ML011870065

License DPR-35, Amendment 13

ML011870065
Person / Time
Site:	Pilgrim
Issue date:	11/01/1975
From:	Ziemann D Office of Nuclear Reactor Regulation
To:	Howard J Boston Edison Co
References
Download: ML011870065 (38)

Text

DISTRIBUTION YDocket NRC PDR TJCarter EP LA Local PDR EP P NOV 0 7975 ORB #2 Reading P Collins OELD SVarga Woeket N'o. 50-29? OI&E (3) CHebron NDube AESteen BJones (4) DEisenhut Ecston Fdisor Corpary JmCGought ACRS (16)

ATTN: ,r. J. E. toward JSaltzman, OAI TBAbernathy, D¶FIE Vice President - NJuclear PdDiggs 5On oylston Street PWO'Connor Eoston, Massachusetts 02199 DLZiemann KRgoller Gentlemen: SKari BSCharf (15)

The Copnission has issued the einclosed Arerdrent N~o. 13 to Facility Operotirr License No. DPR-35 for the Pilgrir Nuclear Power Station.

This amendrent includes Change Fo. 15 to the Technical Specifications, ADpendiy A, and is in response to your requests dated February 14, 1975, Jure 2, 1975, ard July 29, 1975.

The amendrent incorporates into the Pilgrim Nuclear Power Station Technical Specifications changer to the Administrative Controls.

Chpnpes to your proposal were necessary to meet our requirerents.

These have been discussed with your staff. The Technical Specifications are bosed on the Regulatory positions described in guides 1.8, "Personnel Selection and Training", 1.16, "Reportinr of Operating Information Appendix A Technical Specifications", Revision 4, and l.??, "Oualitv Assurance Prcprar Peouirerents".

We renuest that you use the fornats presented in the Apperdices to Yerulatory Cuide 1.16, Pevisior 4, for reportinF operating infornation anP that you report events of the type described under the section "Events of Potential Public Interest". Instructions for using these reporting forrats are contpined in Rerulatory Guide 1.16 (a copy is enclosed for your use), and AEC report OCE-SS-O01 titled "Instructions for Preparation of Data Entry Sheets for Licensee Fvent Report (LER)

File" of which you were previously provided a copy. This report is modified by updated instructions dated Auqust 21, 1975 which are enclosed. Copy requirerents are sulrarizae in Regulatory Guide 10.1, "Corpilation of Reporting Fecuirevents for Persons Subject to AEC Repulateons", a copy of which is also enclosed. This Cuide will assist you in identifying reports that are required by the Commission's regulations set forth in Title 10 Code of Federal Repulations but are not contained in your Technical Specifications. Reports that are required by the regulations have not been repeatee ir your Technical Specifications.

Copies of the related Safety Evaluation and the Federal Pegister Notice also are enclosed. I Sincerely, Col

/-

OF FICE R 2, U ....... ernr L. n, Chle. E .ra SURNAME PWOConno.r...... Diviion....a.ctor.Lie* *s * ..DLZemann AT F o r m.

10i /7S 101.

13O 10/ /7 i /75.....

Form AEC-318 (Rev. 9-53) A*ECM 0240 " U. S; GOVERNMENT PRINTING* OFF'ICER 1 7 - .

Boston Edison Company NOV 0 1915

Enclosures:

1. Amendment No. 13

2. Regulatory Guide 1.16

3. Updated Instructions

4. Regulatory Guide 10.1

5. Safety Evaluation

6. Federal Register Notice cc w/enclosures:

Mr. J. E. Larson Anthony Z. Roisman, Esquire Nuclear Licensing Administrator - Operations Berlin, Roisman and Kessler Boston Edison Company 1712 N Street, N. W, RFD #1, Rocky Hill Road Washington, D. C. 20036 Plynouth, Massachusetts 02360 Plymouth Public Library Mr, G. Carl Andognini North Street Nuclear Operations Manager Plymouth, Massachusetts 02360 Boston Edison Company 800 Boylston Street Mr. David F. Tarantino Boston, Massachusetts 02199 Chairman, Board of Selectmen 11 Lincoln Street Mr. J. A. Smith Plymouth, Massachusetts 02360 Pilgrim Station Manager Boston E~dison Co>npany cc w/enclosures and cy of BECo's RFD 1l, Rocky Hill Road filings dtd. 2/14/75, 6/2/75 Plymouth, Massachusetts 02360 and 7/29/75:

Henry Kolbe, 1. D.

Mr. E. F. Kearney Acting Commissioner of Public Healtl" Operations Engineering Group Manager Massachusetts Department of Public Boston Edison Company Health RFD #1, Rocky Hill Road 600 Washington Street Plymouth, Massachusetts 02360 Boston, Massachusetts 0211]

Mr, W, H, Sides Quality Assurance Manager Boston Edison Company RFD #1, Rocky Hill Road Plymouth, Massachusetts 02360 Mr. D. G. Stoodley, Counsel Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. 0. C. 20555 IOSTON EDISON COMPANY DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION AMENDNENT TO FACILITY OPEAT\T:NC LI CENSE Amendment No. 13 License No. DPR-35 I.' The Nuclear Regulatory Commission (the Commission) has found that:

A. The applications for amendment by Boston Edison Company (the licensee) dated February 14, 1975, June 2, 1975 and July 29, 1975, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 Commission; C. There is reasonable assurance (i) that the activities authorized 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 Commission's regulations; and D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

2. Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Facility License No. DPR-35 is hereby amended to read as follows:

"1B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change No. 15."

3. This license amendment is effective 30 days after the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 0Den; - L. by:

Dennis L. Ziemann Dennis L. Zibm-am-n---h4*X Operating Reactors Branch #2 Division of Reactor Licensing

Attachment:

Change No. 15 to the Technical Specifications Date of Issuance: NOV 04 1975 O.

S U R N A M E 1-

...ii iiiiiiii iiiiiiii iiiii iiiii ii Re D A TE3 "I'll . . 020. ......

.. ... ........................................... ............. P O. .

FOrm AE*C-318 (Rev. 9-53) A*EC*[ 0240

• U. S; GOVERNMENT PRINTINGS OFFICES 1974-526-166

ATTACHMENT TO LICENSE AMENDMENT NO. 13 CHANGE NO. 15 TO THE TECHNICAL SPECIFICATIONS FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO, 50-293 Replace pages i, ii, 1, 2 and 208 through 230 with the attached pages i, ii, 1, 2 and 208 through 231. The revised pages have marginal lines indicating where the changes appear.

TABLE OF CONTENTS Page No.

1.0 DEFINITIONS I SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 1.1 FUEL CLADDING INTEGRITY 2.1 6 1.2 REACTOR COOLANT SYSTEM 2.2 22 LIMITING CONDITIONS FOR OPERATION SURVE£I LLANCE REQUIREMENT 3.1 REACTOR PROTECTION SYSTEM 4.1 26 3.2 PROTECTIVE INSTRUMENTATION 4.2 42 3.3 REACTIVITY CONTROL 4.3 80 A. Reactivity Limitations A 80 B. Control Rods R 81 C. Scram Insertion Times C 83 D. Control Rod Accumulators D 85 E. Reactivity Anomalies E 86 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 95 A. Normal Operation A 95 B. Operation with Inoperable Components B 96 C. Sodium Pentaborate Solution C 97 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 103 A. Core Spray and LPCI Subsystems A 105 B. Containment Cooling Subsystem B 106 C. HPCI System C 107 D. RCIC System D 108 E. Automatic Pressure Relief Subsystem E 109 F. Minimum Low Pressure Cooling System Diesel Generator Availability F 110 H. Maintenance of Filled Discharge Pipe H i1 J. Average Planar LHGR J 112A 3.6 PRIMARY SYSTEM BOUNDARY 4.6 123 A. Thermal and Pressurization Limitations A 123 B. Coolant Chemistry B 124 C. Coolant Leakage C 125 D. Safety and Relief Valves D 126 E. Jet Pumps E 127 F. Jet Pump Flow Mismatch F 127 G. Structural Integrity G 127 H. High Energy Piping (outside containment) H 127A i

1.0 DEFINITIONS The succeeding frequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.

A. Safety Limit - The safety limits are limits below which the rea sonable maintenance of the cladding and primary systems are as sured. Exceeding such a limit is cause for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself re sult in serious consequences but it indicates an operational de ficiency subject to regulatory review.

B. Limiting Safety System Setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the auto matic protective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represent margin with normal operation lying below these settings. The margin has been established so that with proper op eration of the instrumentation the safety limits will never be ex ceeded, C. Limiting Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system per formance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be opera ted safely and abnormal situations can be safely controlled.

D.

1s

1.0 DEFINITIONS (Cont'd)

E. Operable - A system or component shall be considered operable when it is capable of performing its intended function in its required manner.

F. Operating - Operating means that a system or component is perform ing its intended functions in its required manner.

G. Immediate - Immediate means that the required action will be ini tiated as soon as practicable considering the safe operation of the unit and the importance of the required action.

H. Reactor Power Operation - Reactor power operation is any operation with the mode switch in the "Startup" or "Run" position with the reactor critical and above 1% design power.

I. Hot Standby Condition - Hot standby condition means operation with coolant temperature greater than 212*F, system pressure less than 600 psig, the main steam isolation valves closed and the mode switch in startup.

J. Cold Condition - Reactor coolant temperature equal to or less than 2126F.

K. Mode - The reactor mode is that which is established by the mode selector-switch. The modes include shutdown, refuel, startup and run which are defined as follows:

1. Startup Mode - In this mode the reactor protection scram trips, initiated by condenser low vacuum and main steam line isolation 9

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY The Pilgrim Station Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.

6.2 ORGANIZATION A. OFFSITE The Company organization for station management and technical support shall be as shown on Figure 6.2.1.

B. FACILITY The Facility organization shall be as shown on Figure 6.2.2 and:

I. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2.1.

2. At least one licensed Operator shall be in the control room when fuel is in the reactor.

3. At least two licensed Operators shall be present in the control room during reactor startup, scheduled reactor shutdown and during recovery from reactor trips.

4. An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

5. ALL CORE ALTERATIONS performed while fuel is in the reactor vessel after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

6.3 FACILITY STAFF QUALIFICATIONS of The qualifications with regard to educational and experience backgrounds the faeility staff at the time of appointment to the active position shall meet the requirements as described in the American National Standards Institute N18.1-1971, "Selection and Training of Personnel for Nuclear Power Plants."

6,4 TRAINING A. A retraining and replacement training program for the facility staff shall be maintained under the direction of the Pilgrim Station Manager.

under B. A retraining program for the licensed operators shall be maintained and shall meet or the direction of the Senior Nuclear Training Specialist 5.5 of ANSI N18.1-1971 exceed the requirements and recommendations of Section and Appendix A of 10 CFR Part 55.

208

Office Of the President sident tcutive Vice Presiden' lor Vice Presidents

-ear Safety Review Audit Committee Nuclear Operations Staff Assistants Nuclear Licensi Administrator Operations Figure 6.2.1 Boston Edison Company - Organization for Operation of Pilgrim I

- 209 -

STATION MANAGER (ORC) (NSRAC)

(

PILGRIM STATION ORGANIZATION Figure 6.2.2 Code: NSRAC - Member of NSRAC ORC - Member of ORC RO - NRC Reactor Operator License SRO - NRC Senior Reactor Operator License 210

PILGRIM _NUCLEAR POWER STATION MINIMUM SHIFT CREW COMPOSITION TECHNICAL SPECIFICATION TABLE 6.2-1 STATION CONDITION

• C R E W MINIMUM NUMBER ON DUTY t I OPERATING Licensed Senior Reactor Operator 1 Licensed Reactor Operator 2 Unlicensed Operator 2 COLD SHUTDOWN Licensed Senior Reactor Operator 1 Licensed Reactor Operator 1 Unlicensed Operator

• Higher grade licensed operators may take the place of lower grade licensed or unlicensed personnel.

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6.5 REVIEW AND AUDIT A. OPERATIONS REVIEW COMMITTEE (ORC)

1. FUNCTION The ORC shall function to advise the Pilgrim Station Manager on all matters related to nuclear safety.

2. COMPOSITION The ORC shall be composed of the:

Chairman: Station Manager Member: Management Systems Coordinator Member: Chief Operating Engineer Member: Chief Technical Engineer Member: Chief Maintenance Engineer

3. ALTERNATES Alternate members shall be appointed in writing by the ORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate in an ORC quorum at any one time.

4. MEETING FREQUENCY The ORC shall meet at least once per calendar month and as convened by the ORC Chairman.

5. QUORUM A quorum of the ORC shall consist of the Chairman and two members including alternates.

6. RESPONSIBILITIES The ORC shall be responsible for:

a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) any other proposed procedures or changes thereto that affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

c. Review of all proposed changes to the Technical Specifications.

d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technical Specifications and shall prepare and forward a report covering evaluation and recommenda to tions to prevent recurrence to the Nuclear Operations Manager and the NSRAC Chairman.

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f. Review of facility operations to detect potential safety hazards.

g. Performance of special reviews and investigations and reports thereon as requested by the NSRAC Chairman.

h. Review of the Station Security Plan and implementing procedures and changes to the plan and procedures.

i. Review of the Emergency Plan and implementing procedures and changes to the plan and procedures.

7. AUTHORITY The ORC shall:

a. Recommend to the Pilgrim Station Manager written approval or disapproval of items considered under 6.5.A.6(a) through (d).above.

b. Render determinations in writing with regard to whether or not each item considered under 6.5.A.6(a) through (e) above constitutes an unreviewed safety question.

c. Provide immediate written notification to the Nuclear Operations Manager and the Nuclear Safety Review and Audit Committee of disagreement between the ORC and the Pilgrim Station Manager, however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1 above.

8. RECORDS The ORC shall maintain written minutes of each meeting and copies shall be provided to the Nuclear Operations Manager and NSRAC Chairman.

B. NUCLEAR SAFETY REVIEW AND AUDIT COMMITTEE (NSRAC)

1. FUNCTION The NSRAC shall function to provide independent review and audit of designated activities in the areas of:

1. huclear power plant operations;

2. nuclear engineering;

3. chemistry and radiochemistry;

4. metallurgy;

5. instrumentation and control;

6. radiological safety;

7. mechanical and electrical engineering;

8. quality assurance practices

2. COMPOSITION The NSRAC Chairman and other members shall be appointed by the Vice Presdient - Nuclear, or such other person as he shall designate. Each NSRAC member shall have at least a bachelor's degree in engineering or physical science and a minimum of five years of professional experience.

Experience may be substituted for the degree requirement on the two for one year basis.

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The colle, ve competence of the committee '.1 be maintained as changes-to the membership are made. The-mnembership shall consist of a minimum of five persons.

3. ALTERNATES Alternate members shall be appointed in writing by the Vice President Nuclear or the Chairman to serve on a temporary basis; however, no more than two alternates shall participate in a quorum at any one time.

4. CONSULTANTS Consultants shall be utilized as determined by the NSRAC Chairman to provide expert advice to the NSRAC.

5. MEETING FREQUENCY The NSRAC shall meet at least once per six months.

6. QUORUM 15 A quorum of the NSRAC shall consist of the Chairman and a majority of the NSRAC members including alternates. No more than a minority of the quorum shall have line responsibility for operation of the facility.

7. REVIEW The NSRAC shall review:

a, The safety evaluations for:

(1) Changes to procedures, equipment or systems and (2) Tests or experiments completed under the provision of Section 50.59, 10 CFR to verify that such actions did not constitute an unreviewed safety question.

b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

d. Proposed changes in Technical Specifications or operating license.

e. Violations of applicable statutes, codes, regulations, orders, Technical Specifications, license reqfuirements, or of internal procedures or instructions having nuclear safety significance.

f. Significant operating abnormalities or deviations from normal and expected performance of plant equipment that effect nuclear safety.

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g. All event! hich are required by regulatio. ' Technical Specificat>T6 ns to be reported to the NRC in writing within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

h. Any other matter involving safe operation of the nuclear plant which NSRAC deems appropriate for consideration or which is referred to NSRAC by the onsite operating organization or by other functional organizational units within Boston Edison.

i. Reports and meeting minutes of the Operations Review Committee.

8. AUDITS Audits of facility activities shall be performed under the cognizance of the NSRAC. These audits shall encompass:

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

b. The training and qualifications of the entire facility staff at least oncei per year.

c. The results of all actions required by deficiences occurring 15 in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per six months.

d, The performance of all activities required by the Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per two years.

e, The Emergency Plan and implementing procedures at least once per two years.

f, The Station Security Plan and implementing:procedures at least once per two years.

f. Any other area of facility operation considered appropriate by the NSRAC or the Vice President-Operations and Engineering.

9. AUTHORITY The NSRAC shall report to and advise the Vice President-Nuclear on those areas of responsibility specified in Section 6.5.B.7 and 6.5.B.8.

10. RECORDS Records of NSRAC activities shall be prepared, approved and distributed as indicated below:

a. Minutes of each NSRAC meeting shall be prepared, approved and forwarded to the Vice President-Nuclear, NSRAC members, and others the Chairman may designate, within 14 days following each meeting.

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b. Reports of reviews encompassed by Section 6.5.B.7 e, f, g and h above, shall be prepared, approved and forwarded to the Vice President-Nuclear, with a copy td the Pilgrim Division Head, within 21 days following the completion of the review.

c. Audit reports encompassed by Section 6.5.B,8 above shall be forwarded to the Vice President-Nuclear and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6.6 REPORTABLE OCCURRENCE ACTION The following actions shall be taken in the event of a reportable occurrence:

A. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9.

B. Each Reportable Occurrence Report submitted to the Commission shall be reviewed by the ORC and submitted to the NSRAC and the Nuclear Operations Manager.

6.7 SAFETY LIMIT VIOLATION violated:

The following actions shall be taken in the event a Safety Limit is A. The provisions of 10 CFR 50.36(c)(1)(i) shall be complied with immediately.

B. The Safety Limit Violation shall be reported to the Commission, the Nuclear Operations Manager and to the NSRAC Chairman immediately.

C. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the ORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

D. The Safety Limit Violation Report shall be submitted to the Commission, the NSRAC Chairman and the Nuclear Operations Manager within 14 days of the violation.

6.8 PROCEDURES be established, A. Written procedures and administrative policies shall the requirements and implemented and maintained that meet or exceed N18,7 - 1972 and recommendations of Sections 5.1 and 5.3 of ANSI provided in Appendix "A" of USNRC Regulatory Guide 1,33, except as 6.8.B and 6.8.C below.

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B. Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the ORC and approved by the Station Manager prior to implementation and periodically as set forth in station procedures, C. Temporary changes to procedures of 6.8.A above may be made provided:

1. The intent of the original procedure is not altered.

2. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's license on the unit affected.

3. The change is documented, reviewed subsequently by the ORC, and approved by the Station Manager within 7 days of implementations D. Compliance with the provisions of 6.8.A relating to ANSI N18.7 - 1972 and Regulatory Guide 1.33 is not required until one year from the date of this specification.

6.9 REPORTING REQUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Director of the appropriate Regional Office of Inspection and Enforcement unless otherwise noted.

,A. Routine Reports

.1. Startup Report. A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or character istics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that wo-re re'uire 1 to olbta'-ni satisfactory operation shall also be described. Any addii-ional speci.fic deta;i is reoiireoi in license coud itions based on other coimmitments shall be included in this report.

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Startup reports shal.l be subhlmf-ted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or conmmencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.

2. Annual 0perating Report.- Routine operating reports covering the operation of the unit during the previous calendar year should be submitted prior to March 1 of each year.- The initial report shall be submitted prior to March 1 of the year following initial criticality.

The annual operating reports made by licensees shall provide a comprehensive summary of the operating experience gained during the year, even though some repetition of previously reported information may be involved. References in the annual operating report to previously submitted reports shall be clear.

Each annual operating report shall include:

a.. A narrative summary of operating experience during the report period relating to safe operation of the facility, including safety-related mainte nance not covered in item A.2.b.(5) below.

b. For each outage or forced reduction in power 2/ of over twenty percent of design power level where the reduction extends for greater than four hours:

(1) the proximate cause and the system and major component involved (if the outage or forced reduction in power involved equipment malfunction);

(2) a brief discussion of (or reference to reports of) any reportable occurrences pertain ing to the outage or power reduction; (3) corrective action taken to reduce the proba bility of recurrence, if appropriate; 1/ A si gle submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.

2/ The term "forced reduction in power" is normally defined in the electric power industry as the occurrence of a component failure or other condition which requires that the load on the unit be reduced for corrective action immediately or up to and including the very next weekend, Note that routine preventive maintenance, surveillance and calibration activities requiring power reductions are not covered by this section.

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(4) operating time lost as a result of the outage or powerIjeduction (for scheduled or forced outages,-! use the generator off-line hours; for forced reductions in power, use the approximate duration of operation at reduced power);

(5) a description of major safety-related corrective maintenance performed during the outage or power reduction, including the system and component involved and identifica tion of the critical path activity dictating the length of the outage or power reduction; and (6) a report of any single release of radio activity or radiation exposure specifically associated with the outage which accounts for more than 10% of the allowable annual values.

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated,1 7an rem exposure according to work and job functions,- e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various dulty functions may be estimates based on pocket dosimeter, TLD, or film 'badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

d. Indications of failed fuel resulting from irradiated fuel examinations, including eddy current tests, ultrasonic l*tests, 1- L.46t.J. or visua`

L examiuationls C..T'..Lu.

during the report period.

in the electric power 3/ The term "fOLCeUd ouiage is normally defined other condition industry as the occurrence of a component failure or from service for corrective which requires that the unit be removed the very next weekend.

action immediately or up to and including 2 7

-- This t!bulation s-.,,pplements the raquirements of § 0.40 of 10 CFR Part 20.

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3. Monthly Operating Report. Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the 'Office of Inspection and Enforcemento U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the appropriate Regional Office, to arrive no later than the tenth of each month following the calendar month covered by the report.

3, Reportable Occurrences Reportable occurrences, including corrective actions and measures to prevent reoccurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of o9currence. In cage of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.

1. Prompt Notification With Written Followup. The types of events listed below shall be reported as expeditiously as possible, but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the appropriate Regional Office, or his designate no later than the first working day following the event, with 15 a written followup report within two weeks. The written followup report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide com plete explanation of the circumstances surrounding the event.

a. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function.

Note: Instrument drift discovered as a result of testing neU not be U-3epot+/-Led unduer LZLIs -itew buL Iliay be reportable under items 6.9,B.l.e, 6.9.B.l.f, or 6.9,B.2.a below.

b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition is less conservative than the least conservative aspect of the limiting condition for operation established in the technical specifications.

Note: If specified action is taken when a system, is found to be operating between the most conservative and the least conservative aspects of a limiting condition for operation listed in the technical specifications, the limiting condition for operation is not considered to have been violated and need not be reported under this item, but it may be reportable under item 6.9.B.2.b below.

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c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.

the limits Note: Leakage of valve packing or gaskets within in technical spec for identified leakage set forth this item.

ifications need not be reported under disagreement with

d. Reactivity anomalies, involving balance under the predicted value of reactivity power operation, greater steady state conditions during reactivity than or equal to 1% Ak/k; a calculated less conservative balance indicating a shutdown margin short specifications; than specified in the technical a reactor correspond to term reactivity increases that of less than 5 seconds or, if sub-critical, period of more than 0.5%

an unplanned reactivity insertion criticality.

Ak/k or occurrence of any unplanned which of one or hKore components

e. Failure or malfunction the fulfillment by itself, prevents or could prevent, requirceaents of system(s) used to of the functional in the SAR.

cope with accidents analyzed 15 inadequacy which prevents

f. Personnel error or procedural the fulfillment of the or could prevent, by itself, systems required to cope with functional requirements of accidents analyzed in the SAR.

reduced redundancy Note: For items 6.9.B.l.e and 6.9.B.l.f system function loss of that does not result in a Ireported under this s.ction but may be need not'be aad 6.9.B.2.c bdlow.

reportable under items 6.9.B.2.b man-made events that,

g. Conditions arising from natural or plant shutdown, as a direct result of the event require measures systems, or other protective operation of safety required by technical specifications.

or accident analyses

h. Errors discovered in the transient analyses as described or in the methods used for such for the or in the bases in the safety analysis report have or could have technical specifications that less conservative permitted reactor operation in a manner than assumed in the analyses.

or components that

i. Performance of structures, systems, measures to requires remedial action or corrective conservative than prevent operation in a manner less in the safety analysis assumed in the accidentoanalyses bases; or discovery report or technical specifications during plant life of conditions not specifically con or technical sidered in the safety analysis report remedial action or corrective specifications that require or development of an measures to prevent the existence unsafe condition, 221

Note: This item is intended to provide for reporting of potentially generic problems.

2. Thirty Day Written Reports. The reportable occurrences discussed below shall be the subject of written reports to the Director of the appropriate Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.

a. Reactor protection system or engineered safety feature instrument settings which are found to be less conserv ative than those established by the technical specifica tions but which do not prevent the fulfillmrent of the functional requirements of affected systems.

b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.

Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system configurations as described in items 6.9.B.2.a and 6.9.B.2.b need not be reported except where test results themselves reveal a degraded mode as described above.

C. Observed inadequacies in the implementation of admin istrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.

d. Abnormal degradation of systems other than those specified in item 6.9.B.l.c above desighed to contain radioactive material resulting from the fission process.

Note: Sealed sources or calibration sources are not inclukded under this item. Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical specifications need not be reported under this item.

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C. Unique Reporting Requirements

1. Radioactive Effluent Release Report A report shall be submitted to the Commission within 60 days after January 1 and July 1 of each year specifying the quantity of each of the principal radionuclides released to unrestricted areas in liquid and gaseous effluents during the previous 6 months. The format and content of the report shall be in accordance with Regulatory Guide 1.21 (Revision 1) dated June 1974.

2. Environmental Program Data A report containing the data taken in support of the Pilgrim Station environmental monitoring program shall be submitted within 60 days after January 1 and July 1 of each year. The format and content of the report shall include the following items for each environmental medium:

a, Sample Type (1) Biological (to the extent practicable, list species, life stage, age, weight or size, biological condition, tissue or material sampled, sample weight, etc.).

15 (2) Non-Biological (identify type - soil, air, etc,; list actual area, depth and weight or volume sampled as appropriate, whether grab or continuous sample, number of samples, etc.)

b. Sample Location (also supply map showing locations)

c. Collection Period (continuous samples) or Date of Collection

d. Critical Pathway e, Radionuclide

f. Concentration (OCi/ml or g, specifying wet or standard dry 2 or other appropriate units weight) and/or Deposition (ICi/m of measure). (List conversion factors relating sample activity and volume - depth X cross section - to total area deposition average and maximum values for each medium, etc.)

g. Background Value h, Analytical Method-/

5/ Estimates of the error associated with the measurement of each environmental medium should be repeated.

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i. Comparison of Observed Concentrations. Depositions and Estimated Doses with Predicted Values (based on effluent measurements).

j. Remarks (be sure to explain any unusual measurements or deviations).

3. Special Reports Special reports shall be submitted as indicated in Table 6.9.1.

6.10 RECORD RETENTION A, The following records shall be retained for at least five years:

1. Records of facility operation covering time interval at each power level.

2. Records of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

3. Reportable Occurrence Reports.

4. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

5. Records of reactor tests and experiments.

6. Records of changes made to Operating Procedures.

7. Records of radioactive shipments.

8. Records of sealed source leak tests and results.

9. Records of annual physical inventory of all source material of record.

B. The following records shall be retained for the duration of the Operating License:

1. Record and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

2. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

3. Records of facility radiation and contamination surveys.

4. Records of radiation exposure for all individuals entering radiation control areas.

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TABLE 6.9.1 Reference Submittal Date Area 4.7.C.c Upon completi)

a. Secondary Containment of each test.ý-)

Leak Rate Testing (1) 4.6.G. Five years after

b. In-service Inspection initial fuel loading Evaluation 4.6.A. Five years after

c. Bases for revision of initial fuel loading 180°F Temperature Bases Pressurization Limit 4.8.B. Ten days after the

d. Gross Gaseous Release release occurs.

>0.05 Ci/sec for 48 Hours NOTES: 1. Each integrated leak rate test of the secondary containment shall be the subject of a summary technical report. This report should include data on the wind speed, wind direction, outside and inside temperatues 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.

90 days after completion

2. The report shall be submitted within be based on the commercial of each test. Test periods shall service date as the starting point.

2Z

5. Records of gaseous and liquid radioactive material released to the environs.

6. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.

7. Records of training and qualification for current members of the plant staff.

8. Records of in-service inspections performed pursuant to these Technical Specifications.

9, Records of Quality Assurance activities required by the QA Manual.

10. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

11. Records of meetings of the ORC and the NSRAC.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 RESPIRATORY PROTECTION PROGRAM A. ALLOWANCE Pursuant to 10 CFR 20.103(c)(1) and (3), allowance may be made for the use of respiratory protective equipment in conjunction with activities authorized by the operating license for this facility in determining whether individuals in restricted areas are exposed to concentrations in excess of the limits specified in Appendix B, Table I, Column 1, of 10 CFR 20, subject to the following conditions and limitations:

1. The limits provided in Section 20.103(a) and (b) shall not be exceeded.

2. If the radioactive material is of such form that intake through the skin or other additional route is likely, individual exposures to radioactive material shall be controlled so that the radioactive content of any critical organ from all routes of intake averaged over 7 consecutive days does not exceed that which would result from inhaling such radioactive material for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at the pertinent concentration values provided in Appendix B, Table I, Column 1, 10 CFR 20.

226

3. For radioactive materials designated "Sub" in the "Isotope" column of Appendix B, Table I, Column 1 of 10 CFR 20, the concentration value specified shall be based upon exposure to the material as an external radiation source. Individual exposures to these materials shall be accounted for as part of the limitation on individual dose in 20.101. These materials shall be subject to applicable process and other engineering controls.

B. PROTECTION PROGRAM In all operations in which adequate limitation of the inhalation of radioactive material by the use of process or other engineering controls is impracticable, the licensee may permit an individual in a restricted area to use respiratory protective equipment to limit the inhalation of airborne radioactive material, provided:

1. The limits specified in 6.12.1 above are not exceeded.

2. Respiratory protective equipment is selected and used so that the peak concentrations of airborne radioactive mateiial inhaled by an individual weari-pg the equipment do not exceed the pertinent concentration values specified in Appendix B, Table I, Column 1, of 10 CFR 20. For the purposes of this subparagraph, the concentra tion of radioactive material that is inhaled when respirators are worn may be determined by dividing the ambient airborne concentration by the protection factor specified in Table 6.12-1 for the respirator protective equipment worn. If the intake of radioactivity is later determined by other measurements to have been different than that initially estimated, the latter quantity shall be used in evaluating the exposures.

3. The licensee advises each respiiator user that he may leave the area at any time for relief from respirator use in case of equipment mal function, physical or psychological discomfort, or any other condition that might cause reduction in the protection afforded the wearer.

4. The licensee maintains a respiratory protective program adequate to assure that the requirements above are met and incorporates practices for respiratory protection consistent with those recommended by the American National Standards Institute (ANSI-28.2-1969). Such a program shall include:

a. Air sampling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper selection of respiratory protective equipment.

b. Written procedures to assure proper selection, supervision, and training of personnel using such protective equipment.

c. Written procedures to assure the adequate fitting of respirators; and the testing of respiratory protective equipment for opcrability immediately prior to use.

d. Written procedures for maintenance to assure full effectiveness of respiratory protective equipment, including issuance, cleaning and decontamination, inspection, repair, and storage.

227

e. Written operational and administrative procedures for proper use of respiratory protective equipment including provisions for planned limitations on working times as necessitated by operational conditions.

f. Bioassays and/or whole body counts of individuals (and other surveys, as appropriate) to evaluate individual exposures and to assess protection actually provided.

5. The licensee shall use equipment approved by the U. S. Bureau of Mines under its appropriate Approval Schedules as set forth in Table 6.12-1. Equipment not approved under U. S. Bureau of Mines Approval Schedules shall be used only if the licensee has evaluated the equipment and can demonstrate by testing, or on the basis of reliable test information, that the material and performance charac teristics of the equipment are at least equal to those afforded by U, S. Bureau of Mines approved equipment of the same type, as specified in Table 6.12-1.

C. REVOCATION The specifications of Section 6.12 shall be revoked in their entirety upon adoption of a change to 10 CFR 20, Section 20.103, which would make such provisions unnecessary.

6.13 HIGH RADIATION AREA In lieu of the "control device" or "alarm signal" required by para graph 20.203(c)(2) of 10 CFR 20, an acceptable alternate to controlling access to a high radiation area is as follows:

A. Each High Radiation Area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiation Work Permit and any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose in the area.

B. Each High Radiation Area in which the intensity of radiation is greater than 1000 mrem/hr shall be subject to the provisions of 6.13.A above, and in addition locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Watch Engineer on duty.

228

TABLE 6.12-1 PROTECTION FACTORS FOR RESPIRATORS PROTECTION FACTORS GUIDES TO SELECTION OF EQUIPMENT Particulates BUREAU OF MINES APPROVAL SCHEDULES*

and vapors FOR EQUIPMENT CAPABLE OF PROVIDING AT and gases LEAST EQUIVALENT PROTECTION FACTORS DESCRIPTION MODES except tri- 3 / *or schedule superseding for tium oxide - equipment of type listed (

T. AIR-PURIFYING RESPIRATORS Facepiece, half-mask 4/, 7/ NP 5 21B 30 CFR 14.4 (b) (4)

Facepiece, full7/ NP 100 21B 30 CFR 14.4 (b) (5); 14F 30 CFR IT. ATMOSPHERE-SUPPLYING RESPIRATOR

1. Airline respirator Facepiece, half-mask CF 100 19B 30 CFR 12.2 (c) (2) Type C (i)

Facepiece, full CF 1,000 19B 30 CFR 12.2 (c) (2) Type C (i)

Facepiece, full 7/ D 100 19B 30 CFR 12.2 (c) (2) Type C (ii)

Facepiece, full PD 1,000 19B 30 CFR 12.2 (c) (2) Type C (iii)

Hood CF 5/ 6/

Suit CF 5/ 6/

2. Self-contained breathing apparatus (SCBA)

Facepiece, full 7/ D 100 13E 30 CFR 11.4 (b) (2) (i)

Facepiece, full PD 1,000 13E 30 CFR 11.4 (b) (2) (ii)

Facepiece, full R 100 13E 30 CFR 11.4 (b) (1)

COMBINATION RESPIRATOR Any combination of air-\ Protection factor 19B CFR 12.2 (e) or applicable purifying and atmosphere for type and mode of schedules as listed above supplying respirator operation as listed

___________________________above 1/, 2/, 3/, 4/, S/, 6/, 7/ (These notes are on the following pages)

TABLE 6.12-1 (Continued) 1/ See the following symbols:

CF: continuous flow D: demand NP: negative pressure (i.e., negative phase during inhalation)

PD: pressure demand (i.e., always positive pressure)

R: recirculating (closed circuit) 2/ (a) For purposes of this specification the protection factor is a measure of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radioactive material outside the respiratory protective equipment to that inside the equipment (usually inside the facepiece) under conditions of use. It is applied to the ambient airborne concentration to estimate the concentration inhaled by the wearer according to the following formula:

Concentration Inhaled= Ambient Airborne Concentration Protection Factor (b) The protection factors apply:

(i) only for trained individuals wearing properly fitted respirators used and maintained under supervision in a well-planned respiratory protective program.

(ii) for air-purifying respirators only when high efficiency (above 99.9% removal efficiency by U. S. Bureau of Mines type dioctyl phthalate (DOP) test) particulate filters and/or sorbents appropriate to the hazard are used in atmospheres not deficient in oxygen.

(iii) for atmosphere-supplying respirators only when supplied with adequate respirable air.

3/ Excluding radioactive contaminants that present an absorption or submer sion hazard, For tritium oxide approximately half of the intake occurs by absorption through the skin so that an overall protection factor of not more than approximately 2 is appropriate when atmosphere-supplying respirators are used to protect against tritium oxide. Air-purifying respirators are not recommended for use against tritium oxide. See also footnote 5, below, concerning supplied-air suits and hoods.

4/ Under chin type only. Not recommended for use where it might be possible for the ambient airborne concentration to reach instantaneous values greater than 50 times the pertinent values in Appendix B, Table 1, Column 1 of 10 CFR Part 20.

5/ Appropriate protection factors must be determined taking account of the design of the suit or hood and its permeability to the contaminant under conditions of use. No protection factor greater than 1,000 shall be used except as authorized by the Commission.

230

6/ No approval schedules currently available for this equipment. Equipment must be evaluated by testing or on basis of available test information.

7/ Only for shaven faces.

NOTE 1: Protection factors for respirators, as may be approved by the U. S.

Bureau of Mines according to approval schedules for respirators to protect against airborne radionuclides, may be used to the extent that they do not exceed the protection factors listed in this Table.

The protection factors in this Table may not be appropriate to circumstances where chemical or other respiratory hazards exist in addition to radioactive hazards. The selection and use of respirators for such circumstances should take into account approvals of the U. S. Bureau of Mines in accordance with its applicable schedules.

NOTE 2: Radioactive contaminants for which the concentration values in Appendix B, Table I of this part are based on internal dose due to inhalation may, in addition, present external exposure hazards at higher concentrations. Under such circumstances, limitations on occupancy may have to be governed by external dose limits.

231

SAFETY EVALUATION BY THE OFFICE OF NVCLEAF REACTOR REGULATIOIN SUPPORTINC AMF NENcý,T NO.13 TIC FACILITY LICENSE NO. DPFr-5 CFANGE NO.__I_ TO TFCH*'TCAL SPECIFICATIONS BOSTON EDISON COMPANY PILGRIM NUCLEAR POWEP STATION DOCKET NO. 5O=-2Qa Introduction Py letters dated February 14, 1975, June 2, 1975, and July 29, 1975, Boston Edison Company (BECO) proposed changes to the Technical Specifications appended to Facility Operatinr License No. DPF-35, for the Pilgrim Nuclear Power Station. The proposed changes involve changes to the administrative controls includinF changes to the reportinp requireaents.

Discussion The proposed chanpes would be administrative in nature and are intended to provide uniform license requirements. Areas covered by the proposed uriform specifications include licensee staffing qualifications and ranarement procedures involved vith operating the reactor, reporting requirements, deletion of references to abnormal occurrence, a respiratory protection program, and a change to the Corporate and Station Organization Charts.

!embers of the facility staff should meet the requirements set forth in Guide 1.8, "Personnel Selection and Training" which endorses proposed ANSI N18.1, which was subsequently issued as ANSI N18.1-1971. Provisions operations should be in accord with for independent review of facility Guide 1.33, "Cuality Assurance Program Requirements" which endorses proposed standard ANS 1.2, which was subsequently issued as ANSI !8.7-1972.

In Section P08 of the Energy Reorganization Act of 1974 "abnorral occurrence" is defined as an unscheduled incident cr event which the Commission determines is significant from the standpoint of public health or safety. The term "abnormal occurrence" is reserved for usage by NPC. Regulatory Guide 1.16, "Peportinp of Operating Inforration Appendix A Technical Snecifications", revision 4, enumerates required O AFI E . .. ' .

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reports consistent with Section 208. The proposed change to required reports identifies the reports required of all licensees not already identified by the regulations and those unicue to this facility. The proposal would formalize present reporting and would delete any reports no longer needed for assessment of safety related activities. In addition, a radiation protection program delineates use of respiratory equiprent in the event personnel are to be exposed to concentrations in excess of Part 20 concentrations.

The proposed chanFes to the BECo Corporate and Station Organization Charts reflect reor 'anizations that have taken place within the Nuclear portions of the Corporate structure effective July 1, 1975.

Evaluation The new guidance for reporting operating information does not identify any event as an "abnormal occurrence". The proposed reporting require ments also delete reportinp of information no longer required and duplication of reported Information. The standardization of required reports and desired format for the information will permit more rapid recognition of potential problems.

Identifyinp minimum acceptable qualifications for facility personnel should assure capable performance fro, the facility staff. Other administrative requirerents also restated by the specifications assure uniformity and conformance to the desired features in the review, staffing, and procedures. Incorporating tihe currently accepted respiratory protection program at this time assures that a consistent method of using respira tory equipment is immediately available whenever needed. Similar changes are being approved for all power reactor licensees, so all licensees will have the same requirements presented in a uniforr manner.

UFCo has proposed to delete the requirenent that the Station Manager hold a Senior Operator License. Since the Station Manager is rot required by ANSI N18.1 - 1971 to hold a Senior Operator License and the Pilgrim Station Manager satisfies the reQuirements of ANSI F18.1 - 1071 Section 4.2.1 relating to required training:and,-experience-for~the~position,..

this change-i* acceptable.

During our review of the proposed chanres, we found that certain modi fications to the proposal were necessary to have conformance wi*t the desired Pegulatory position. These chanpes were discussed with [lia* o staff and have been incorporated into the proposal. p'r ;ýrýd.) J

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- 3-We have concluded that the proposal as modified improves the licensee's program for evaluating plant performance and the reportinF of the operating information needed by the Commission to assess safety related activities and is acceptable. The facility staff aualifications and training program conforn to Guide 1.8 and therefore are acceptable.

consistent The administrative procedures and facility review and audit are The modified reporting program is with Guide 1.33 and are acceptable.

Regulatory Guide 1.16, "Reporting consistent with the guidance providedby of Operating Information - Appendix A Technical Specifications",

the Pevision M. The administrative controls are consistent with requirements being incorporated in Technical Specifications for new licensed fscilities.

The ERCo Corporate and Station Manager changes allow improved communication the between the Corporate manarerent and the Station IManager and augments in the areas of station organization by the addition of new positions waste management, nuclear training, outage planning, health physics change supervision, performance engineering and material contro0- These are acceptable to the staff.

Conclusion that:

We have concluded, based on the considerations discussed above, in the (1) because the change does not involve a significant increase and does probability or consequences of accidents previously considered the change does not involve a significant decrease in a safety r,-arin, reasonable not involve a significant hazards consideration, (2) there is assurance that the health and safety of the public will not be endangered will be by operation in the proposed manner, and (3) such activities and the issuance conducted in compliance with the Commission's regulations and security of this amendment will not be inimical to the common defense or to the health and safety of the public.

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UNITED STATES NUCLEAR REGULATORY COM1ISSION DOCKET NO. 50-293 BOSTON EDISON COMPANY NOTICE OF ISSUANCE OF AMENDMENT TO FACILITY OPERATING LICENSE Notice is hereby given that the U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 13 to Facility Operating License No. DPR-35 issued to the Boston Edison Company (the licensee) which revised Technical Specifications for operation of the Pilgrim Nuclear Power Station (the facility) located in Plymouth County, Massachusetts. The amendment is effective 30 days after the date of its issuance.

The amendment revises the administrative controls section of the Pilgrim Nuclear Power Station Technical Specifications to provide uniforw staffing qualifications, management procedures and reporting requirements.

The application 6or the amendment complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commiss6on's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendment. Prior public notice of this amendment is not required since the amendment does not involve a significant hazards consideration.

For further details with respect to this action, see (1) the application for amendment dated February 14, 1975, (2) Amendment No. 13 to License No. DPR-35, with Change No. 15, and (3) the Commission's concurrently issued related Safety Evaluation. All of these items are oFC E ) ................. .......

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available for public inspection at the Commission's Public Document Room, 1717 H Street, N. W., Washington, D. C. and at the Plymouth Public Library on North Street in Plymouth, Massachusetts 02360. A copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, Attention: Director, Division of Reactor Licensing.

Dated at Bethesda, Maryland, this ý) t\Jq).4e yv..A&t f FOR THE NUCLEAR REGULATORY COtIISSION Original Signed by:',

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ACTION ROUTING AND TRANSMITTAL SLIP I TO (Name, office symbol or location) I IITIALS CIRCULA"E DATE COORDINATION OELD - f/concurrences 2 INITIALS FILE DATE INFORMATION DLZiemann - f/signatures 3 INITIALS NOTE AND SE*lJIRN Reba - for final checks DATE PERCON V ER SAT ION 4 INITIALS SEE ME DATE SIGNATURE REMARKS Attached for your concurrence are five packages (Dresde Station, Quad Cities Station, Cooper, Pilgrim and Calvert Cliffs) of nine from ORB 2 which incorporate standard reporting requirement sections into the Appendix A Technical Specifications. One package, Pilg also revises the entire administrative controls sectic.n It is requested that, in the interest of review con sistency, these packages (and the 4 future reporting requirements packages) be assigned to one OELD re viewer.

Questions may be directed to the PM for the particular case to Mike (Exts.or 7403, Fletcher,. ,coordinator , for reporting,_-i 7450),* "

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Do NOT use this form as a RECORD of approvals, concurrences,'~e~

disapprovals, clearances, and similar actions FROM (Name, office symboor location)

DLZiemann 'ýT~

4 5 9 OPTIONAL FORM 41 1--8 1 4-1 552-103 GPO 5041-101 AUGUST 1967 GSA FPMR ( 41CFR) 100-11.206 StA) eto - S' ..

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