program to ensure that switch settings on certain safety related MOV's are selected, set and maintained correctly. The schedule to accomplish the above cannot go beyond 15 November 1987.
The full impact of IEB 85-03 on PNPS 1) NRC IE Bulletin #85-03 is being evaluated. A schedule will be dated 11/15/85 developed and submitted to the NRC at the conclusion of this evaluation to address the concerns identified in the bulletin.
EOF BECo will construct a suitable Emergency 1) NRC Ltr. #83-269 dated Offsite Facility (EOF) by June 1986. November 3,1983
- 2) BECo Ltr. #84-007 dated January 10, 1984
- 3) BECo Ltr. #84-050 dated April 5, 1984
- 4) BECo Ltr. #85-164 dated September 10, 1985.
a
, (Attachment 2 continued) Page 5 of 6 i Commitment Description ADDITIONAL ITEMS Item Description Refererce Document (s)
Fuel Pool Re-rack 1 BECo will replace the existing spent fuel N/A racks in the fuel pool with higher-density racks to provide storage for the full existing licensed amount of spent fuel.
Radwaste Betternent Program The scope of this program is being N/A developed, and currently includes the replacement of existing cartridge-type fuel pool filters with a back-flushable filtration system. The radwaste concentrator is being removed. The objectives of the program are:
- 1) to decrease personnel radiation exposure
- 2) to reduce the volume of radwaste shipments; and 3) to increase system reliability The replacement of the fuel pool filter is progressing well.
Plant Computer Replacement Plant process computer will be updated with N/A state-of-the-art hardware /sof tware.
Additional testing has been identified that will extend the coupletion date to Mar.:h 1987.
Co6rol Room Recorders Riplace obsolete Contro'I Room recorders N/A wi+h more reliable recorders.
l l
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~
( Atta: ment 2 continued) Pagg 6 of 6 Cocur : ment Description ADDITIONAL ITEMS (continued)
Item Description Reference Document (s)
Refueling Bridge Replacement BECo will install a more reliable refueling N/A ,
bridge to match Pilgrim's track gauges and '
other critical dimensions.
Secondary Containment Damper Repla:ement BECo will perform a complete change-out of 1) BECo Ltr. #85-84 the dampers using a more reliable model. (LER #85-10-00)
RCIC Ramp Generator This project has been cancelled.
BWR Nuclear Instrument Dry Tube Inspection / Replacement
- GE has informed BECo of corrosion in SIL 409 the instrument dry tube. BECo performed an inspection during the last outage and plan to inspect and replace, as necessary, during RF0 #7.
BWR In-vessel Instrument Line Inspection BECo is planning to perform inspection of SIL 420 in-vessel instrument line in accordance with SIL 420.
BWR Shroud Head Bolt Inspection BEco is planning to perform inspection SIL 433 of shroud head bolt recommended in the SIL 433.
l l
l i
i
, (Attachment 3) Page 1 of 1 Progress Since June 1985 _rdate Schedule A Status ATWS: ATWS Schedule Implementation Standby Liquid Control System modification schedule was submitted.to the NRC via BECo Ltr.
- 85-178 dated October 7, 1985.
Equipment Qualification Complete Schedule B Status Hydrogen Water Chemistry - Conduct Phase I Complete -
Feasibility Studies Testing will be performed during Cycle 6.
Regulatory Guide 1.97 BECo is preparing a response to the NRC Request for Additional Information (NRC Ltr.
dated December 12, 1985) on a schedule to be negotiated with NRC ProjectManager.
Schedule C Item Status New refuel bridge Is ready for shop acceptance test.
(Attachment 4) page 1 of i Summary o' Changes and Reasons Associated wi:9 Requiatory Requirements Schedule A Environmental Qualification ihe implementation date was csanged from March 31
, 1985 to November 30. 1985 as per NRC Letter dated
- 85-020 dated' January 29, 1985.
March 28, 1985 approving BECo's Extension Letter compliance on November 30, 1935. BEco achieved Environmental Qualification Schedule B i
RIP The Radiation remain Improvement in the approval Program was essentially completed - a few procedures process.
I&E has an inspector following the completion of these procedures as well as the continued implementation of RIP actions that are identified in the rertew.
One significant hardware modification which resulted from the review is the modification of the HP access comtrol area. This modification has been included in Schedule B and is targeted for completion prior to the start of RF0 #7.
Hydrogen Water Chemistry BECo Station.is developing a test system siellar to the one installed at the Dresden SPOS/ Computer Replacement The original date of December 1966 for completion of SPOS modification has been extended to March 31, 1987 in consideration of anticipated installation lead time.
9907
, Regulatory Affairs and Programs Correspondence Control Sheet OutgoingNRCLetter2.86.034 Distribution:
J. E. Howard A. L. Orsen J. D. Keyes Licensing File NRC Inspector (PNPS)
. W. S. Stowe Nuclear Oversight Tean Yankee Atomic PNPS Document Control NUS Corporation J. M. Fulton W. Wendland (ANI)
H. F. Brannan Peter Walsh D. Bryant W. G. Lobo .
K. Roberts C. H. Minot )
W. F. Hickey J. Nicholson l
Subject:
{
Long Term Plan Semlannual Report of January 1986 Issued: _ 3/24/86 1
{
Summary: i Schedule A, S and C ltems.This semiannual report of January 1 Keywords: LTP JAN 86 REPORT '
RA&P
Contact:
W. C. Lcbc Clas si fica tion: C Assignments / Commitments update report for July 1936 submittal to NRC. Provide input o Responsibility: K. P. Roberts Plan Due RA&P: NA Date Due RA&P: 6/18/86 Date Due NRC: 7/18/86
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SNM. ECT: GENERIC LETTER 83-28. ITEMS 3.1.3 AND 3.2.3 (TAC 5 .
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. 8tt%:[ Pilgrim Nuclear Power Station {
p y In Generic Letter 83-28, Items 3.1.3 and 3.2.3, the staff reques you review the existino Technical Specifications for P11erim Stati th determine whether any post-maintenance testing requirements might depra rather than enhance safety- tv Your responses dated November 7, 1983 and June 28, 1985, stated that'nd -
post-maintenance testino requirements had been identified in the Pilorim Technical Specifications that would degrade safety of the reactor trip system or other safety-related components. The staff, with the assistance of a contractor, has reviewed your responses and has found them to be
' acceptable.
This concludes the staff's action on Items 3.1.3 and 3.2.3 relative to Pilgrim Station.
Sincerely,
. . . P i W
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i John A. Zwolinski, Director Y ' M'1 " RWR Pro.iect Directorate #1 e,, $" , -
Division of BWR Licensino cc: See next page 1
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...A Pr. William D. Harrington Boston Edison Company Pilgrim Nuclear Power Station cc:
Mr. Charles J. Mathis, Station Mgr.
Boston Edison Company P.FD #1, Rocky Hill Road Plymouth, Massachusetts 02360 Resident inspector's Office -
U. S. Nuclear Regulatory Commission Post Office Box 867 Plymouth, Massachusetts 02360 Chairman, Board of Selectmen 11 Lincoln Oreet Plymouth, Massachusetts 02360 Office of the Commissioner Massachusetts Department of Environmental Quality Engineering One Winter Street Boston, Massachusetts 02108 Office of the Attorney General 1 Ashburton Place 19th Floor Boston, Massachusetts 02108 p Mr. Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street Boston, Massachusetts 02111 Regional Administrator, Region 1 U. S. Nuclear Regulatory Commission l 631 Park Avenue King of Prussia, Pennsv1vania 19406 Mr. James D. Keyes Boston Edison Company 25 Braintree Hill Office Park Braintree, Massachusetts 0?l84 I
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- - - - - \
. e hh July 31,1986 Docket No. 50-293 Mr. James M. Lykn -
Chief Operating Officer Roston Edison Company 800 Boylston Street Boston, Massachusetts 02199
Dear Mr. Lydon:
SUBJECT:
Generic letter 83-28. Item 1.2 Post-Trip Review (TAC #53619) 4 i
Re: Pilgrim Nuclear Power Station By letter dated Ilovember 7,1983, the Boston Edison Cowpany responded to our Generic letter 83-28 with reoard to required actions based upon the generic f implications of the Salem ATWS events. We have completed our review of your j response concerning Item 1.2 and conclude that the post-trip review data and 3 information capabilities are acceptable.
A copy of our Sa'ety Evaluation relative to this matter is enclosed. l Sincerely, QEIGINAL SIWLD BY John A. Zwolinski, Director BWR Project Directorate #1 Division of BWR Licensing
Enclosure:
As stated DISTRIBUTION Docket File NRC PDR L"eal J PDR BWD1 Rdg.
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EJordan GHolahan BGrimes JPartlow NThompson j JZwolinski
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o Mr. James M. Lydon Pilgrim Nuclear _ Power Station Boston Edison Company CC' Mr. Alfred E. Pedersen, Station Manaaer Boston Edison Company RFD #1, Rocky Hill Road Plymouth, Massachusetts 02360 Resident Inspector's Office U. S. Nuclear Regulatory Commission Post Office Box 867 Plymouth, Massachusetts 02360 Chairman, Board of Selectmen 11 Lincoln Street Plymouth, Massachusetts 02360 l Office of the Commissioner Massachusetts Department of Environmental Quality Engineering i One Winter Street l Boston, Massachusetts 02108 Office of the Attorney General 1 Ashburton Place 19th Floor Boston, Massachusetts 02108 l- Mr. Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street, 2nd Floor Boston, Massachusetts 02111 Regional Administrator, Reaion I U. S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Mr. James D. Keyes Boston Edison Company 25 Braintree Hill Office Park Braintree, Massachusetts 02184 l
l 1
- . . -- _ _9
/ o g UNITED STATES 8 e NUCLLAR REGULATORY COMMISSION
- r. WASHINGTON, D. C. 20555 g * , e * ,l SAFETY EVALUATION RY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATIVE TO GENERIC LETTER 83-28, ITEM 1.9 - POST-TRIP REVIEW BOSTON EDISON COMPANY PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293
- 1. INTRODUCTION ,
On February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant (SNPP) failed to open upon an ;.utomatic reactor trip signal from the reactor protection system. This incident occurred during the plant start-up and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal. The failure of the circuit breakers has been determined to be related to the sticking of the undervoltaoe trip attachment. On February 22, 1983, during the start-up of the SNPP, Unit 1, an automatic trip signal occurred as the result of steam generator low-low level. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip. Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (EDO) directed the staff to investigate and report on the generic implications of these occurrences. The results of the staff's inquiry into these incidents are reported in NUREG-1000. " Generic Implications of ATWS Events at the Salem Nuclear Power Plant." As a result of this investigation, the Commission .
requested (by Generic letter 83-28 dated July 8,1983) all licensees of !
operating reactors, applicants for an operating license, and holders of construction pennits to respond to certain generic concerns. These concerns are categorized into four areas (1) Post-Trio Review. (2) Equipment Classifica-tion and Vendor Interface, (3) Post-Maintenance Testino, and (4) Reactor Trip System Reliability Improvements.
The first action item, Post-Trip Review, consists of Action Item 1.1, " Program Description and Procedure" and Action Item 1.2, " Data and Information Capability." This safety evaluation addresses Action Item 1.2 only.
- 2. REVIEW GUIDELINES The following review guidelines were developed after initial evaluation of the various utility responses to Item 1.2 of Generic Letter 83-28 and incorporate the best features of these submittals. As such, these review guidelines, in effect, represent a " good practices" approcch to post-trip review.
A. The eouipment that provides the digital sequence of events (SOE) record and the analog time history records of an unscheduled shutdown should provide a reliable source of the necessary information to be used in the post-trip review. Each plant variable which is necessary to determine the cause and progression of the events followina a cost trio should be Q (9 0WI M~C m
. 2 monitored by at least one recorder (such as a sequence-of-eve".s recorder or a plant process computeti for digital parameters; and stri: charts, a plant process computer or analog recorder for analog (time history) I variables. Perfonnance characteristics guidelines for SOE and time history recorders are as follows:
o Each sequence of events recorder should be capable of detecting and recording the sequence of events with a sufficient time discrimins-tion capability to ensure that the time responses associated with !
each monitored safety-related system can be ascertained, and tha! a determination can be made as to whether the time response is within acceptable limits based on the accident analyses in Chapter 14 of the plant FSAR. The recomended guidelines for the SOE time discrimination is approximately 100 milliseconds. If curremt SOE recorders do not have this time discrimination capability the licensee should show that the current time discrimination capability is sufficient for an adequate reconstruction of the course of the reactor trip and post-trip events. As a minimum this sheuld include the ability to adequately reconstruct the transient and accident scenarios presented in Chapter 14 of the FSAR.
o Each analog time history data recorder should have a sample interval small enough so that the incident can be accurately reconstructed '
following a reactor trip. As a minimum, the licensee sht:uld be able to reconstruct the course of the transient and accident sectuences evaluated in the accident analyses in Chapter 14 of the FSM. The recommended guideline for the sampic interval is 10 secords. If the time history equipment does not meet this guideline, the licensee should show that the time history capsbility is sufficiert to accurately reconstruct the transient and accident sequences presented in Chapter 14 To support the post-trip analysis of the cause of the trip and the proper functioning of involved safety related equipment, each analog history data recorder should be capable of updating and retaining information from approrfmately five minutes prior to the trip until at least ten minutes after the trip.
o All equipment used to record sequence of events and time history infonnation should be powered from a reliable and non-interruptible power source. The power source used need not be Class IE.
B. The sequence of events and time history recording equipment steuld monitor sufficient digital and analog parameters, respectively, to assure that the course of the reactor trip and post-trip events can be reconstructed. The parameters mcnitored should provide sufficiemt information to determine the root cause of an unscheduled shutdcwn, the progression of the reactor trip, and the response of the plant parameters and protection and safety systems to the unscheduled shutdown.
Specifically, all input parameters associated with reactor trips, safety injections and other safety-related systems, as 'all as output parameters sufficient to record the proper functioning of these systems, should be recorded for use in the post-trip review. The parameters deened j necessary, as a minimum, to perform a post-trip review that would '
determine if the plant remaired within its safety limit desior emvelope are presented in Table 1. They were selected on the basis of staff l
j
~
. 3 engineering judgment following a complete evaluation of utility submittals. If the licensee's SOE recorders and time history recorders do not monitor all of the parameters suggested in these tables, the licensee should show that the existing set of monitored parameters are sufficient to establish that the plant remained within the design envelope for the accident conditions analyzed in Chapter 14 of the FSAR.
C. The information gathered by the secuence of events and time history recorders should be stored in a manner that will allow for data retrieval and analysis. The data may be retained in either hardcopy (e.g.,
computer printout, strip chart record), or in an accessible memory (e.g.,
magnetic disc or tape). This information should be presented in a readable and meaningful format, taking into consideration good human factors practices such as those outlined in NUREG-0700.
D. Retention of data from all unscheduled shutdowns provides a valuable reference source for the determination of the acceptability of the plant vital parauter and equipment response to subsequent unscheduled shutdowns. Information gathered during the post-trip review is to be retained for the life of the plant for post-trip review comparisons of i subsequent events.
- 3. EVAL.UATION By letter dated November 7,1983, the Boston Edison Company proviosd infoma-tion regarding its post-trip review program data and infomation capabilities for Pilgrim Station. We have evaluated the licensee's submittal against the review guidelines described in Section 2. Deviations from the guidelines were discussed with representatives of the licensee by telephone on February 12 and ,
26, 1986. A brief description of the licensee's responses and the staff's I evaluation of the responses against each of the review guidelines follows: ,
A. The licensee has described the performance characteristics of the equip-ment used to record the sequence of events and time history data needed for post-trip review. Based on our review of the licensee's submittal i and the infomation provided by the licensee during the telephone conversa-tions, we find that the sequence of events recorder and time history recorder characteristics confom to the guidelines described in Section 2A, and are acceptable. '
B. The licensee has established and identified the parameters to be monitored and recorded for post-trip review. The parameters selected by i the licensee include all but one of those identified in Table 1. While diesel generator Status is not included as a parameter, there are indicators in the control room that provide this information.
Consequently, we find that the licensee's selection of parameters meets the intent of the guidelines described above in Section 2B and is, therefore, acceptable.
C. The licensee described the means for storage and retrieval of the ,
information gathered by the secuence of events and time history recorders, i and for the presentation of this infomation for post-trip review and I analysis. Based on our review, we find that this information will be presented in a readable and meaningful fomat, and that the storage, retrieval and presentation confom to the above guidelines of Section 2C, 1
4 D. During the February 26, 1986 telephone conversation, the licensee stated that the data and information used during post-trip reviews are being retained in an accessible manner for the life of the plant. Based on this information, we find that the licensee's program for data retention conforms to the above guidelines of Section 2D and is acceptable.
4 CONCLUSION Based on our review of the licensee's submittal and the telephone conversations with the licensee, we conclude that the licensee's post-trip review data and information capabilities for Pilgrim Station are acceptable.
Principal Reviewer: Joel J. Kramer Dated:
9
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TABLE 1 BWR PARAMETER LIST SOE Time History
]
Recorder Recorder Parameter / Signal J x Reactor Trip x Safety Iniection x Containment Isolation x Turbine Trip x Control Rod Position x (1) x Neutron Flux, Power x (1) Main Steam Radiation (2) Contair. ment (Dry Well) Radiation x (1) x Drywell Pressure (Containment Pressure)
(2) Suppression Pool Temperature x (IT x Primary System Pressure x (1) x Primary System level x MSIV Position x (1) Turbine Stop Valve / Control Valve Position
- Turbine Bypass Valve Position x Feedwater Flow L x Steam Flow (31 Recirculation: Flow, Pump Status x (1) Scram Discherge Level x (1) Condenser Vacuum
. g SOE Time History Recorder Recorder Parameter /Sioral x AC and DC System Status (Bus Voltaae)
(3) (4) Safety Injection; Flow. Pump / Valve Status x Diesel Generator Status (On/Off, Start /Stop)
(1) Trip parameters (2) Parameter may be recorded by either an SOE or time history recorder.
(3) Acceptable recorder options are: (a) system flow recorded on an SOE recorder, (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder.
(4) Includes recording of parameters for all applicable systems from the following: HPCI, LPCI, LPCS, IC, RCIC.
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August 12, 1986 Ql1 Docket No.: 50-293 DISTRIBUTION: $[N U i Docket Files NRC PDR NThcepson Mr. James M. Lydon Local PDR JZwolinski Chief Operating Officer PD#1 Rdg File Pteech Boston Edison Company RBernero CJamerson 800 Boylston Street OGC-BETH (info only) ACRS, 10 i Boston, Massachusetts 02199 EJordan DHavercamp, R:I BGrimes JChung, R:I
Dear _Mr. Lydon:
JPartlow GHolaban Pilgrim File
SUBJECT:
GENERIC LETTER,83-28, ITEMS 3.1.1, 3.1.2, 3.2.1, 3.2.2, 4.5.1 (TACNUMBERSf2948,53785,54095)
RE: Pilgrim Nuclear Power Station In Generic Letter 83-28, we recuested that you review the maintenance and testing procedures at Pilgrim Station for safety-related components in the reactor trip system (Items 3.1.1 and 3.1.2) and all other safety-related equipment (Items 3.2.1 and 3.2.2). We also revested that you review the provisions for on-line functional testing of the reactor trip system, including independent testing of the diverse trip features (Item 4.5.1).
The Boston Edison Company's responses relative to the above items have been reviewed and found acceptable. Enclosed is a copy of the staff's Safety Evaluation supporting this conclusion.
This concludes the staff's action relative to these items with respect to Pilgrim Station.
~
Sincerely, Original signed by:
John A. Zwolinski, Director BWR Project Directorate #1 Division of BWR Licensing
Enclosure:
As stated cc w/ enclosure:
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- .20
wr, James W. Lydon Ros .ca Edison ComM ny Pilgrim Fuclear Power Station ec:
Mr.11' red E. Pedersen, Station Manaaer Bostor. Edison Company DFD dl, Rockv Hill Road Plyrcuth, Massachusetts 07360 Resident Inspector's Office II. S. Nuclear Reculatory Commission Post Office Box 867 Pl.n uth, Massachusetts 02360 Chaiman, Board of Selectmen 11 (Jncoln Street Plymuth, Massachusetts 02360 Office of the Commissioner Massachusetts Department of Environmental Quality Engineering One hinter Street ,
Rosten, Massachusetts 02108 i Office of the Attorney General 1 Ashburton Place 19th cloor }
Rostca, Massachusetts 02108 Pr. D3 bert F. Pallisev, Director Radiation Control Procram Massachusetts Department of PuMic Health 150 Trenont Street, 2nd Floor Boster, Massachusetts 02111 Reciocal Administrator, Reafor I '
U. S. Nuclear Regulatory Commission ,
631 Pari: Avenue King c' Prussia, Pennsylvania 10406 Mr. James D. Keyes Boston Edison Company ,
25 Praintree Hill Office Park Braintree, Massachusetts 0210.4 i i
i o
Safety Evaluation 3 pending final recommendations by the BWR owner's grow: and a Nuclear Utility Task Action Committee (NUTAC), as well as c: e vendors. In a letter dated A:-il 25, 1986, the licensee clarified re responsibil-ities of the s':e personnel in determining the post ,c-k maintenance and operability tests.
Based on the a::ve, tne staff concluded that the licarsee has complied with ::sitions for actions 3.1.1 and 3.1.2 :# Generic Letter 83-28.
2.3 Actions 3. 2.1 a-c 3.2.2, Post-Maintenance Testing ( A~1 Dther Safety Related Comporects)
Position Licensees and applicants shall submit a report documenting the ex-tending of tes; and maintenance procedures and Techn':a1 Specifica-tions review to assure that post-maintenance operabi'f ty testing of all safety-related equipment is required to be condu::e: and that the testing demonstrates that the equipment is capable c' ;+rforming its safety functices Defore being returned to service.
Licensees and a::licants shall submit the results of :reir check of vendor and engineering recommendations (all other sa'aty-related components) to assure that any appropriate test guican:e is included 4 in the test an: maintenance procedures or the Technia7 Specifica- I tions, where re; sired.
Discussion In letters date: Nover.ber 7, 1983, and April 25, 195f, the licensee stated that the ost-maintenance testing of the safe:/- elated com-ponents was re;u'. red by Stati:1 Procedure 1.5.3, Mair:e .ance Requests.
The specific test and procedure / method to do the reqLired post-main-tenance testing are determined by both the maintenance engineer and operating supervisor. The complete review of the safety-related SS&Cs was completed and the revf ew results were validated by Operations, Maintenance, Engineering, and Quality Assurance. The licensee further indicated that a review of maintenance procedures was completed and a Procedure Update Program (PUP) was instituted as par: c# the performance improvement effort for maintenance procedures. The FUF, a one time effort, has been completed including the vendor manual validation, and subsequent additional procedural revisions will :e handled by existing organ'zational procedures. The licensee als: stated that surveillance frequencies in Technical Specifications #cr the safety-related systems were initially formulated using vend:- information and established probability techniques, and that veno:* or other recommendations would be updated and incorporated int: :recedures as operating experience and new information are develope:.
.c. - )
infety Evaluation /.
The licensee is further consee-ing an evaluation of relevant industry and station failure data to assess appropriate. actions concerning post-maintenance testing re:.'-ements, and stated that the results :f the BWR Owner's Group and GE's recommendations and results would be incorporated into the progrars.
Based on the above and the l':e see's commitment to incorporate future vendor recommendations 1 to a maintenance procedure, the sta'f concluded that the licensee *as complied with the NRC staff positiot for action 3.2.1 and 3.2.2 fc- Seneric Letter 83-28, 2.4 Action 4.5.1, Reactor Trip System Reliability (System Functional Testing)
Position On-line functional testing of t~e reactor trip. system, including in-dependent testing of the dive-se trip features, shall be performed or, all plants.
The diverse t-10 features to be tested include the breaker undervoltage and shur- rip features on Wes'tinghouse, B&W a*d CE plants; the circuitry usec fce power interruption witn the silictn controlled rectifiers on B&W :lants; and the scram pilot valve and backup scram valves (includir; all initiating circuitry) on GE plants.
Discussion In a letter dated November 7,1953, the licensee indicated that on-line functional testing of the reactor protection system, and inde-pendent testing of the diverse t-ip features were performed in accort-ance with appropriate Technical 5 specification requirements, which included initiating circuitry. T ogic checks, and the actuating device.
In a letter dated June 28, 1954 the licensee further stated that a specific surveillance test fc- t9e backup ' scram valves was not perfc med routinely. The licensee has explained, and the staff agrees, -
that the reactor trip system currently is not designed to permit periodic on-line functional testing of backup scram valves.
Justification for not making modifications to permit on-line testing has been reviewed separately by the staff under Action Item 4.5.2 of Generic Letter E3-26. The staff found that such modifications are not require:'.
Safety Evaluati:- 5 l
' Based :- the above, the staff concluded that ;re licensee has complitc with the NRC staf f position for Acti:n 4.5.1 of Generic Letter 53-28.
3.0 Conclusion The staff c:ccludes that the licensee has compliet with Actions 3.1.1, 3.1.2, 2.2.', 3.2.2, and 4.5.1 of Generic Letter E3-28.
Dated: August 12, 1986 Principal Contributor:
J. Chung, Divisi:n of Reactor Safety, Region I l
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f o UNITED STATES g
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NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20555
%, # January 17, 1986 Docket No. 50-293 bS Mr. William D. Harrington Senior Vice President, Nuclear Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199
Dear Mr. Harrington:
SUBJECT:
DRAFT TECHNICAL EVALUATION REPORT (TER) FOR SALEM ATWS ITEM 1.2 (GENERIC LETTER 83-28)
Re: Pilgrim Nuclear Power Station The staff has completed a preliminary review to assess the completeness and adequacy of licensee responses to Generic Letter 83-28 Item 1.2. Your response for the Pilgrim Station was found to be incomplete as indicated below.
1 The enclosed Draft Technical Evaluation Report (TER) provides a technical I evaluation representing the staff's initial judgment of the areas evaluated.
We would appreciate your cooperation in providing the additional infonnation that will permit us to complete our review. The information needed is indicated in the conclusion and on the attached pages of the TER. It would appear that,the needed information on your facility could be obtained by telephone cor.ference after your receipt of the praft TER. Your NRC project manager will be working with your licensing staff to arrange an acceptable time to conduct the necessary conference calls.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, l John . Zwolinski, Director BWR Project Directorate il Division of BWR Licensing
Enclosure:
Draft TER cc w/ enclosure:
See next page
@ -(r tl) : 6 % tf d lo ? _ } ff'
V-Mr. William D. Harrington Boston Edison Company Pilgrim Nuclear Power Station cc:
l Mr. Charles J. Mathis, Station Mgr.
Boston Edison Company RFD #1, Rocky Hill Road Plymouth, Massachusetts 02360 Resident Inspector's Office U. S. Nuclear Regulatory Commission Post Office Box 867 Plymouth, Massachusetts 02360 i
Mr. David F. Tarantino Chairman, Board of Selectman ,
11 Lincoln Street Plymouth, Massachusetts 02360 Office of the Commissioner Massachusetts Department of Environmental Quality Engineering i One Winter Street Boston, Massachusetts 02108 ,
Office of the Attorney General '
1 Ashburton Piace 19th Floor i Boston, Massachusetts 02108
, Mr. Robert N. Hallisey, Director i Radiation Control Program.
Massachusetts Department of Public Health ,
150 Tremont Street Boston, Massachusetts 02111 Regional Administrator, Region I ,
U. S. Nuclear Regulatory Commission i 631 Park Avenue King cf Prussia, Pennsylvania 19406 Mr. A. Victor Morisi Boston Edison Company 25 Braintree Hill Park Rockdale Street J Braintree, Massachusetts 02184 I
A t O *'4 9',:
UNITED STATES
!(p . $ NUCLE AR REGULATORY COMMISSION WASHINGTON. O C. 20555 /
\...../ August 5, 1987 Docket No.: 50-293 Boston Edison Company M/C Nuclear ATTN: Mr. Ralph E. Cird Senior Vice President - Nuclear s 800 Boylston Street Boston, Massachusetts 02199
Dear Mr. Bird:
The Conrnission has issued the enclosed Amendment No. 102 to Facility Operating License No. DRP-35 for the Pilgrim Nuclear Power Station. This amendment consists of changes to the Technical Specifications in response to your application dated May 29, 1987 as supplemented by a letter dated July 15,1987 containing minor explanatory details.
This amendment revises the Technical Specifications to reflect modifications to the Standby Liquid Control System in response to the Anticipated Transient Without Scram Rule,10CFR50.62(c)(4). The revisions are to Technical Specification 4.4.A and 4.4.C for surveillance requirements; Section 3.4.C for sodium pentaborate solution chemical characteristics; deletion of Figure 3.4.2; changes to the bases; a change to Table 6.9.1, changes to Figure 3.4.1 and several minor administrative changes.
A copy of our Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's Bi-Weekly Federal Register Notice. l l
Sincerely, I AN L Richard H. Wessman, Senior Project Manager Project Directorate I-3 Division of Reactor Projects 1/11
Enclosures:
- 1. Amendment No. 102 to DPR-35
- 2. Safety Evaluation cc w/ enclosures:
See next page l
9 g,3 i 1613b bf f r/zr
1 Docket No.: 50-293
~
Boston Edison Company M/C Nuclear ATTN: Mr. Ralph E. Bird Senior Vice President - Nuclear 800 Boylston Street ,
Boston, Massachusetts 02199 !
Dear Mr. Bird:
l The Commission has issued the enclosed Amendment No.102 to Facility Operating License No. DRP-35 for the Pilgrim Nuclear Powr Station. This amendment consists of changes to the Technical Specifications in response ,
i to your application dated May 29, 1987 as supplemented by a letter dated July 15, 1987 containing minor explanatory details.
This amendment revises the Technical Specifications to reflect modifications to the Standby Liquid Control System in response to the Anticipated Transient Without Scram Rule.10CFR50.62(c)(4). The revisions are to Technical Specification 4.4.A and 4.4.C for surveillance requirements; Section 3.4.C for sodium pentaborate sclution chemical characteristics; deletion of Figure 3.4.2; changes to the bases; a change to Table 6.9.1, changes to Figure 3.4.1 and several minor adeninistrative changes. ;
A copy of our Safety Evaluation is also enclosed. The Nctice of Issuance will I' be included in the Conrnission's Bi-Weekly Federal Register Notice.
I Sincerely, j
/5 Richard R, Wessman, i Senior Project Manager Project Directorate I-3 i A
Division of Reactor Projects I/II l
Enclosures:
- 1. Amendment No.102 to DPR-35
- 2. Safety Evaluation cc w/ enclosures:
See next page S M OFC :PDI.3 ./ pn i. i r . nc e -
J
Mr. Ralph G. Bird Boston Edison Company. Pilgrim Nuclear Power Station l
l cc:
Mr. K. P. Roberts, Nuclear Operations Boston Edison Company Nuclear Pilgrim Nuclear Power Station ATTN: Mr. Ralph G. Bird i Boston Edison Company Senior Vice Pr?sident - Nuclear '
RFD #1, Rocky Hill Road 800 Boylston Street Plymouth, Massachusetts 0?360 Boston, Massachusetts 02199 Resident Inspector's Office Mr. Richard N. Swanson, Manager U. S. Nuclear Regulatory Comission Nuclear Engineering Departmer.t i Post Office Box 867 Boston Edison Company l Plymouth, Massachusetts 02360 ?5 Braintree Hill Park Braintree, Massachusetts 02184 Chairman, Board of Selectmen 11 Lincoln Street Ms. Elaine D. Robinson Plymouth, Massachusetts 02360 Nuclear Information Manager Pilgrim Nuclear Power Station Office of the Conrnissioner RFD #1, Rocky Hi.ll Road Massachusetts Department of Plymouth, Massachusetts 02~4C Environmental Quality Engineering l One Winter Street Boston, Massachusetts 02108 Office of the Attorney General 1 Ashburton Place 19th Floor Boston, Massachusetts 0?108 Mr. Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street, 2nd Floor Boston, Massachusetts 0?111 Regional Administrator, Region ! )
U. S. Nuclear Regulatory Comission '
631 Park Avenue King of Prussia, Pennsylvania 19406 I I
Mr. James D. Keyes i Regulatory Affairs and Programs Group '
Leader Boston Edison Company 75 Braintree Hill Park Braintree, Massachusetts 0?l84 l
l l
Mr. Ralph G. cird Boston Edison Company. Pilarim Nuclear Power Station cc:
Mr. K. P. Roberts, Nuclear Operations Boston Edison Company Nuclear Pilgrim Nuclear Power Station ATTN: Mr. Ralph G. Bird Boston Edison Company Senio Vice President - Nuc RFD #1, Rocky Hill Road 800 Boylstor. Street Plymouth, Massachusetts 0?360 Bostor., Massachusetts 02199 pesident Inspector's Office Mr. Richard 4. Swanson, Manager U. S. Nuclear Regulatory Cornission Nuclear Engineering Department Post Office Bor 867 Boston Edison Company Plymouth, Massachusetts 02360 ?5 Braintree Hill Park Braintree. Massachusetts 02184 Chairran, Board nf Selectmen 11 Lincoln Street Ms. Elaine D. Robinson Plymouth, Massachusetts 02360 Nuclear Information Manager Pilgrim Nuclear Power Station Office of the Commissioner RFD #1, Rocky Hill Road Massachusetts Department of Plymouth, Massachusetts 02360 Environmental Quality Engineering One Winter Street Poston, Massachusetts 02108 Office of the Attorney General 1 Ashburton Place 19th Floor Boston, Massachusetts 0?108 Mr. Robert M. Hallisey, Director Radiation Control Program Massachusetts Cepartment of Public Health 150 Tremont Street, 2nd Floor j Boston, Massachusetts 02111 1 Regional Administrator, Region i U. S. Nuclear Pegulatory Commission !
631 Park Avenue l King of Prussia, Pennsylvania 19406 Mr. James D. Keyes Regulatory Affairs and Programs Group l
Leader Boston Edison Company
?5 Braintree Hill Park Braintree, Massachusetts 0?l84
2
(?) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.102, are hereby incorporated in the 'ticense.
The licensee shall operate the facility in accordance with the -
Technical Specificatices.
- 3. This license amendment is e#fective immediately. [
FOR THE NUCLEAR REGULATORY C0*ISSION Victor Nerses, Acting Director Project Directorate I-3 Division of Reactor Projects I/II Atta chtren t:
Changes to the Technical Specifications Date of Issuance: August 5,1917 I
t '
ATTACHMENT TO LICENSE AME% " ENT NO. 102 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain ve-tical lines indicating the areas of. change. The corresponding overlea' pages are provided to maintain docucent completeness.
Remove Pages insert Pages 95 95 96 96 97 97 97a*
98 98 99 99**
100 100 101 101 102 102**
??? 225
- Denotes rew pace
- Pace f atentionally deleted i
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[IMITINQff)NDIT10NS F0D ODERAV10N
_ SURVEILLANCE RET IREMENTS 3,4 STANDBY LIOUID CONTROL SYSTEM 4.4 STAND 8Y L:1 :D CONTROL SYSTEM Aeolicability: i
, ADolicabiliS:
Applies to the operating status Applies to t5e surveillance of the Standby Liquid Control System. requirements of the Standby 1 Liquid Control System.
Obiective:
Obiective:
To assure the availability of a {
system with the capability to To verify the coerability of the Standby Liquid Control System.
shutdown the reactor and maintain the shutdown condition without the use of control rods.
Specification:
Soeci ficatigt:
A. Normal System Availability A. Normal System Availability
- 1. During periods when fuel is I in the reactor and prior to The operability of the Standby startup from a cold liquid Cortrol System shall be condition, the Standby Liquid verified by the performance of Control System shall be the following tests: I operable, except as specified 1.
in 3.4.B below. This system At least once per month need not be operable when the each pump loop shall be 4 reactor is in the Cold functionally tested by '
Shutdown Condition, all recirculating operable control rods are demineralized water to the fully inserted and test tank.
Specification 3.3.A is met. 2:
At least once during each i operating cycle:
- a. Check that the system relief valves trip full open at pressures less than 1800 psig, and restat on a falling pressure greater than 1275 psig.
- b. Marually initiate the system, except explosive valves.
Ruso boron solution through the recirculation path and back to the Standby Llosie Control Solstion Tank. Check f that each pump flow ,
rate meets or exceeds !
39 GFW against a '
system head of 1275 '
psi;.
Anendnent No. 102 95
. : w : ' : '. :- C M 3 F02 QPERATION SURVEILLANCE REQUIREMENTS -
3,: S MDBr L102:0 CONTROL SYSTEW 4.4 STANO3Y L10VID CONTRD; S : I-
- c. Manually initiate on of the Standby Liquic Control System loops and pump demineralize:
water into the reacte-vessel.
This test checks explosion of the charge associated witr the tested loop, proper operation of the valves, and pump operability. The replacement charges te be installed will be selected from the same i manufactured batch as the tested charge.
- d. Both systems, including both explosive valves, shall be tested in the course of two operating cycles.
B. Ooeration with Inocerable B. Surveillance with Incoeratte Comoonents: Comoonents:
- 1. From and after the date 1. When a component is found that a redundant to be inoperable, its component is made or redundant component shall found to be inoperable, be demonstrated to be Specification 3.4.A.1 operable immediately and shall be considered daily thereafter until the fulfilled and continued inoperable component is operation permitted repaired. !
provided that the component is returned to i an operable condition within seven days.
Amendment No. 102 96 i
_ _ _ _ _ _ _ _ _ _ l
.:u: :', 3 :?.:~ : *.3 04 0 ERATICN SURVEILLAN:E REO.:REMENTS 3.4 Siar6E' _::.:: CN 43 SfSTEM 4.4 S T A hCE'* LIO.ID CONTROL SYSTEM C. Soci;- :entacorate Solution C. Soc *wn Pentaborate Solution at a t're wren the Standby The f:llowing tests shall be L':;': :: : ci System is per'Ormed to verify the re:. e: :: ce o;erable the ava't arility of the Liquid fe'ic. : conditions snall be 'Cor:rol Solution:
met-
- 1. V:lume: Check at least
- 1. t*e 9et volume - ence per day.
- -:entratio" Of ne
- ': Con:r:1 Solution 2. Temperature: Check at-
'- ite Itaute control least once per day.
- i a snali be maintained a: e:ui e: in Figure 3. Concentra tion: Check at 3.*' least once per month.
Also check concentration-
- 2. ~re Temperature of the anytime water or boron is
- .ic contrcl solution added to the solution, or l 1*a be maintainec above the solution temperature
~
!!'; !f the solution is at or below 48'F.
ti :e a:L e falls to or
- e':. 48'F, the system 4. Ec-ichment: Check B'
.' te ' low tested to enrichment level by test
.e-*'y so Dath. arytime boron is added to tr.e solution and during 3 T*e enrichment of the ea:h refueling outage.
. :.10 control solution Enrichment analyses shall ce received within 30 days sa1 be maintaine at a E :sotcpe enrichment of test performance. If e.:eeding 54.5 atom net received within 30 pe tent. days, see Table 6.9.1 for reporting requirements.
D. Tnere are two operational i l considerations associated with the Standby Liquid Control sodium pentaborate l solution requirements. The first ccmsideration involves sodiu c+ntaborate concentration / volume requirement s . The second consideration involves B' isotopic enrichment. The relate Limiting Conditions for Operation are delineated below:
AmendmentNo. 102 97 W Me e
, 3 '. 4 STANDBY CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.4 STANDBf LIQUID CONTROL SYSTEM 4.4 STANDBY LIOUID CONTROL S*S~Ew D.1 If specification 3.4. A. E. or C.1 or C 2 cannot be met the reactor shall be placed in a Cold Shutdown Condition with all operable control rods fully inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.2 If the enrichment requirements of specification 3.4.C.3 are not met, a check I'
shall be made to ensure tnat sodium pentaborate solutien meets the original design criteria by comparing tne enrichment, concentration and volume to established criteria. If the sodium centaborate solution does not meet the original design criteria, the reactor sna11 te placed in a Cold Shutd:an Condition with all operable control rods fully inserted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.3 :# the sodium pentaborate solution meets the original design criteria, but the enrichment requirements of specification 3.4.C.3 are net net, bring the B isotopic enrichment to greater than 54.5 atom percent within seven days from the time of ;
receipt of the enrichment report. If after this time period the enrichment requirements of specification j 3.4.C.3 are still not met, submit a report to the NRC within seven days and advise them of plans to bring the solution up to a demonstratable 54.5 atom percent B isotopic enrichment. _
Arendment flo.102 g,
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,1 Page 99
('m:entionally deleted) ,
-ae crescsec 'S :-=aca! -
tre ATKS P.ule, IOC:^: -
ecuivalar: in con ct' .:-
- s i ng '.2 weight ce r
- c a- .'
Or0005ed 3DDr0 acne! J 'a3+'-
d;"rc.ad in "ae NO" +v: .
in g _ } .' ' -,ene, lc
-r
-ir'~u- : 3 n c e r, , e 9 = ~ - - -~
me**C* rent Of 52.; ?':- **-
l l
Ane 1 ent tio. 102 99 i
t
t
- SASES:
3.4 & 4.4 STANDBY LIQUID CON M OL SYSTEM A. The requirements for SLC capabil'ty to shut down the reactor are identified via the statton W.: lear S p. era ional M' 'Analysi s ( Appendi x G 'to"'thefsi#, Y(neffiqfe 3
l W. ' ff V n9"C
- ,;ffh. 1more than one operable control rcc is withdrawn, the basic
"~% . ' shutdown reactivity requirement for the core is satisfied and the Standby Liould Control systern is not required. Thus, the basie reactivity requWrrte Son stDexf@9t;uirie tha!,prioyya st4ndby 1d atti d ig determinant required. Theofdesign when ot th.e.herfi,,ve,',of sTind4nqtiftf ttl cpcntrol 5gstem ,/0h~t7df system is to provide t% unpattWttyadidsrbgi% cheer.eac1 tor from full power to a cchl d xen i assuming that none of' thlyVitl,,on,-fr,qe C rcfrain tonMff gutdoc .cond,ltion rbd's%rt tf( -,
inserted. To meet ttrh s2t@ertiust:athfrdita,ndb'A Lol.Wtd sontMic system is designed to in ect of b a concentration eauivaY tt7c.arT6C quantity "ifdW dt%a%oron yNn ,*ffr'tfIeatpr,opuces fYfSd reactor core. The 700 wicira'ui% heff.corfctsfitoHtiorilbrd the reactor core is requireq tq tirin reggpr front ful o r to a th ree pe rc en t a E $1.Iftti'tt' i s~.t,he,fa c6r It'i on'.cij.)dn s'ftfe .,,
ho t to c01 d r e a c t i V i ty/ TDff erepct. :MmordgXbtfd>rni J t? etC. The system wi11 inject this tgrpn solgign i,rgips_s ,t an ,4125 minutes. The maximumtinre Yhviremetr for in[e fng the Doron s01ution was seiecret itnrovereine th1EctteicafnNtb/ilttwe1s ,
insertioncausecbycgld,cg.off.erea.,ctgrfgl,1ogigtge,,gnon poison peak. g n g
~'a dW .er:9 s t u -a 11ri t , which is j to ge,.e,t, The S tandby 10CFR50.62 Liquid. fon.t,Tpl,,sys (Requirements for'htlTct'icrf
~
tg i.s ob a sK!,,s,q C re,q,ui re4, Tnt 1c1da,
'from , , ,
Transients Without-ScramdhTN5thfets' trtirrdsi(pte-Watarydooret huclear Power Plaftts),. . The- 5 tan @y, .Li,qtti(.Q,
~
o ntrol system must have the eauivalent' con't' rot cYpa'O ty ( fnf(c' tion rate) of 86 gpm
. at 13 percent by wt natural socium pentaborate for a 251" CfR50.62 diame ter reaeto.r requirements. This.pras.sgr,e correspon~ds .veTos.an SAL o
J4,qrdeg,tp equivalent corrtrcgti.s.fy y)h ,-'
capacity (injectiotr raw) 'of' 6fr gpou at F3 percertt' by wt.
natural sodium perttabora,te. Soluttcn .for ths .P.NP.S.,TAactor pressure vessel diameter of OlB". Thisequiv'arency requirement is fulfilled ty a 'contbitrrtiorr of' concentration, ff e enetthment
_ _ and flow rate af Jodiurn g concentration ud~ 54'.5't. enrTchhient penyborAte solug'iort.r.
of B' is~otope .A ginipujrt a.421 af a 39 GPM ic : : an pungr f1t>Wa't"F sot 13!$les :thet%0eS'%rivel'0CPA60. ifs 21entsieency
-'.: 1:c :ercerf.e@gM_;. cem days is allewad ic oring the bcron i er enmec 4-tc art %Ve'the c'o'neintWtttmt#1(Me cwWen mpy t5M0% vised to I x :11acce can n!teiyQ, r>il remenAh4,h'as f b~een acde,d to evaivate e the asolUtlon'tJnc.r,ejtseg' isq
, -" w e-m tap 4b'WtMo Mbe't5i tettchel(fda t'detrtagrUs%rtdtWrs (prileterttu eompiy l
,, ... - , r 3 :.,%h,eneg gtherJ'? enfc i ppegjgeggt ,1s3 gt 9eh y abeen aaraec
..' e3 a e e"* 5 EYpen etit%r'eWpurdWYdb t'Tttjccregdi'edtes $ttt ttWe MkitWloi e ' n .
..--.2.
e ,JAs cycle, t ,,,1 nr combtoiItadcrgi is suff'icient t
maintaint,h,, cump th4 :t(Hi,14u r-1,qgj performance. agg[.{Lt;ygil'y )
^ The or I
- h. .
N '-
- "Fidt1 c'&'ttine 'tb ' fill 1 ect'eWsthls 3 i.fau!i($ncbrit,r d t oD.sytetem is during a refueling outage. Various components of the system are individually tested periodically, thus making rnore frequent testing of the entire system unnecessary.
/.mendmen No. 102 100
_ _ _ _ _a
EASES:
3.4 & 4,4 STANDBY LIOUD CONTROL SYSTEM (Cont'd) w l T' t ?% minimum limitatigbg.the relief valve settisg is intended
,,.J Nto prevent the loss of sodium pentaborate solution via the lif ting of a relief v41 ye at :op low a prt_ssyre. The upper limit on the relief vadea 'Nt d : 4hff*ffogegysteFitticVitn r t from overpressure. are acceptable since they aceouately B. Only one of the two shdBE)'Mhuh MtItAhSMsg loops is needed for operating the system. One inoperable pumping circuit does not immediately threaten the shutdoun capability, and reactor operation can continue while the circuit is being repaired. As surance- tteed $4Dergetid@ nggysfgpl MS&ngeffprghd ts intended function and that th; long term Lvera v411ab111ty of the system is notM%NIttqdta"!96 s YUr %pWbbMbf two 5y5 tem by an a11owatde tetsdomtwt199tR f.3Ffd)F70tjmeP6/s6W)ft third of the normal surveilla This method cetermines an ecutor47qeg ohceYrfre i vi uenc . hf 0MPMPg. mental Additional conservatA sintdrss, introduced by reducing the allowable out of service time to seven days, and by increased testing of the operable redundant component. ,
.- "h! quantity cf B' stored in the Standby Liquid Centrol System Storage Tank is sufficient to bring the concentration of B' in the r eactor to tne point vberF steureattore wi+1la :e. Start 40wn and i te provice a minimum. 25 p,e,rcen,t paf. gin, be,ycAd Jf! .a. mount needed l to shutdown the reactor to 'alTodor possitrie MeWe*tt mixing
~
j cf the thenical solut torM1rthe reactors wedep.F er: I Level indica: ion and 'a'liit'm= iirf68te' b.e'the~r!~th s'[i jt%n volume f has changec, onit*r rrWjhteindi'tateW possj trit $td.t'*Qrt . ,
conc entra t ion change ._ _ Teat. .),n.t,erv,a l s,,f,of, J, eve l por_i, tori ng have l been established th, considera' tion oY these fa'ctcri.
- Temperature and liquid levei~rlatts'for the systro are annunciated in the, control.roo.m ,... _ ,,, , ..
The solution shal1 be kept at"least .10'F above-twe mawhnum ,
_ s atu r atj on t e mpef A.t.ur.e, t o, ya r,d, ,a,g a_i r :
Ninimum' solution ~ temperature Ts ,~48'f.ist, This isbp,r,on,,prn tMtation.
it*Fa'bove the !
.. ~ : '
~
'57twatitm ' temp'e'ratibiTor .the rurx,inur a+10wled 1terthum i
- c.or=-. -
e r, y pent (hnr#te, Eqnc2.f)ff.a tJg Of g 4.gHtrg . P,e,.rcgt n. .. gg
'Hus- a ci htWpmlnettr (concentration, pump flow rate, ard enrichment) '
is tested at an interval consistent with the potential for that :
parameter to vary anr, also to assure proper equit. ment l performance. Enrichment testing is required wher material is received and when c'iemical addition occurs since change cannot i occur by any process other than the addition of rew chemicals !
to the Standby Liquid Control solution tank. l l
Amendrent 30. 102 101
A v.4.
f'
'; rim vni:5 would provide for the use of e requirements of 10CFR50.62(c)(4) and selected by BECo and the associated TS ted, based on the considerations
-easonabie assurance that the health and
'ageret by operation in the proposed De cerdu:ted in compliance with the
.ance c' this amendment will not be
-curity or to the health and_ safety of the
.. 4, Page 132 (intentional. celeted) l i
l Arenament No.102 - - -
. 102 l
TABLE 6.9.1 s .n v ',
, .3 y
r, Reference Subaittal Cate a4,2 , d tainment 4.7.C.c Upon completion of te' h.
- esting (1) _, .. ,,ga.cjte,st,(2).. ,
l
- b. (Deleted) -
- o:5.
- c. (Deleted)
- ;ELI-3*.'M ~~ ' A * -i;a*ec a e d'.! :v e , ; X7:5. ,5?'a , 'ette-
- d. Gross Gaseous Release 4.1.4. Ten days af ter the 0.05 C1/sec for 48 release occurs Hours
.: -at - c 4- ' ;,199 with NPC
- e. Standby Liquid Control 3.4.Q.3 , Fourtyn solution enrichment out
- !" SM' *""ffteigto(r f if'donha,ys, after of specification complying enrichment report or lack of receipt of such a report witnir the required thirty days, if enrichment compliance cannot be achieved wit *:in seven days.
NOTES: 1. Each integrated leak rate test of the seconcary containment shall be the subject of a sumary technical report. This report shall include data on the wind speed, wind direction, outside and inside temperatures during the test, concurrent reactor building pressure, and emergency ventilation flow rate. The report shall also include analyses and interpretations of those data which demonstrate compliance with the specified leak rate limits.
- 2. The report shall be submitted within 90 days after completion of
.., each test. Test periods shall be based on the comercial service data as the starting point.
.w
~ f. . . , .
Amendment l1o . 102 225
f&~
- - s. d 4
- . :. - .; hyp Ay
[ ., umTro sTATas JCLEAR REGULATORY COMMISSION W" ' . .
(
{ wAsmcTom. o. c. 2 cess -
)
s -. I N s i BY THE OFFICE OF NUCLEAR REACTOR REGULATION A At14G TO STANDBY LIQUID CONTROL SYSTEM
.&E' .:%2d.O .. BOSTON EDISON COMPANY
'3
- C;i. .
. PILGRIM NUCLEAR POWER STATION
.f.;Pi~
'%. DOCKET NO. 50-293 N'~'.ff LO INTRODUCTION
' 'N ,
By letters dated May 29 and July 15, 1987, .
Company (BEco), to U.S. Nuclear Regulatory Cossnission, BECo pr6pos 1 the Technical Specifications (TS) for Pilgrim.
R.G. Bird in the Standby Liquid Contml System (SLCS) in order to meet the d the Anticipated Transient Without Scram (ATWS) Rule,10CFR50.62 The changes' address -_
thoron, enriched in the isotope B-10, in the sodium pentaborate' solu (c)(4). The proposed changes are to TS Sections 3.4. A. 3.4.C. 3.4.D,Y.'4~.'A.".,
- i 4.4.C, Figures 3.4-1 and 2, Bases 3.4, 4.4 and Table 6.9.1 all associated with SLCS.
The changes include achinistrative adjustments to page numbers' and section mumbers caused by the technical changes. In addition, in the revision of Table 6.9.1 the requirement for a 5 year Inservice Inspection (ISI) report was deleted since the report has already been submitted.
2.0 EVALUATION ,
l The) ' hanges for Pilgrim are intended to meet the requirteents of the" .
RSO.62(c)(4). The ATWS Rule requires that the SLCS be equivalen tal' capacity to a system with an 86 gpm injection rate,
.., u ,.v, wsing'13 weight percent unenriched sodium pentaborate solution. Of the several proposed approaches presented in the General Electric report, Reference 1, and approved in the NRC evaluation, Reference 2. BECo has chosen to use enriched (in B-10) boron. Using the calculation methods of Reference 1 results in a urinimum concentration of 8.42 weight percent sodium pentaborate when using an enrichment of 54.5 atom percent B-10, an injection of 39 gpm and a water sess of
- 8 7<VtJLhf 9-L - Qe.
j,
- .y .
y-
~
d ' '
.m .
~
_ _ / 2
,. w < -
., k ssel). The new limits are reflected in the revised
. d, n , Ngure 3.4.1.
.x
~
The tempera ure/ concentration requirements of existing Figure 3.4.2 are no longer required because the curve extends down only to 9.4% sodium pentaborate concentration and is based upon naturally enriched sodium pentaborate. The proposed revised concentration limits, in proposed TS Figure 3.4-1 allow a maximum concentration of 9.22% enriched sodium pentaborate. At*9.22% enriched sodium pentaborate concentration, the temperature required to preclude sodium pentaborate precipitation (with a 10*F margin) is 48'F. The controlled ~, 4 vp r
building temperatures provide assurance that it will be difficult for, SLC solution to approach this limit, and system alanns provide operator ' ,
notification of such a potential event. Because of the 10*F rurgin y^ y ,
potential sodium pantaborate precipitation at monitored concentration ( 1, the 48'F temperature limit provides equivalent protection to that conside j in the original safety evaluation. The 48'F temperature limit, whichk ,a..-
m included in the proposed TS, preempts the previous temperature - concintritfo P
..l[
curve provided in Figure 3.4.2. Accordingly, the staff finds the proposed TS Section 3.4.C.? and deletion of Figure 3.4.2 to be acceptable.
Having selecteti the enriched boron option of compliance with the ATWS Rule, BECo, following an approved approach, has elected to have the sodium pentaborate fonulated at the chemical vendor's facility. The boron enrichment test will therefore be done prior to the acceptance for use on the !
site. The boron enrichment test also will be done anytime boron is added to the solutiog,end each refueling outage. If the enrichment level is less than 53.5 atom hk, a . period of seven days is allowed to bring the boron enricivaskt i con iance. If at the end of the seven day period, compliance can. net be assured, the licensee is required to submit a report, within seven days, to the NRC advising the NRC of the licensee's plan to comply with the A1WS Rule. These are all acceptable procedures. They have been agreed upon as elements of an appropriate approach for compliance with the ATWS Rule in discussions between the staff and industry (BWR Owners Group ATWS Comnittee, Ref. 3). The proposed changes in T/S Sections 4.4.C.4 and 3.4.0. to l
.w r., . y 7.
$&f0
~
l* ;
.N 4
. z... $$ m
' "s x 3 ,a . M
- p.m
[ $$r, -
,' acceptable.
[ '[ \ 1c ~ P La p Mffication 3.4 and 4.4 were revised to reflect the M changes. The revised bases are acceptable since they huately e$1afn'the bases for the proposed requirement.s in the Technica( Specifications. ">- % 9
>: n The administrative changes, including the deletion of the requiht for the 5 year ISI report which has already been submitted, meet t(t elfhtbility '
- f. ,
{ criteria for categorical exclusion set forth in 10CFR51.22(c)(1d)$ I
. o \
to 10CFRSI.22(b) no additional environmental impact statement of envi I Y
~
assessment need be prepared for those items. ;
]
+ g, _ a t
2f ' ,
3.0 ENVIRONMENTAL CONSIDERATION
S ' '
. W.
This amendment involves a change in the installation or use of a facility ' , i component located within the restricted area as defined in 10 CFR Part 1
- 20. The staff had detennined that the amendment involves no significant '
increase in the amounts, and no sigr.ificant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational readiation exposure. The Commission has previously published a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 951.22(c)(9). Pursuant to 10 CFR 651.22(b), no environmental impact statement ohkvironmental assessment need be prepared in connection with the issuance of the amendment.
i f6ijf}} ~f Vi _ , g5 yQ:>p ene. y .- .y : 3 y' QQaskyp.w + ~ f g: 4gg
. Mt%# .s % "
cg;f lin44
'%@W.: i f m
' ' d
-} ' t
' 1 [ig-d.YRS 'p-
.~
n.ug, .- s for Pilgrim which would provide ffr the use of _
.4 . . w in the SLCS to meet the requirements of 10CFR50.52(c)(4) and
[,$ h,. kive changes. TheapproachselectedbyBEcoandthedssociatedTS i
.'.k.
".c ,h.V . p:regeleptdle.f N f The staff has concluded, based on the consider (tions u s .
,ji Ms9.ussed c"dm p .s shove, that: (1) there is reasonable assurance that. t$.
safety' of the public will not be endangered by operation proposed in the',J health
~ *ssener, and (2) such activities vill be conducted in compliance)~with the.
' 7 Consission's regulations and the issuance of this amendment wilf not' ' # ".
inimical to the common defense and security or to the health anhsaf j og-public. - a
~
s .?v
;W l' ~$
Principal contributors: G. Thomas, O. Gormley
- 4 3
Date: AUG 5 1967 > - 'a~,
. '*,$jk ,
,Bn.
+
4 REFERENCES !
- 1. " Anticipated Transients Without Scram: Response to NRC ATWS Rule, 10CFR50.62", NEDE-31096-P, December 1985. i j
- 2. " Safety Evaluation of Topical Report (NEDE-31096-P) ' Anticipated Transients Without Scras: Response to A1WS Rule,10CFR50.62'", letter from G. Lainas (NRC) October 21 1986.
- 3. Minutes to BWR Owner's Group informal meeting on April 1,1987 with NRC to discuss ATWS Technical Specification Bases, Bethesda, MD, April 3, 1987 1
I I I 4 i f i i
)
J S
l C36/f SAIC-85/1524-14 1 1 l REVIFs OF LICENSEE AND APPLICANT RESPONSES l TO NRC GENERIC LETTER 83-28 { (Required Actions Based on Generic Implications of i Salem ATWS Events), Item 1.2 '
" POST-TRIP REVIEW: DATA AND INFORMATION CAPABILITIES" FOR PILGRIM STATION (50-293)
I l Technical Evaluation Report Prepared by 4 Science Applications International Corporation 1710 Goodridge Drive .i McLean, Virginia 22102
)
Prepared for - U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract No. NRC-03-82-096 , 4 i l- -+ ~ ~ *
$A(p
1 I l FOREWORD l This report contains the technical evaluation of the Pilgrim Station response to Generic Letter 83-28 (Required Actions Based on Generic Implica-tions of Salem ATWS Events), Item 1.2 " Post Trip Review: Data and Informa-tion Capabilities." For the purposes of this evaluation, the review criteria, presented in part 2 of this report, were divided into five separate categories. These are:
- 1. The parameters monitored by the sequence of events and the time history recorders,
- 2. The performance characteristics of the sequence of events recorders,
- 3. The performance characteristics of the time history recorders,
- 4. The data output format, and
- 5. The long-term data retention capability for post-trip review ma terial .
All available responses to Generic Letter 83-28 were evaluated. The plant for which this report is applicable was found to have adequately responded to, and met, categories 2 and 4. The report describes the specific methods used to determine the cate-gorization of tne responses to Generic Letter 83-28. Since this evaluation report was intended to apply to more than one nuclear power plant specifics 1 regarding how each plant met (or failed to meet) the review criteria are not presented. Instead, the evaluation presents a categorization of the responses according to which categories of review criteria are satisfied and which are not. The evaluations are based on specific criteria (Section 2) derived from the requirements as stated in the generic letter, i l
TABLE OF CONTENTS Section Page l Introduction. . . . . . . . . . . . . . . . . . . . . . . . . I j
- 1. Background. . . . . . . . . . . . . . . . . . . . . . . . . . 2
- 2. Review Criteria . . . . . . . . . . . . . . . . . . . . . . . 3
- 3. Evaluation. . . . . . . . . . ................ 8 ,
i
- 4. Conclusion. . . . . . . . . . . . . . . . . . . . . . . . . . 9
- 5. Re fe re n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . 10 C. SupM'rzub- Oo0vhevi PoA. MLeccM - - -
- II i
4 1 l l l l
1 INTRODUCTION i SAIC has reviewed the utility's response to Generic Letter 83-28, item ! 1.2. " Post-Trip Review: Data and Information Capability." The response (see references) contained sufficient information to determine that the data and information capabilities at these plants are acceptable in the following areas. e The sequence-of-events recorder (s) performance charac-teristics. . j e The output f,ormat of the recorded data. However, the data and information capabilities, as described in the submittal, either fail. to meet the review criteria or provide insufficier.t information to allow determination of the adequacy of the data and informatic t capabilities in the following areas.
~
e The parameters monitored by both the sequence-of-events and time history recorders. e The time history recorder (s) performance characteris- ' tics. e The long-term data retention, recore keeping, capa-bility. j l 1 1
)
( L___-_---______. l
l l
- 1. Background On February 25, 1984, both of the scram circuit breakers at Unit 1 of the Salem Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident occurred during the plant startup and the reactor was tripped manually by the operator about 30 seconds after the initiation of the automatic trip signal The failure of the circuit breakers has :een determined to be related to the sticking of the under voltage trip atta:hzent. Prior to this incideat; on February 22, 1983; at Unit 1 of the Salen Nuclear Power Plant an automatic trip signal was generated based on stear generator low-low level during plar.t startup.
In this case the reactor was tripped manually by the operator almcst coinci-dentally with the automatic trip. At that time, because the utility did not have a requirement for the systematic evaluation of the reacter trip, no investigation was perfortred to determine whether the reactor was tripped automatically as expected or manually. The utilities' written procedures required only that the cause of the trip be determined and identified the responsible personne' that could autnorize a restart if the ca;se of the trip is known. Fol;os ' g t9e second trip which clearly indicated the problem with the trip breakers, the question was raised on whether the circuit breakers had functi:ned properly during the earlier incident. The most useful source of information in this case, namely the sequence of events printout which would hav? indicated whether the reactor was tripped automatically or manually daring the February 22 incident, was nc: retained after the incident. Thus, r.: judgment on the proper functioning cf the trip system during the earlier ircitant could be made. Following these incidents; on February 28, 1983; the NRC Executive Director for Operations (EDC), directed the staff to investigate and report on the generic implications of tnese occurrences at Unit 1 of the Salem Nuclear Power Plant. The res;1ts of the staf f's inquiry into t*e generic implications of the Salem Unit incidents is reported in NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant." Based on the results of this study, a set of required actions were developed and included in Generic Letter 83-28 which was issued on July 8,1983 and sent to all licensees of operating reactors, applicants for operating license, and construction permit holders. 'ne required actions in this generic letter ; consist of four categories. 79ese are: (1) Post-Trip Review, (2) Equipment l 4 2
Classification and Vender Interface, (3) Post Maintenance Testing, and (4) Reactor Trip System Reliability Improvements. The first required action of the generic letter, Post-Trip Review, is the subject of this TER and consists of acticn item 1.1 " Program Description and Procedure" and action item 1.2 " Data and :nformation Capability." In the next section the review criteria used to assess the adequacy of the utilities' responses to the requirements of action item 1.2 will be discussed.
.2. Review Criteria The intent of the Post Trip Review requirements of Generic Letter 83-28 is to ensure that the licensee has adeqaate procedures and data and information sources to understand the cause(s) and progression of a reactor trip. This understanding should go beyond a simple identification of the course of the event. It should include the ca;: ability to determine the root
-cause of the reactor trip and to determine whether safety limits have been exceeded and if so to what extent. Sufficier.t information about the reactor trip event should be available so that a decision on the acceptability of a reactor restart can be made.
Tne following are the review criteria developed for the requirements of Generic Letter 83-28, action item 1.2: T9e equipment that provides the digital sequence of events (50E) record and the analog time history records of an unscheduled shutdown should pro-vide a reliable source of the necessary information to be used in the post trip review. Each plant variable which is necessary to determine the causeis) and progression of the event (s) fcilowing a plant trip should be monitered by at least one recorder [such as a sequence-of-events recorder or a plant process computer for digital parameters; and strip charts, a plant process computer or analog recorder for ana'o; (time history) variables]. Each device used to record an analog or digital plant variable should be described in sufficient detail so that a determination can be made as to whether the following performance characteristics are met: l i 3 I i l 1
e Each sequence-of-events recorder should be capable of detecting and recording the sequence of events with a suf ficient time discrimination capability to ensure that the time responses asso-ciated with each monitored safety-related system can be ascer- ; tained, and that a determination can be made as to whether the l time response is within acceptable limits based on FSAR Chapter 15 Accident Analyses. The recommended guideline for the SOE time discrimination is approximately 100 msec. If current SOE l recorders do not have this time discrimination capability the licensee or applicant should show-that the current time discrimi-nation capability is sufficient for an adequate reconstruction of tt.e course of the reactor trip. As a minimum this should-include the ability to adequately reconstruct the accident scenarios pre-sented in Chapter 15 of the plant FSAR. e Each analog time history data recorder should have a sample inter-val small enough so. that the incident can be accurately reconstructed following a reactor trip. As a minimum, the licensee or applicant should be able to reconstruct the course of the accident sequences evaluated in the a:cident analysis of the plant FSAR (Chapter 15). The recommended ;Jideline for the sample interval is 10 sec. If .the time history equipment does not meet this guideline, the licensee or applicant should show that the current time history capability is sufficient to accurately recon- i struct the accident sequences presented in Chapter 15 of the FSAR. e To support the post trip analysis of the cause of the trip and the proper functioning of involved safety related equipment, each analog time history data recorder should be capable of updating and retaining information from approximately five minutes prior to the trip until at least ten minutes after the trip. e The information gathered by the sequence-of-events and time history data collectors should be stored in a manner that will allow for retrieval and analysis. The data may be retained in either hardcopy (computer printout, strip chart output, etc.) or in an accessible memory (magnetic disc or tape). This information should be presented in a readable and meaningful format, taking i i 4 I
into consideration good human factors practices (such as those outlined in NUREG-0700). e All equipment used to record sequence of' events and time history information should be powered from a reliable and non-interruptible power source. The power source used need not be safety related. The sequence of events and time history recording equipment should monitor sufficient digital and analog parameters, respectively, to assure that the course of the reactor trip can be reconstructed. The parameters monitored should provide sufficient information to determine the root cause of the reactor trip, the progression of the reactor trip, and the response of the plant parameters and systems to the reactor trip. Specifically, all input parameters associated with reactor trips, safety injections and other safety-related systems as well as output parameters sufficient to record the proper functioning of these systems should be recorded for use in the post trip review. The parameters deemed ne:essary, as a minimum, to perform a post-trip review (one that would determine if the plant remained within its design envelope) are presented on Tables 1.2-1 and 1.2-2. If the appli-cants' or licensees' SOE recorders and time history recorders do not monitor all of the parameters suggested in these tables the applicant or licensee should show that the existing set of m:nitored parameters are sufficient to establish that the plant remained withiq the design envelope for the appro-priate accident conditions; such as those analyzed in Chapter 15 of the plant Safety Analysis Report. Information gathered during the post trip review is required input for future post trip reviews. Data from all unscheduled shutdowns provides a valuable reference source for the determination of the acceptability of the plant vital parameter and equipment res;onse to future unscheduled shut-downs. It is therefore necessary that information gathered during all post j trip reviews be maintained in an accessible manner for the life of the l pl a nt . i l
)
5
l Table 1.2-1. PWR Parameter List SOE Time History Aecorder Recorder Parameter / Signal x Reactor Trip (1)x Safety Injection x Containment Isolatior. (1)x Turbine Trip x Control Rod Position (1) x x Neutron Flux, Power x x Containment Pressure (2) Containment Radiatior. x Containment Sump Level (1) x x Primary System Pressu-e (1) x x Primary System Temperature (1)x , Pressurizer Level (1)x Reactor Coolant Pump Status (1) x x Primary System Flow (3) Safety Inj.; Flow, Punp/ Valve Status x MSIV Position x x Steam Generator Pressare (1) x x Steam Generator Level (1) x x Feedwater Flow (1) x x Steam Flow (3) Auxiliary Feedwater System; Flow. Pump /Value Statas x AC and DC System Statas (Bus Voltage) x Diesel Generator Statas (Start /Stop, On/Off) ! x PORV Position (1): Trip parameters (2): Parameter maj te monitored by either an SOE or time history recorder. (3): Acceptable recorder options are: (a) system flow recorded on an SOE f recorder, (t) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder. 6 1 h I
Table 1.2-2. BWR Parameter List SOE Time History Recorder Recorder Parameter / Signal 1 x Reactor Trip Safety Injection : x x Containment Isolation x Turbine Trip x Control Rod Position ) Meutron Flux, Power I x (1) x x (1) Main Steam Radiation (2) Containment (Dry Well) Radiation x (1) x Drywell Pressure (Containment Pressure) (2) Suppression Pool Temperature x (1) x Primary System Pressure x (1) x Primary System level x MSIV Position x (1) Turbine Stop Valve / Control Valve Fosition x Turbine Bypass Valve Position x Feedwatcr Flow x Steam Flow (3) Recirculation; Flow Pump Status x (1) Scram Discharge Level x (1) Condenser Vacuum x At and DC System Status (Bus Voltage) (3)(4) Safety Injection; flow. Pump / Valve Status x Diesel Generator Status (On/Off, Start /Stop) (1): Trip parameters. (2): Parameter may be recorded by either an SOE or time history recorder. (3): Acceptable recorder options are: (a) system flow recorded on an SOE recorder, (b) system flow recorded on a time history recorder, or f l '(c) equipment status recorded on an SOE recorder. l (4): Includes recording of parameters for all applicable systems from the j follofing: HPCI, LPCI, LPCS, IC, RCIC.
)
i 7 l i _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . l
- 3. Evaluation _
The parameters identified in part 2 of this report as a part of the review criteria are those deemed necessary to perform an adequate post-trip - review. The recording of these parameters on equipment that meets the guidelines of the review criteria will result in a source of information , that can be used to determine the cause of the reactor trip and the plant m response to the trip, including the responses of important plant systems. The parameters identified in this submittal as being recorded by the sequence of events and time history recorders do not correspond to the parameters specified in part 2 of this report. The review criteria require that the equipment being used to record the i sequence of events and time history data required for a post-trip review meet certain performance characteristics. These characteristics are intended to ensure that, if the proper parameters are recorded, the record- - ing equipment will provide an adequate source of information for an effec-tive post-trip review. The information provided in this submittal does not , indicate that the time history equipment used would meet the intent of the performance criteria outlined in part 2 of this report. Information supplied in the submittal does indicate that the SOE equipment meets the , performance criteria specified in part 2 of this report. , The data and information recorded for use in the post-trip review should be output in a format that allows for ease of identification and use of the data to meet the review criterion that calls for information in a readable and meaningful format. The information contained in this submittal indicates tnat this criterion is met. The dats and information used during a post-trip review should be retained as part of the plant files. This information could prove useful '. during future post-trip reviews. Therefore, one criterion is that infor-matior, used furing a post-trip review be maintained in an accessible manner for the life of the plant. The information contained within this submittal does not indicate tdt this criterion will be met, b 8 ,
4 Conclusice l The information supplied in response to Generic Letter 83-28 indicates that the current post-trip review data and information capabilities are adequate in the following areas:
- 1. The recorded data is output in a readable and meaningful format.
- 2. The sequence of events recorders meet the minimum performance characteristics.
The information supplied in response to Generic Letter 83-28 does not indicate that the post-trip review data and information capabilities are adequate in the following areas:
- 1. Based upon the information contained in the submittal, all of the parameters specified in part 2 of this report that should be recorded for use in a post-trip review are not recorded.
- 2. Time history recorders, as described in the submittal, do not meet the minimum performance characteristics.
- 3. The data retention ;rc:edures, as described in the submittal, may not ensure that the information recorded for the post-trip review is maintained in an accessible manner for the life of the plant.
It is possible that the current data and information capabilities at this nuclear power plant are adequate to meet the intent of these review criteria, but were not completely described. Under these circumstar.ces, the { 1 licensee should provide an updated, more complete, description to show in i more detail the data and information capabilities at this nuclear power plant. If the information provided accurately represents all current data and information capabilities, then the licensee should show that the data and information capabilities meet the intent of the criteria in part 2 of 1 this report, or detail future modifications that would enable the licensee to meet the intent of the evaluation criteria. 9
REFERENCES NRC Generic Letter 83-28. " Letter to all licensees of operating reactors, applicants for operating license, and holders of construction permits regarding Required Actions Based on Generic Implications of Sales ATWS Events." July 8, 1983. NUREG-1000 Generic Implications of ATWS Events at the Salem Nuclear Power Plant, April 1983. Letter from W.D. Harrington, Boston Edison Company, to D.B. Vassallo, ' NRC, dated November 7,1983, Accession Number 8311090331 in response to Generic Letter 83-28 of July 8,1983, with attachment. e
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W ' :. ,- rs performance characteristics: Satisfactory g ..j_-:gy-,
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- . computer: 16.Sesec time discrimination W1 h,_ _. . ~ible i YMfc gg..smj[)l5g. et %% power supply Qil igf ! P<
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MvM, A W#.Ti.m,.p~ Alstory recorders performance characteristics: s f..
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Plant computer: 5 sec sample interval for the perio( fen 2
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pre-trip to 2.5 minutes post-trip ' c. Strip charts are also used. g .
- 4. Data output format: Satisfactory p; ..; , .
.r' ?..;x . ,
SOE: output includes time, event descriptor, and sensor ID. ! Time history: output includes time, parameter value, and sensor ID.
- 5. Data retention capability: Unsatisfactory l Data is retained but for an unspecified period.
*, T'.,
y Q.a H
I i Desirable BWR Parameters for Post-Trip Review (circled parameters are not recorded) SOE Time History Recorder Recorder Parameter / Signal 1 x Reactor Trip l
@ Safety Injection
@ Containment Isolation I x Turbine Trip x Control Rod Position j x (1) x Neutron Flux, Power '
x (1) Main Steam Radiation
@ Containment (Dry Well) Radiation x (1) x Drywell Pressure (Containment Pressure) l (2) Suppression Pool Temperature l
x (1) x Primary System Pressure x (1) x Primary System level - i
@ MSIV Position x (1) Turbine Stop Valve / Control Valve Position ;
@ Turbine Bypass Valve Position j x Feedwater Flow x Steam Flow (3) Recirculation; Flow, Pump Status x (1) Scram Discharge Level x (1) Condenser Vacuum
@ AC and DC System Status (Bus Voltage)
(3)(4) Safety Injection; flow. Pump / Valve Status
@ Diesel Generator Status (On/Off.
Start /Stop) (1): Trip parameters. (2): Parameter may be recorded by either an SOE or time history recorder. (3): Acceptable recorder options are: (a) system flow recorded on an SOE recorder, (b) system flow recorded on a time history recorder, or (c) equipment status recorded on an SOE recorder. (4): Includes recording of parameters for all applicable systems from the following: HPCI, LPCI, LPCS, IC, RCIC. lA
va;. ur a BOSTON EDISON E!!!! I l NUCLE; OPERATI0t:5 PRD:EDURE Vendor Manual Control Submi tted :.(f/ .L 3' ' f? , . _, _ 5 - .~ - [ Preparer
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- 1) Y N'frN) diept. Mgr, Da:a QA Approval: h /S w % j lo/l a:e QAM l QA Program related: hYes O 'e Approval: NEE 2 VPN0
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.W Vg&QA" 0l" A Ci.s
_VPN S j~ 02:e 10/19/85 Effective Date: b4df[9[ 9, g NOTI: This revision is an extensive re-write; therefore, revision notations in the margin are not used.
)
s Sage 2 of 10 WOP 84A4 Pevised 1 19/85 VENDOR MANUAL CONTROL 1.0 PURPOSE The purpose of this procedure is to establish a system for tre control of-Vendor Manuals associated with the installation, operation, ande maintenance activities of Pilgrim Nuclear Power Station. ~1f s procedure identifies and assigns responsibilities related to the Venoor Manual Control System. 2.0 SCOPE This procedure governs activities associated with the following areas of Vendor Manual control: 2.1 Required Instructions 2.2 Required Vendor Manuals 2.3 Vendor Manual Classification 2.4 Vendor Manual Authorization 2.5 vendor Manual Preparation 2.6 Vendor Manual Revision 2.7 Authorization to Own Vendor Manuals 2.8 1ssuance of Vendor Manuals 2.9 Maintenance of Vendor Manuals by Controlled Copy Holeers 2.10 Vendor Manual Indexes 2.11 Historical Files 3.0 APPLICABILITY This procedure is applicable to Vendor Manuals involved wiin the installation, operation and maintenance activities of Pilgrim Nuclear Power Station. m,.L , ,, ,,,o ^:... '. : . . . ^. . ,.......
4.0 REFERENCES
4.1 ANSI N45.2.9 - 1974: Requirements for Collection,, Storage and Maintenance of QA records for Nuclear Power Plants 4.2 ANSI Na5.2.ll - 1974: QA Requirements for the Desigt of Nuclear l Power Plants 4.3 ANSI N45.2.13 - 1976: QA Requirements for Control o' Procurement
-- of items and Services i
Page 3 of 10 NOP84A4 Revised 9/19/85 4.4 BE0AM Volume !!, Section 3, 5, 6 and 7
4.5 N0P83A7
Controlled Documents
4.6 NOP83ES
10CFR50.59 Safety Evaluations 5.0 DEFINITIONS 5.1 Vendor Manual - A document that may include drawings, pictures, instructions, as appropriate, provided by a Vendor or supplier, to { provide technical information pertaining to the installation, operation, and/or maintenance of a system, structure or component. 5.2 vendor Manual Control - The process utilized to assure the adequacy and availability of controlled vendor Manuals at Pilgrim Nuclear Power Station. 5.3 Safety Related Vendor Manual Revision - A revision to Class 1 controlled Vendor Manual resulting from a change to, addition of, or removal of, a safety related item, system or structure; or resulting from a change in the operation or maintenance of a safety related item, system or structure. 5.4 Validation - The process of annotating the generic Vendor Manual / Instruction (Revision) to reflect the actual configuration or specific equipment at PNPS. 5.5 Owner of the Vendor Manual - The Cognizant N00 Chief responsible i for the specific Vendor Manual, including authorization for issuance or change and for assignment of Controlled Copy Holders of same. Any overlap or duplicate responsibility will be resolved , through the NOM. 5.6 Controlled Copy Holder - The assigned holder of a controlled document (in thi! case, a Vendor Kanual) as identified by the Owner of the Vendor Manual. 5.7 Vendor Manual Change Request (VMCR) - A form used to administrative 1y authorize, approve and control changes to Vendor Manuals (Exhibit 1 to this procedure). 6.0 RESPONSIBILITIES 6.1 N11R$tleMen@t@fppM6 memMMWFP. 1 I 6.1.1 Ensures Completion of the VMCR (Exhibit 1). sr 6.1. 2 Validation of Vendor Manuals e
, i Page 4 of 10 NOP84A4 Revised 9/19/85 6.1.3 Reviewing appropriate Class i Vendor Manuals and submitting to the NOM f or approval 6.1.4 Approving appropriate Class 2 Vendor Manuals and submitting to the Nuclear Operations Support Department 6.2 The Nuclear Operations Manager's responsibilities include:
6.2.1 Approving Class 1 Vendor Manuals 6.2.2 Submitting approved manuals to the Nuclear Operations Support Department 6.3 The Nuclear Operations Support Department's responsibilities include: 6.3.1 Receipt and processing of Vendor Manual mark-ups and Vendor Manual Change Request forms > 6.3.2 Assignment of PNPS Vendor Manual Numbers and PNPS Revision Pumbers 6.3.3 Coordination of the activities required for the formal preparation of Vendor Manuals 6.3.4 Turnover of approved Vendor Manuals and VMCR's to RMG for issuance and distribution 6.4 The N include:
~' '
6.4.1
~~ ~ ' ' ' ' - " " " " " '"
6.4.2 MedDW9M8089F
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6.4.3 Immeset senesesMemurter 6.4.4 Maintenance of histo:ical files 6.4.5 Coordination of inventory
6.5 61993tyfesponsibilitiesinclude
6.5.1 Maintenance of their Vendor Manuals according to RMG transmittals and instructions 6.5.2 Acknowledgment on each transmittal and return to RMG in a timely manner 6.5.3 fMVtWte e d timely correction of identified discrepancies.
Page 5 of 10 NOP94A4 Revised 9/19/E5 7.0 REQUIREMENTS 7,1 Reauired Instructions 7.1.1 Vendor Manual Control System is organized, described ard administered by this N0P. 7.1.2 Departments shall implement their assigned responsibilities ; through their own procedures and/or work instructions, where applicable. 7.2 Required Vendor Manuals 7.2.1 Vendor' Manuait used:to perfonsW6 shall be. validated'anf cottN1TWW Contral Systeer 7.2.2 Vendor Manuals that provide information or instr'uctions concerning structures, systems or components should be controlled by the Vendor Manual Control System. 7.3 vfndor Manual Classification ._. 7.3.1 Vendor Manuals shall be c lassified by their Owners witt
- respect to plant operation and public safety.
7.3.2 The following classifications shall be applied to the Vendor Manuals in the control system: 7.3.2.1 - pertonmens 7.3.2.2 h I 5
%M l 59eC1ffth 7.4 Vendor Manual Authorization 7.4.1 Addition or deletion of controlled Vendor Manuals can be initiated by any Department utilizing the VMCR, and v
Page 6 of 10 N0P84A4 Revised 9/19/85 submitting the specific Vendor Manual direction to the apparent / designated owner. 7.4.2 The NOM (or designee) shall approve additions to or deletions f rom the Vener Manual Control System via
-signature on the completed VMCR.
7.5 Vendor Manual Preparation 7.5.1 New Vendor Manuals shall be prepared through the Owner of the Manual prior to addition to the Vendor Manual Control System. 7.5.2 The information requested in the VMCR must be provided by the owner as completely as possible for processing of l Vendor Manuals. Additionally, a list of effective pages and their current revision shall be provided. 7.5.3 8Meral*fortitfMder I . 7.6 Vendor Manual Revision 7.6.1 Class 1 Vendor Manuals 7.6.1.1 The proposed revision and the VMCR (including justification for revision) shall be reviewed by the Owner of the Vendor Manual (i.e. if a 4 maintenance / control revision, the Maintenance Group reviews), and approved by the NOM. 7.6.1.2 The cognizant Chief is responsible for deciding (documenting) if a Safety Evaluation is required and preparing / obtaining this Evaluation. 7.6.2 Class 2 Vendor Manuals 7.6.2.1 The proposed revisions shall be approved and authorized by the Cognizant Chief. 7.6.3 uppeammumv- , - , if applicable. Approved PDC's that are outstanding and not yet closed out will be annotated on the cover sheet of the controlled copies of each specific Vendor Manual affected.
Page 7 of 10
- NOP84A4 Revised 9/19/ E5 7.7 Authorization to Own Controlled Vendor Manuals 7.7.1 A list of Controlled Copy Holders shall be prepared by the Owners and maintained by RMG. Changes to these lists are authorized by the Owner of the Manual and approved on the VMCR. Such changes are forwarded to the RMG Leader via the YMCR.
7.7.2 One controlled copy'of eactnyonporgNaquelasWW495W tof the..PilgrJa-Document +Contrebtenteer 7.7.3 'OneYontrolled -topy efCMt9PMW d to' the Braintree .DorumentToMPtAdWWeesyand wil1 be available to be logged out by Braintree personnel for reference. 7.8 Issuance of Vendor Manuals 7.8.1 Controlled and uncontrolled issuance of Vendor Manuals shall be accomplished in accordance with Ref. 4.5. 7.8.2 Additional copies of each Manual shall be reproduced by PNPS DCC for issue to other parties when determined necessary. These copies are uncontrolled documents and do not have to N; maintained current. 7.9 Maintenance of Vendor Ranuals by Controlled Cooy Holders 7.9.1 Maintenance shall be accomplished in accordance with Ref. 4.5. 7.9.2 In the event of a lost or destroyed manual, the Controlled Copy Holder shall take the following action: 7.9.2.1 Notify DCC of the loss or destruction. OCC will issue a replacement cratrolled manual. 7.9.2.2 DCC will update the Vendor Manual index to reflect this replacement action. 7.9.i.3 If, at a later date a manual that was lost is ' found, it will be returned to DCC for disposition. 7.10 Vendor Manual Indices l 7.10.1 RMG shall prepare an index of controlled Vendor Manuals which will include the following information for each as l
.- applicable:
7.10.1.1 PNPS Vendor Manual Number
I I Page 8 of 10
- N0P84A4 Revised 9/19/85 7.10.1.2 Vendor Name 7.10.1.3 vendor Manual Title 7.10.1.4 Cross References (i.e. Vendor's Manual Number, Old BECo. P.O. Number) 7.10.1.5 PNPS Vendor Manual Revision Number 7.10.1.6 Vendor Manual Classification 7.10.1.7 Owner of the Vendor Manual 7.10.2 An index comprised of deleted Vendor Manuals shall be prepared and shall contain the following information:
7.10.2.1 PNPS Vendor Manual Number 7.10.2.2 PNPS Number of superseding Vendor Manual (if applicable) 7.10.3 A cross reference index, Vendor's Manual Number versus PNPS Vendor Manual Number shall be prepared. 7.10.4 Indices of controlled Vendor Manuals shall be maintained by RMG. They shall be updated upon receipt of a new Vendor Manual, a revision to an existing manual, or deletion of an existing manual, and shall then be issued to the following standard distribution: 7.10.4.1 CME 7.10.4.2 COE 7.10.4.3 Technical Section Head 7.10.4.4 DCC Plymouth 7.10.4.5 DCC Braintree 7.11 Historical Files 7.11.1 The record copy of all Vendor Manuals shall be maintained by RMG in accordance with Ref. 4.1. 7.11.2 Additionally, the following records shall be maintained in accordance with Ref. 4.1.: , 7.11.2.1 Lifetime: 7.11.2.1.1 Revised material from Vendor Manuals 7.11.2.1.2 Deleted Class 1 Vendor Manuals 7.11.2.1.3 Vendor Manual Change Requests V
I . . l onge 9 of 10 tCP84A4 levised
. P.9/85 l
7.11.2.2 Two years from completion: 7.11.2.2.1 Vendor Manual inventory results 9 1
4 Page 10 of 10 NOP84A4 Revised 9/19/85 VENDOR MANUAL CHANGE RE0 VEST RType V10.00 QA Record MANUAL TYPE (Check one) [] Chemistry J] Architectural JCivil [] R . P . [] Electrical []I&C [] Mechanical
- 1. PNPS Vendor Manual Number:
- 2. Vendor Manual
Title:
- 3. Revision ((] Deletion Date:
((]ManualAddition
- 4. PNPS Manual Revision Number:
- 5.
- Manual Cross-
Reference:
- 6. Manual Pages Affected:
- 7. Vendor Name:
- 8. Equipment Number:
9.** Component Cross-Reference No.:
- 10. Owner of Manual:
- 11. Manual Classification:
- 12. Comments / Justification:
Manual Control Initiated by: Date: 13. Manual Validated by: Odte: 14.
- 15. Manual Reviewed by (owned) Date: __
- 16. Manual Approved For Distribution and Use Date:
- 17. Manual Distribution Matrix
)
- 18. Disposition of Safety Evaluation - S.E. ho. (See Ref. 4.6)
If Not Required, Provide Justification: . i 1 I
*All other manual #'s (i.e. Vendor #, BECo PO f. etc.)
** Mark #'s, Serial #'s, Series #'s, Model #'s, etc.
"~ Exhibit 1 l I
----_-__m. _ _ _ _ _ _ _ _ _
80 N Page 1 Of 1: nop es:
# EDISON.
NUCl. EAR OPERATIC *NS PROCEDURE OPERATING EXPERIENCE REVIEW PROGRAlf Submitted: f[w t/f/gy Prepa er Ja't e' 9 Lk&ah Dept. Mgr/ Cate QA Approval: QAM
/bu% A//f//'/ Cate QA r: gram related: Q Yes O h:
Approval: DNO [v 3/II/ satt f M ZbibY E (Q ~ ' ~ Cate k . l SVPN Q ' Ca t'e Effective Date:
)
yp v#6p' a;r Ji[re -
Page 2 of 14 NOP 8401 OPERATING EXPERIENCE REVIEW PROGRAM I. POLICY STATEMENT It is the policy of the Boston Edison Company to utilize the accumulated experience of the Nuclear Industry to improve and maintain the quality of , operations at Pilgrim Nuclear Power Station. To fulfill this policy, a program of assessment of operating experience is to be implemented as described within this document. II. OBJECTIVE OF THE OPERATING EXPERIENCE RCVIEW PROGRAM A. Industry and internal operating experiences should be evaluated and appropriate actions taken to improve the plant safety and 1 reliability. B. Elements of this N0P are not to replace or modify previously established requirements or responsibilities addressed in NOP 83Al, Nuclear Operations Organization er NOP 83A9, Management Corrective Action Program. III. SCOPE The Operating Experience Review Program (OERP) shall address both external and internal sources of operating experience. A. External sources of operating experience are limited to:
- 1. NRC Bulletins and Notices
- 2. INPD SOERs j
- 3. INPO SERs
- 4. General Electric SILs and TILs
- 5. INPO 08MRS
- 6. Nuclear NETWORK Inf ormation pertinent to plant operating experience.
B. Internal sources of operating experience information are:
- 1. Post Trip Review Reports
- 2. Failure and Malfunction Reports
- 3. Plant Licensee Event Reports
- 4. Reports generated on a non-routine basis as part of an investigation at a plant event 1
a Page 3 of 14 NOP 8401 IV. DEFINITIONS A. SER - Significant Event Reports are reports generated by INPO based upon Industry events and are issued to provide timely information to other utilities. SERs describe events and cause of the evemt. SERs are precursors of 50ERs and are prompt bulletins to the industry. B. SOER - Significant Operating Experience Reports are generated by I INP0 based upon industry events and are issued after more detailed l information is accumulated. SOERs contain specific recommendations to address the cause or , prevention of the initiating event. SOERs are categorized as requiring; immediate attention (RED TAB), Prompt Attention (YELLOW TAB), Normal Attention (GREEN TAB), C. O and MR - Operating and Maintenance Reminder; are issued by LWP0 and serve to provide additional suggested betterments for har6 ware problems. D. OER Item - a report or document containing operating experience information to be evaluated. E. Action Item - the result of an OER review that requires an action to address and close-out an OER item. F. SIL - Service Information Letter is issued by GE and contains recommendations concerning Nuclear Steam Supply system equipment. , i G. TIL - Technical Information Letter is issued by GE and contaims j recommendations concerning Turbine-Generator system equipment, and I by their nature are generally of commercial nature. V. RESPONSIBILITIES A. NUCLEAR OPERATIONS MANAGER Responsible for the assignment of appropriate group (s) to ace,mplish action items resulting from the OER process, submitted by the Onsite Safety and Performance Group leader. Refer 6ction items to other managers as appropriate. I l
I l Page 4 of 14 NOP 8401 l B. NUCLEAR MANAGEMENT SERVICES MANAGER Responsible for the evaluation and corrective actions resulting f rom NRC items as conducted by the Regulatory Af f airs and Program Group. C. NUCLEAR OPERATIONS SUPPORT MANAGER Responsible for accomplishing corrective actions referred by the NOM. This is to incorporate other Compliance and or Licensing interfaces. D. NOCLEAR ENGINEERING MANAGER Responsible for the assignment of appropriate discipline group (s) to accomplish action items resulting from OER process and requested by the Onsite Safety and Performance Group Leader. E. 0A MANAGER Responsible to provide an annual independent review to determine ef festiveness of the overall Operating Experience Review Program. F. NUCLEAR TRAINING MANAGER Responsible to incorporate approved action items into training programs . G. N00. NED. NMSO. and NOSD. GROUP LEADERS These groups are responsible for:
- 1. The review and acceptance of action items.
- 2. Dissemination of Operating Experience information to personnel under their supervision. The PNPS COE is to provide traceability of this dissemination to licensed and non-licensed personnel as appropriate.
- 3. Completien of approved action items.
- 4. Provide status of Actions items to PS&CC
- 5. Identification of events or occurrences of interest to the industry to the OSS&P Group Leader.
H. PS&CC GROUP IEA0_ER Responsible to mai'atain a tracking system for action items resulting from the OER process and track the items to completion, and to solicit needed input data.
Page 5 of 14 NOP B401 I. OSS&P GROUP LEADER
\
Responsible for:
- 1. Coordination of prcgram requirements with or. site and offsite groups.
- 2. Supervision of OSS&P Group activities to accomplish OER Reviews suggested action items and information distribution.
- 3. Implement procedures for the review, disposition dissemination of OER items.
- 4. Provide a monthly report to the Nuclear Organization Managers on the status of OER items.
J. RMG - PILGRIM STATION
- 1. Responsible for copying services and distribution of OER 4 information transmittals as required.
- 2. Forward copies of incoming NRC correspondence and incoming GE SIL's and TILS to OSS&P Gr oup Leader.
VI. REFERENCES s' A. INPO Good Practice T5403, " Industry Event Operating Experience Review Program".
'B. INPO Good Practice, T5406, 'In-House Operating Experience Review /
Program' C. INPO Performance Objectives, TS.3 of the Operating Experience Review Program
/.
D NUREG 0737, Item I.C.5 E. BEco Letter, " Response to 1983 INP0 Evaluation Report' F. BECo QA Manual G. NOP B3Al Nuclear Operations Organization H. NOP 83A9 Management Corrective Action Program VII. OER PROGRAM REQUIREMENTS Operating experience review incorporates six dif ferent elements. These elements are: o Review of OER items for applice.bility, and significance (INPO and BEQAM Criteria).
i i
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Page 6 of 14 NOP 8401 o Evaluation of items for preliminary corrective action (for SERs) Review of recommendations for action items (other items). ' o Assignment approval and completion of action items o Dissemination of information o Use of Nuclear NETWORK o Reporting of OER Status . The requirements of major steps of each of.these elements are described in the following sections. Each affected Manager, Chief Engineer or Group Leader shall implement methods or procedures to meet the applicable requirements of each element. A. Review of OER items for applicability and significance
- 1. All incoming items shall be entered into the incoming OER Log. Incoming items f rom either external or internal sources, as described in Sections III, A and B shall be screened for applicability and significance. Applicability, INPO significance criteria and BEQAN significance are exhibits 2, 3 and 4 respectively.
- 2. Any items becoming applicable shall be referred to in Section VII.D for dissemination of information and be reviewed for significance. Non-applicable inf ormation shall be logged as closed.
- 3. Any items determined to be significant or conditionally significant by INPO criteria shall be referred for evaluation. l
- 4. Any '. tem determined to be significant by BEQAM criteria shall be evaluated per NOP 83A9.
- 5. Exceptions:
NRC Bulletins and notices will be compared for existing INP0 documents and applicable INPO documents shall be forwarded to Regulatory Af f airs and Programs Group Leader for information feedback. Evaluation and actien on NRC items remain the responsibility of RA&P. 1
l 1 l Page 7 of 14 NOF S401 i i B. Evaluation ]
- 1. SERs will be evaluated for preliminary corrective action or if j current procedures adequately address the ites. Action Items resulting f rom this evaluation shall enter the Action Item Process VII.C. I
- 2. 50ER sit's will be evaluated on a per recommendatice basis.
Recommendations shall be evaluated against currett procedures and design configuration for applicability. Where current procedures or design do not meet the recommendations intent an action ites will be issued for action as outlinec in Section VII.C. Recommendations which are met by existin5 procedures or design shall be closed-out.
- 3. Internal items shall undergo an evaluation for rcot cause and corrective action or concurrence with results of a cause and corrective action determination. Resulting acticn items will !
be issued for action per Section VII.C. Information distribution of internal items shall be made consistent with the need for providing " lessons learned" f ree a 5 1ven event per section VII.O. Additionally internal items will be reviewed for output via Nuclear NF.TWORK per VII.E.
- 4. Evaluations may be requested of other cognizant $roups where a technical speciality exists. The priority of the requested "
evaluation will be specified in a cover meno or LSR. The request shall clearly state whether a Regulatory or IN?0 commitment is involved. C. Action Item Process
- 1. The result of an evaluation of an OER item may require an action prior to closeout. For purposes of this procedure, such results will be referred to as Action Items.
- 2. Action Items will be submitted to the OSS&P Grous Leader for review for appropriateness. The OSS&P Group Leater will then submit or discuss the action item with the Cognizant Group Leader or Chief Engineer.
Page 8 of 14 NOP 8401
- 3. The Cognizant Chief or Group Leader will review the action l item for agreement / acceptance. If the Cognizant Chief or f Group Leader does not agree with the proposed action item or assignment, he/she will so inform the OSS&P Group Leader with the reasons and possible alternatives. The Cognizant Chief and OSS&P Group Leader shall attempt to resolve the difference for acceptance prior to proceeding. 1
- 4. An accepted action item will be submitted by the OSS&P Group I Leader to the Cognizant Manager for assignment to the Cognizant Chief Engineer or Group Leader. An unaccepted action item which cannot be resolved will be presented to the Cognizant Manager for review and disposition. If agreement cannot be reached, the matter will be referred to the next higher level of management.
- 5. Steps 2,3, and 4 of this section may be performed within the context of a forum of Chief Engineers and Group Leaders for discussion assignment and disposition of Action Items. An Action Item may be dispositioned as accepted and assigned not applicable, or applicable but action not warranted for sound j engineering or managerial cause. Such dispositions shall be !
doc un.onted .
- 6. When an action item is accepted, the OSS&P Group Leader will l inform the PS&CC Group of the item for entry into the t'acking {
systam. A member of PS&CC will request an action plan from ) the implementing group. Based upon this action plan, PS&CC will track the action ites to completion.
- 7. The implementing department will proceed according to the 1 1
action plan. When action is complete, documentation will be j forwarded to the OSS&P Group Leader.
- 8. The OSS&P Group Leader will review the action item documentation for close-out. The OSS&P ticup Leader will inform PS&CC of the close-out and the item will be logged as I
closed in the OER Log. I I
)
Page 9 of 14 NOP 8401 D. Dissemination of Information
- 1. All applicable OER items will be disseminated for information to the appropriate groups.
]
- 2. The OSS&P Group member performing initial eview and evaluation shall indicate on a transmittal sheet the appropriate groups receiving information.
The reviewer will also convert abstract or lengthy documents into more understandable format. The reason for applicability to PNPS and the possible consequences shall be clearly and concisely stated. The original OER items any be attached for further information.
- 3. The OSS&P Group Leader will review the informational ,
transmittal prior to dissemination.
- 4. Group Leaders, Managers or Chief Engineers reieiving such informational transmittals shall review an: dispatch the i information by methods appropriate to their department. SERs should also be reviewed for impact and preliminary l preventative action. The receiving person is responsible for
)
information dissemination of the work unit.
- 5. An exception to Step D.4 above is that material received by the PNPS Chief Operating Engineer (COE) shall be formally docunanted as reviewed by Operations personnel. Operations personnel who receive information may be Watch Engineers, Operating Supervisors, licensed and unlicessed NPO's and Radwaste operators depending on the nature or subject of the information. The COE shall determine whici personnel will be information recipients. Such completed do:umentation is to be returned to the OSS&P Group Leader for record retention.
- 6. The ORC shall be apprised of the status of OER activities, especially items determined to be INP0 Sig11ficant, by a brief sunination at ORC meetings as soon as practicable af ter such determination is made.
Page 10 of 14 , NOP 8401 E. -Use of Nuclear NETWORK
- 1. Internal events determined to be of interest to the industry by the OSS&P Group Leader shall be entered into the operatifs experience category of Nuclear NETWORK. l
- 2. The approval of the Nuclear Operations Manager or designee shall be abtained prior to entry into NETWORK.
- 3. Examples of information of interest to the industry are not .l necessarily plant trips but their causes; additionally, off-normal conditions resulting from equipment failures are also appropriate NETWORK entry items.
- 4. INP0 SERs and O&MRs are also first received via NETWORK. This will serve as the primary entry point for this information. ;
- 5. OSS&P Group members accessing Nuclear NETWORK shall maintain a logbook of information sent and SERs and O&MRs received.
F. OER Status Reports
- 1. In order to keep management aware of OER activities, a report will be issued by the OSS&P Group leader on a monthly basis.
- 2. Content of this report shall be as a minimum:
o New items received with brief description and disposition. o Items closed-out with description and by what group /da9artment. o Status of action items over 3 months old. o Listing of information items sent out in the report period and departments receiving the information.
- 3. The report shall be directed to the Director, Nuclear Operations with copies to the VP-NE&QA, NSRAC Chairman, NOM, NMSM, and other managers as requested.
- VIII. EXHIBITS
- 1. OER Program Requirements Flow Chart
- 2. Applicability Screening Criteria
- 3. INP0 Significant Screening Criteria
- 4. BEQAM Significance Screening Criteria i
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Page 12 of 14 NOP 8401 l APPLICA81LITY SCREENING CRITERIA Items are Applicable if System, equipment or process are: o Generic o From BWR of equivalent design o System or Component Similar to PNPS o Procedural or Operator Error or Equipment f ault which could occur at PNPS , o INPO or NRC designation as applicable l Exhibit 2
1 Page 14 of 14 NOP 8401 l BEQAM SIGNIFICANCE SCREENING CRITERIA l l
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DIRECTIONS: i Carefully review the problem report description (discuss the problem description with the report originator and other responsible individuals to obtain a good understanding of the problem) and evaluate whether it meets any of the following criteria for a 'significant" condition adverse to quality. THIS CONDITION REPRESENTS:
- 1. A breakdown in any portion of the Quality Assurance Program Y N conducted in accordance with the requirements of Appendix B to 10CFR50. !
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- 2. A deficiency in design such that the design does not conform to Y N I the criteria and bases stated in the final safety analysis report. I I
- 3. Damage to a structure, system, or component which will require Y N extensive evaluation, extensive redesign, or extensive repair ;
to meet the criteria and bases stated in the final safety { analysis report, or to otherwise establish the adequacy of the j structure, system, or component to perfone its intended safety l function.
- 4. A deviation f rom performance specifications or design drawings Y N which will require extensive evaluation, extensive redesign, g or extensive repair to establish the adequacy of structure ,
system, or component to meet the criteric and bases stated in the final safety analysis report or to otherwise establish the i adequacy of the structure, system, or component to perform its intended safety function.
- 5. The failure of malfunction of, or use of nonconforming material Y N in a structure, system, or component which will require extensive redesign or extensive repair to establish the adequacy of a structure system, or component to meet the criteria and bases stated in the safety analysis report or to otherwise establish the adequacy of l the structure, system, or component to perforin its intended safety l function. )
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- 6. The repetitive recurrence of a deficiency not covered by the items Y N above.
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EGG-NTA--7188 Revised Draft CONFORMANCE TO GENERIC LETTER 83-28 ITEM 2.1 (PART 1) EQUIPMENT CLASSIFICATION (RTS COMPONENTS) SELECTED GENERAL ELECTRIC BOILING WATER REACTOR PLAi!TS HOPE CREEK PEACH BOTTOM 2 AND 3 j PERRY 1 AND 2 PILGRIM 1 R. HAROLOSEN Published September 1986 l EG&G Idaho, Inc. Idaho Falls, Idaho 83415 , 1 1 Prepared for the U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Under 00E Contract No. DE-AC07-761001570 FIN Nos. 06001 and 06002 I Otb ' Ch 2 * $0 P Y' ' ,
1 l~ ABSTRACT This EGGG Idaho, Inc. report provides a review of the submittals from selected operating and applicant Boiling Water Reactor (BWR) plants for conformance to Generic Letter 83-28 Item 2.1 (Part 1). The following
. plants are included in this review.
Plant Name Docket Number TAC Number Hope Creek 50-354 OL Peach Bottom 2 50-277 52865 Peach Bottom 3 50-278 52866 Perry 1 50-440 61705 Perry 2 50-441 OL Pilgrim 1 50-293 52867 FOREWORO This report is supplied as part of the program for evaluating licensee / applicant conformance to Generic Letter 83-28, " Required Actions Based on Generic ImpTTEations of Salem ATWS Events." This work is being conducted for the U. S. Nuclear Regulatory Commission, Of fice of Nuclear Regulation, Division of PWR Licensing-A, by the EG&G Idaho, Inc. The U. S. Nuclear Regulatory Commission funded this work under the authorization B&R 20-19-10-11-3 and 20-19-40-41-3, FIN Nos. 06001 and 06002. e
i ( l CONTENTS A85 TRACT .............................................................. 1) FOREWORD .............................................................. 1)
- 1. INTRODUCTION AND
SUMMARY
......................................... 1
. 2. PLANT RESPONSE EVALUATIONS ....................................... 3 2.1 Hope Creek ................................................... 3 2.2 Conclusion ................................................. 3 2.3 P e a c h B o t t om 2 a nd 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.4 Conclusion ................................................. 5 2.5 Perry 1'and 2 .............................................. 6 2.6 Conclusion ................................................. 6 2.7 Pilgrim 1 .................................................. 7 2.8 Conclusion ................................................. 7
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- 3. GENERIC REFERENCES ............................................... 8 i
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. 1
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- 1. INTRODUCTION AND
SUMMARY
Dn February 25, 1983, both of the scram circuit breakers at Unit 1 of the Salen Nuclear Power Plant failed to open upon an automatic reactor trip signal from the reactor protection system. This incident was terminated manually by the operator about 30 seconds af ter the initiation of the automatic trip signal. The failure of the circuit breakers was determined to be related to the sticking of the undervoltage trip attachment. Prior to this incident, on February 22, 1983, an automatic trip signal was generated at Unit 1 of the Salem Nuclear Power Plant based on steam generator low-low level during plant startup. In this case, the reactor was tripped manually by the operator almost coincidentally with the automatic trip. Following these incidents, on February 28, 1983, the NRC Executive Director for Operations (E00), directed the staf f to investigate and report on the generic implications of the occurrences at Unit 1 of the Salem Nuclear Power Plant. The results of the staf f's inquiry into the generic j implications of the Salem Unit 1 incidents are reported in NURE3-1000, {
" Generic Implications of the ATWS Events at the Salem Nuclear Power j Plant.
- As a result of this investigation, the Commission (NRC) requested (by Generic Letter 83-28, dated July 8,1983 ) all licensees of operating reactors, applicants for an operating license, and holders of construction permits to respond to generic issues raised by the analyses of these two ATWS events.
This ' report is an evaluation of the responses submitted from a selected' group of Boiling Water Reactors (BWRs) for Item 2.1 (Part 1) of Generic Letter 83-28. The results of the review of four individual plant responses are combined and reported on in this document to enhance review efficiency. The specific plants reviewed in this report were selected based on the 1
convenience of review. The actual documents which were reviewed for each evaluation are listed at the end of each plant evaluation. The generic documents referenced in this report are listed at the end of the report.
. Part 1 of Item 2.1 of Generic Letter 83-28 requires the licensee or applicant to confirm that all reactor trip system components are . identified, classified, and treated as safety-related, as indicated in the following statement:
Licensees and applicants shall confirm that all components whose functioning is required to trip the reactor are identified as saf ety-related on documents, procedures, and inf orrr.ation handling systems used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement. 1 1 l l e 2
i f
- 2. PLANT RESPONSE EVALUATIONS 2.1 Hope Creek 50-354 (OL)
. The applicant for Hope Creek (Public Service Electric and Gas Company) provided responses to the requirements of Item 2.1 (Part 1) of
- Generic Letter 83-28 in submittals dated March 30, 1984, Decenter 17, 1984 and May 21, 1985. In the first submittal the applicant described their plan to develop a Master Equipment List (MEL) which would identify the components required to initiate reactor trip and designate these components as safety-related. The MEL imposes quality assurance requirements for the l safety-related components and is the r;ctrolling document for safety-related activities. The applicant 5:sted intentions to be in compliance with Item 2.1 (Part 1) prior to Septeaber 1984.
The second submittal reviewed progress to December 17, 1984 and , outlined a revised program which would meet the requirements of Item 2.1 (Part 1) prior to March 1985., The applicant confirmed in their May 21, 1985 submittal that review of the reactor trip system had been completed and that reactor trip system components were verified to be classified i safety-related on appropriate design documents, however, the MEL had not been completed for all components of the reactor trip system. The ! applicant stated that this effort would be completed by September 30, 1985. 2.2 Conclusion I Based on a review of the applicant's submittals, we find that the applicant's responses confirm that components required to trip the reactor have been designated safety-related and that the MEL is used to control all activities relating to safety-related components. We, therefore, find that the applicant's responses meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-28, and are acceptable. 3
REFERENCES
- 1. Letter, R.L. Mitti, Public Service Electric and Gas Co., to A.
Schwencer, NRC, March 30, 1984.
- 2. Letter, R.L. Mitti, Public Service Electric and Gas Co., to A.
Schwencer, NRC, December 17, 1984.
- 3. Letter, R.L. Nitti, Public Service Electric and Gas Co., to W. Butler,
.- NRC, May- 21, 1985.
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2.3 Peach Bottom 2 50-277 TAC NO. 52865 Peach Bottom 3 50-278 TAC NO. 52866 The licensee for Peach Bottom 2 and 3 (Philadelphia Electric Co.) ] provided responses to the requirements of Item 2.1 (Part 1) of Generic l Letter 83-28 in submittals dated Novencer 4, 1983, April 23, 1984 and 1 May 29, 1985. The responses state that all systems that contribute to the reactor 1 trip function have been identified as safety-related in the current "Q" list and that all components of safety-related systems are safety-related ,
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unless specifically excluded by safety evaluation. The "Q" list is used to identify the applicable codes, standards and procedures to be used for activities relating to the safety-related components, j f ach item or service to be procured is reviewed to determine if it is safety-related. The review is perforned by a congnizant member of the plan staf f or the Engineering and .Research Depar tnent. 2.4 Conclusion Item 2.1 (Part 1) requires licensees to confirm that all components whose functioning is required to trip the reactor are identified as r safety-related on documents, procedures, and information handling systmas used in the plant to control safety-related activities, including maintenance, work orders, and parts replacement. Based on the licensee's submittal we find that the list of components required to trip the reactor is incomplete. We also ; tad that the
- licensee's program does not identify safety-related components on relevant plant documents. The response, therefore, does not meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 and is unacceptable. ,
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REFERENCES
- 1. Letter, S.L. Daltroff, Philadelphia Electric Co., '.o D.G. Eisenhut' NRC, November 4,1983.
. 2. Letter, S.L. Daltroff, Philadelphia Electric Co., to D.G. Eisenhut, NRC, April 23, 1984.
- 3. Letter, S.L. Daltroff, Philadelphia Electric Co., to J.F. Stolz, NRC May 29,1985.
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i 9 2.5 Perry 1 50 440 (OL) and perry 2 50-441 (OL) l I The applicant for Perry 1 and 2 (Cleveland Electric Illuminating Co.) provided responses to the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 in submittals dated April 6, 1984 and August 28, 1985. The applicant reported in the first submittal that the "Q'-list for the plants 1 was undergoing review to verify the correct classification of safety-related components. The "Q'-list is to be used to determine classification for maintenance, work orders and procurement activities. The second submittal reported that the 'Q'-list evaluation had been completed and that all numbered components from the 5 systems that contribute to the reactor trip function had been reviewed and classified as safety-related or ncnsafety-related. The "Q'-list is the safety-related subset of the Perry Equipment Kaster Files System (PEMS) used to setermine l the classification for work orders, maintenance and parts procurement. 2.6 Conclusion Based on the review of the applicant's submittals, we find tsat the applicant has verified that the components necessary to perform reactor trip are classified as safety-related and that this classificatica program imposes safety-related procedures on work orders, maintenance, and procurement activities. We, theref ore, find that the applicant's response meet the requirements of Item 2.1 (Part 1) of Generic Letter 83-25 and are acceptable. { REFERENCES
- 1. Letter, M.R. Edelman, Cleveland Electric Illuminating Co. te 0.G.
Eisenhut, NRC, April 6, 1984.
- 2. Letter, H.R. Edelman, Cleveland Electric Illuminating Co., to 8.J.
Youngblood, August 28, 1985. I ( l 1 1
2.7 Pilarim 1. 50-293. TAC Wo. 52867 The licensee for Pilgrim 1 (Boston Edison Co.) provided responses to the requirements of Item 2.1 (Part 1) of Generic Letter 83-28 in submittals
, dated November 7, 1983 and June 28, 1985. In the submittals the licensee confirmed that the components required to function for reactor trip are
' identified in the plant "Q"-list and are controlled at a quality level which reflects the safety-related functions. Documents (Purchase Orders, Maintenance Requests) used to control activities associated with the "Q" listed equipment are identified as "Q" which designates the use of safety-related procedures.
2.8 Conclusina Based on the review of the licensee's submittals, we find that the licensee has verified that the components necessary to perform reactor trip are c'.assified as safety-related and that the classification program imposes safety-related procedures on maintenance and procurement activities relating to the components. We, therefore, find that the licensee's response meet the requirements of Items 2.1 (Part 1) of Generic Letter 83-28 ar.d are acceptable. REFERENCES
- 1. Letter, W.D. Harrington, Boston Edison Co. to D.B. Vassallo, NRC, November 7, 1983.
- 2. Letter, W.D. Harrington, Boston Edison Co., to D.8. Vassallo, NRC, June 28, 1985.
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- 3. GENERIC Rif ERENCES
- 1. Generic Implications of ATWS Events at the Salem Nuclear Power Plant.
l NUREG-1000 Volume 1. April 1983; Volume 2, July 1983.
- 2. NRC Letter, D.G. Eisenhut to all Licensees of Operating Eeactors, l'
Applicants for Operating License, and Holders of Construction Permits,
" Required Actions Based on Generic Implications of Sales Armis Events (Generic Letter 83-28) " July 8,1983.
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