ML101200158

Forwards Application for Amend to License.Certificate of Svc Encl

ML101200158
Person / Time
Site:	Indian Point
Issue date:	01/10/1979
From:	William Cahill Consolidated Edison Co of New York
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML101200160	List: ML101200158 ML19256A840
References
NUDOCS 7901160222
Download: ML101200158 (27)

Text

William J. Cahill, Jr.

Vice PresidAnt Consolidated Edison Company of New York, Inc.

4 Irving Place. New York, N Y 10003 Telephone (21 2) 460-3819 January 10, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Re: Consolidated Edison Caripany of New York, Inc.

(Indian Point Station, Unit No. 2)

Docket No. 50-247

Dear Mr. Denton:

Consolidated Edison Ccinpany of New York, Ir. (Con Edison) transmits herewith three (3) signed originals and forty (40) copies of a document entitled "Application for Amendment to Operating License," sworn to on January 10, 1979. This Application requests an amendment to the Technical Specifications for Indian Point Unit No. 2. The proposed Technical Specification changes, contained in Attachment A to this Application, would establish limiting conditions for operation (IMs)and surveillance requirements for a gas turbine generator to be utilized as a backup contingency power source as part of the overall Indian Point Unit No. 2 fire protection program. This change was requested by the Regulatory Staff as a result of its review of the fire protection program. A Safety Evaluation of the proposed changes is presented in Attachment B to this Application.

We have been advised by the Regulatory Staff that a filing fee per 10 CFR 170.22 is not required for this Application. It is our understanding that since this change has resulted from the Staff's review of the Indian Point Unit No. 2 fire protection program which was initiated prior to the effective date of the regulation, no such fee is required.

A Certificate of Service is enclosed.

Very truly yours, William J. Cahill, Jr.

Enclosures Vice President cc: Hon. George V. Begany \-

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UNITED STATES OF AMERICA NUCLEAR REGUIATORY COMMISSION In the Matter of )

CONSOLIDATED EDISON COMPANY ) Docket No. 50-247 OF NEW YORK, INC.

(Indian Point Station, )

Unit No. 2) )

CERTIFICATE OF SERVICE I certify that I have, this IA day of January, 1979, served the foregoing document entitled "Application for Amendment to Operating License," sworn to on January 10, 1979, by mailing two (2) copies thereof, first class postage prepaid and properly addressed to the following person:

Hon. George V. Begany Mayor, Village of Buchanan 188 Westchester Avenue Buchanan, New York 10511

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In-the Matter of )

)

CONSOLIDATED EDISON COMPANY ) Docket Nos. 50-3 and 50-247 OF NEW YORK, INC. )

(Indian Point Station, )

Unit Nos. 1 and 2) )

STATE OF NEW YORK )

ss:

COUNTY OF NEW YORK )

AFFIDAVIT OF SERVICE Louis F. Liberatori, Jr., being duly sworn, states:

That he is an Engineer employed by Consolidated Edison Company of New York, Inc., and that he has served the foregoing two (2) documents entitled "Application for Amendment to Operating License," sworn to on February 21, 1979, by mailing two (2) copies thereof, first class postage prepaid and properly addressed to the following person:

Hon. George V. Begany Mayor, Village of Buchanan 188 Westchester Avenue Buchanan, New York 10511

- f~~1 Louis F. Liberatori, Jr Sworfi to before me this day of , 1979 Notary Publi'c

0 ATTACHMENT A Technical Specification Page Revisions Consolidated Edison Company of New York, Inc.

Indian Point Unit No. 1 Docket No. 50-3 February, 1979

3.0 Administrative and Procedural Safeguards 3XY Organization

-3.1.1 The organization for facility management and technical support shall be as shown in Figure 3.1.

3.1.2 The Facility Organization shall be as shown in Figure 3.2.

The Support Facilities Supervisor is responsible for operations at the Unit No. 1 facility.

3.1.3 The Department Manager shall be responsible for overall facil ity operation and shall delegate in writing the succession to this responsibility during his absence.

3.1.4 The operation of the facility, the operating organization, the procedures for operation, and modifications to the facility shall be subject to review by the Station Nuclear Safety Committee. The committee shall report to the Department Manager.

3.1.5 The Nuclear Facilities Safety Committee shall function or pro vide independent review and audit of designated activities in areas of nuclear engineering, chemistry, radiochemistry, metallurgy and non-destructive testing, instrumentation and control, radiological safety, mechanical and electrical engineering, administrative controls and quality assurance practices, and radiological environmental effects.

3.1.6 All fuel handling shall be under the direct supervision of a licensed operator.

• 3.2 Operating Instructions and Procedures 3.2.1 No fuel will be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the Nu clear Regulatory Commission.

3.2.2 Detailed written instructions setting forth procedures used in connection with the operation and maintenance of the nuclear power plant shall conform to the Technical Specifications.

3.2.3 Operation and maintenance of equipment related to safety when there is no fuel in the reactor shall be in accordance with written instructions.

3.2.6 Radiation control procedures shall be maintained and made avail able to all station personnel. Station operation shall adhere to these procedures.

These proc~dures show permissible radiation exposure, and shall be consistent with the requirements of 10CFR2O. The radiation protection program shall be organized to meet-the requirements of 10CFR2O.

• Licensed operator for IP-l or IP-2.

Amendment No.

I NUCLEAR FACILITIES SAFETY COMMITTEE OFFICER IN CHARGE OF POWER GENERATION OFFSITE ONSITE DIRECTOR DIRECTOR FigureNUCEA 3.1E FaiiyMngmnSn ehia upr rAnizTio Amendment No.

0 (CCR) - Central Control Room Position (S) - Continuous Shift Coverage Figure 3.2 Facility Organization Amendment No.

S. S

-8 6.5 The annual radiation exposure reports shall provide 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 man rem exposure according to work and job functions, 2 e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special mainte nance (describe maintenance), waste processing, and refueling. The dose assign ment to various duty 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 as signed to specific major work functions.

Special Reports 6.6 Reports of major safety-related corrective maintenance shall be submitted to the Director, Office of Management Information and Program Control, with 40 copies to the Office of Inspection and Enforcement, U. S. Nuclear Regulatory Commission, Washington, D. C. 20555, no later than 6 months following comple tion of such maintenance.

6.7 Each such report shall include a description of any major safety-related corrective maintenance performed including the system and component involved.

Reportable Occurrences 6.8 The reportable occurrences of specifications 6.8.1 and 6.8.2 below, in cluding corrective actions and measures to prevent recurrence, shall be re ported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.

Prompt Notification with Written Followup Report DS 6.8.1 The types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification by telephone and confirmed by telegraph, mailgram, or facsimile 2

This tabulation supplements the requirements of §20.407 of 10 CFR Part 20.

Amendment No.

transmission to the Director of the Region I Office of Inspection and Enforce ment or his designate, no later than the first working day following the event, with 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 complete 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 speci fied as the limiting safety system setting in the technical specifi cations or failure to complete the required protective function.

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

c. Abnormal degradation discovered in fuel cladding, reactor coolant 3

pressure boundary, or primary containment.

d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power oper ation greater than or equal to 1% Ak/k; a calculated reactivity bal ance indicating a shutdown margin less conservative than specified in the technical specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if subcriti cal, an unplanned reactivity insertion of more than 0.5% Ak/k; or occurrence of any unplanned criticality.

e. Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional require

-ments of system(s) used to cope with accidents analyzed in the SAR.

3Leakage of packing, gaskets, mechanical joints or seal welds within the limits for identified leakage set forth in technical specifications need not be re ported under this item.

Amendment No.

-10

f. Personnel error or procedural inadequacy which prevents or could

- prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR.

g. Conditions arising from natural or man-made events that, as a direct result of the event require plant shutdown, operation of safety sys tems, or other protective measures required by technical specifi cations.

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

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

Thirty Day Written Reports 6.8.2 The types of events listed below shall be the subject of written reports to the Director of the Region I Office of Inspection and Enforcement 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 pro vided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circum stances surrounding the event.

a. Reactor protection system or engineered safety feature instrument settings which are found to be less conservative than those estab

• lished by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.4 Amendment No.

b. Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a 4

limiting condition for operation.

c. Observed inadequacies in the implementation of administrative or pro cedural 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 6.8.1.c above designed to contain radioactive material resulting from the 5

fission process.

6.9 Any references to the term "Safety Analysis Report", "SAR" or "FSAR" for Indian Point Station, Unit No. 1, shall be deemed to refer, as appropriate, to the following exhibits which are a part of the application: F-4 (Rev.-3),

F-6 (Rev.-2), F-7 (Rev.-l), G-3 (Rev.-2), H-14 (Rev.-2), K-4, K-4A, K-4B, K-5 (Rev.-l, but not including Sections 2.1.2 through 2.3.7.4, Section 4, Figures 2-1 through 2-9, Figure 3-17, Figures 4-1 through 4-12, and Appendix A), K-5A1, K-5A10, K-5A11, K-5AIIA, K-5A12, K-5AI3, K-5A14, as amended, K-5A15, K-16, and the document entitled "Final Hazards Summary Report for the Consolidated Edison Indian Point Reactor Core B", as amended.

4Routine serveillance testing, instrument calibration, or preventive mainte nance which require system configurations as described need not be reported except where test results themselves reveal a degraded mode as described.

5 Sealed sources or calibration sources are not included 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.

Amendment No.

ATTACHMENT B Safety Evaluation Consolidated Edison Company of New York, Inc.

Indian Point Unit No. 1 Docket No. 50-3 February, 1979

Safety Evaluation The proposed technical specification revisions, contained in Attachment A to this Application, would effect certain administrative and editorial changes.

The proposed revisions to section 3.1 would require that: (a) the Department Manager rather than the Plant Manager be responsible for overall Unit No. 1 facility operation, (b) the Station Nuclear Safety Committee report to the Department Manager rather than the Plant Manager, and (c) the Nuclear Facili ties Safety Committee report to and advise the President of the Company rather than the Senior Company Officer in Charge of Power Supply. The re vised Figures 3.1 and 3.2 would reflect these and additional pending organi zational changes.

The proposed changes to sections 6.6, 6.7 and 6.8 would replace the monthly operating report with a requirement for a special report. The special report would be submitted, when appropriate, to describe any major safety-related corrective maintenance performed on Unit No. 1. The remaining reporting requirements for the monthly operating report, which are operation-related, would no longer be in effect for Unit No. 1. Since the unit is being main tained in the shutdown and defueled condition, no meaningful information can be provided in response to the operation-related requirements. Accordingly, major safety-related corrective maintenance would continue to be reported for Unit No. 1 while operation-related reporting requirements, which are no longer applicable to Unit No. 1, would be deleted.

The proposed changes do not in any way alter the safety analyses performed for Indian Point Unit No. 1. The proposed changes have been reviewed by the Station Nuclear Safety Committee and the Consolidated Edison Nuclear Facilities Safety Committee. Both committees concur that these changes do not represent a significant hazards consideration and will not cause any change in the types or increage in the amounts of effluents or any change in the authorized power level of the facility.

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the-Matter of )

)

CONSOLIDATED EDISON COMPANY Docket No. 50-247 OF NEW YORK, INC. )

(Indian Point Station, )

Unit No. 2) )

APPLICATION FOR AMENDMENT TO OPERATING LICENSE Pursuant to Section 50.90 of the Regulations of the Nuclear Regulatory Commission, Consolidated Edison Company of New York, Inc. ("Consolidated Edison"), as holder of Facility Operating License No. DPR-26, hereby applies for an amendment to the Indian Point Unit No. 2 Technical Specifications contained Lzn A%,ppendix A to that license. Specifically, Consolidated Edison requests revisions to sections 3.1, 6.1, 6.2, 6.5 and 6.9.1 to reflect pending organizational changes and effect editorial clarifications.

The proposed Technical Specification changes consist of the specific revisions set forth in Attachment A to this Application.

A Safety Evaluation of the proposed changes is set forth in At tachment B to this Application. This evaluation demonstrates

-2 that proposed changes do not represent a significant hazards consideration and will not cause any change in the types or in crease in the amounts of effluents or any change in the authorized power level of the facility.

CONSOLIDATED EDISON COMPANY OF NEW YORK, INC.

By:

William J. l, Jr.4 Vice President Subscribed and sworn to before me this,;/Mf day of February, 1979.

otA-ry "Public ANCELA RODERT Notar.y Public, S 1e -Of Nc.'; 'crk Q::-1Uid in Qi:;:ns Co-ry Com:,ilss npr;-rC ,1

ATTACHMENT A Technical Specification Page Revisions Consolidated Edison Company of New York, Inc.

Indian Point Unit No. 2 Docket No. 50-247 February, 1979

Basis:

Water inventory balances, monitoring equipment, radioactive tracing, boric acid crystalline deposits, and physical inspections can disclose reactor coolant leaks. Any leak of radioactive fluid, whether from the reactor coolant system primary boundary or not can be a serious problem with respect to in-plant radioactivity contamination and cleanup or it could develop into a still more serious problem; and therefore, first indications of such leakage will be followed up as soon as practicable.

Although some leak rates on the order of GPM may be tolerable from a dose point of view, especially if they are to closed systems, it must be recognized that leaks in the order of drops per minute through any of the walls of the primary system could be indicative of materials failure such as by stress corrosion cracking. If depressurization, isolation and/or other safety measures are not taken promptly, these small leaks could develop into much larger leaks, possibly into a gross pipe rupture. Therefore, the nature of the leak, as well as the magnitude of the leakage must be considered in the safety evaluation.

When the source of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. This evaluation will be per formed by the Plant Operating Staff and will be documented in writing and approved by either the Department Manager or his designated alternate. Under these conditions, an allowable primary system leakage rate of 10 gpm within the capacity of one-charging pump and makeup would be available even under the loss of off-site power condition.

If leakage is to the containment, it may be identified by one or more of the following methods:

a. The containment air particulate monitor is sensitive to low leak rates.

The rates of reactor coolant leakage to which the instrument Amendment No. 3.1-18 311

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 _The Department Manager shall be responsible for overall facilityv operation and shall delegate in writing the succession to this responsibility during his absence.

6.2 ORGANIZATION FACILITY MANAGEMENT AND TECHNICAL SUPPORT 6.2.1 The organization for facility management and technical support shall be as shown on Figure 6.2-1.

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

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

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

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

e. ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Oper ator or Senior Reactor Operator Limited to Fuel Handling. This individual shall have no other concurrent responsibilities during this operation.

f. A Fire Brigade of at least five members shall be maintained on site at all times. This excludes four members of the minimum shift crew necessary for safe shutdown of the plant and any personnel required for other essential functions during a fire emergency.

During periods of cold shutdown, the Fire Brigade will exclude two members of the minimum shift crew.

6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions.

Amendment No.

PRNSIDTE NUCLEAR POWERAT RA GEN I CEPRTM N COMMITTEE ENGINEERING TRAI IN ASSURANC MANGE DIRCTO DIECOWR GEENGTINEE Fiue62-FaiitFaaemn nSecnclSupr rgnzto Amendment No.

0 MANAGER NUCLEAR POWER GENERATION DEPARTMENT QUALITY PLANT ASSURANCE MANAGER ENGINEER

& staff (S)

(CCR)

(CCR) - Central Control Room Position (S) - Continuous Shift Coverage Figure 6.2-2 Facility Organization Amendment No.

0 0 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained Under the direction of the Nuclear Training Director and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18el-1971 and Appendix "A" of 10 CFR Part 55.

6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Nuclear Training Director and shall meet or exceed the require ments of Section 27 of the NFPA Code-1976 with the exception of the training program schedule.

REVIEW AND AUDIT 6.5.1 STATION NUCLEAR SAFETY COMMITTEE (SNSC)

FUNCTION 6.5.1.1 The Station Nuclear Safety Committee shall function to advise the Department Manager on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The Station Nuclear Safety Committee shall be composed as follows:

Chairman: Plant Manager Member: Technical Engineering Director Member: Quality Assurance Engineer Member: Chief Operations Engineer Member: Security Supervisor Member: Test and Performance Engineer Member: Instrument and Control Engineer Member: Maintenance Engineer Member: Chemistry and Radiation Safety Director Member: Reactor Engineer Member: Department Manager Member: Refueling Engineer ALTERNATES 6.5.1.3 Alternate members shall be appointed in writing by the SNSC Chairman to serve on a temporary basis.

MEETING FREQUENCY 6.5.1.4 The SNSC shall meet at least once per calendar month and as convened by the SNSC Chairman.

Amendment No. , 6-5

QUORUM 6.5.1.*5 A quorum of the SNSC shall consist of the Chairman or Vice Chairman and five members including no more than two alternates.

RESPONSIBILITIES 6.5.1.6 The Station Nuclear Safety Committee shall be responsible for:

a. Review of 1) all procedures required by Specification 6.8 and changes thereto, and 2) any other proposed procedures or changes thereto as determined by the Department Manager to 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 preparation and forwarding of a report covering evaluation and recommendations to prevent recurrence to the Manager, Nuclear Power Generation Department and to the Chairman of the Nuclear Facilities Safety Committee.

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

g. Performance of special reviews and investigations and the issuance of reports thereon as required by the Chairman of the Nuclear Facili ties Safety Committee.

h. Review of the Plant Security Plan and implementing procedures and submission of recommended changes to the Chairman of the Nuclear Facilities Safety Committee.

i. Review of the Emergency Plan and implementing procedures and sub mission of recommended changes to the Chairman of the Nuclear Facili ties Safety Committee.

AUTHORITY 6.5.1.7 -The Station Nuclear Safety Committee shall:

a-.- - Recommend to the Department Manager, in writing, approval or disap proval of items considered under 6.5.1.6(a) through (d) above.

b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.

Amendment No.

AUTHORITY (Continued)

I

c. Provide immediate written notification to the Chairman, Nuclear Facilities Safety Committee of disagreement between the recommenda tions of the SNSC and the actions contemplated by the Department Manager. However, the course of action determined by the Department Manager pursuant to 6.1.1 above shall be followed.

RECORDS 6.5.1.8 The Station Nuclear Safety Committee shall maintain written minutes of each meeting and copies shall be provided to, as a minimum, the Manager, Nuclear Power Generation Department and the Chairman, Nuclear Facilities Safety Committee.

6.5.2 NUCLEAR FACILITIES SAFETY COMMITTEE (NFSC)

FUNCTI ON 6.5.2.1 The Nuclear Facilities Safety Committee shall function to provide independent review and audit of designated activities in the areas of:

a. reactor operations

b. nuclear engineering C. chemistry and radiochemistry

d. metallurgy and non-destructive testing

e. instrumentation and control

f. radiological safety

g. mechanical and electrical engineering

h. administrative controls and quality assurance practices

1. radiological environmental effects

j. other appropriate fields associated with the unique characteristics

-of the nuclear power plant Amendment No.6- 6-7

h. The Facility Fire Protection Program and implementing procedures at least once per 24 months.

1. A fire protection and loss prevention inspection and audit shall be performed utilizing either qualified offsite licensee personnel or an outside fire protection firm at least once per 12 months.

j. An inspection and audit of the fire protection and loss prevention program shall be performed by an outside qualified fire consultant at least once per 36 months.

k. The environmental surveillance program pertaining to radiological matters and implementing procedures at least once per 12 months.

1. Any other area of facility operation considered appropriate by the NFSC or the President of the Company.

AUTHORITY 6.5.2.9 The NFSC shall report to and advise the President of the Company on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.

RECORDS 6.5.2.10 Records of NFSG activities shall be prepared, approved and distri buted as indicated below:

a. Minutes of each NFSC meeting shall be prepared, approved and for warded to the President and to Senior Company Officers concerned with nuclear facilities within 14 days following each meeting.

b. Reports of reviews encompassed by Section 6.5.2.7 e, f, g and h above, shall be prepared, approved and forwarded to the President and to Senior Company Officers concerned with nuclear facilities within 14I days following completion of the review.

c. -Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Senior Company Officers concerned with nuclear facilities and to the management positions responsible for the areas audited within 30 days after completion of the audit.

Amendment No. 6-11 61

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.

Monthly Operating Report 6.9.1.5 Routine reports of operating statistics., operating and shutdown ex perience and major safety-related corrective maintenance shall be submitted on a monthly basis to the Director, Office of Management Information & Program Control, with 40 copies to the Office of Inspection and Enforcement, U. S.

Nuclear Regulatory Commission, Washington, D.C. 20555, no later than 15 days following the calendar month covered by the report.

6.9.1.6 Each monthly operating report shall include:

a. A tabulation of plant operating data and statistics.

b. A narrative summary of operating experience during the report period C.

relating to safe operation of the facility, including major safety related corrective maintenance not covered in 6.9.1.6.c.5 below.

3 For each outage or forced reduction in power4 of over twenty percent I

of rated power 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 pertaining to the outage or power reduction;

3. Corrective action taken to reduce the probability of recurrence, if appropriate;

4. operating time lost as a result of the outage or power reduction (for scheduled or forced outages, 5 use the generator off-line 3Any safety-related maintenance information not available for inclusion in the monthly operating report for a report period shall be included in a subsequent monthly operating report not later than 6 months following completion of such maintenance.

1 The term 'Iforced reduction in power" is defined as the occurrence of a com ponent 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 calibra tion activities requiring power reductions are not covered by this section.

5The term "forced outage" is defined as the occurrence of a component failure or other condition which requires that the unit be removed from service for corrective action immediately or up to and including the very next weekend.

Amendment No. 6-15 61

0 ATTACHMIENT B Safety Evaluation Consolidated Edison Company of New York, Inc.

Indian Point Unit No. 2 Docket No. 50-247 February, 1979

Safety Evaluation The proposed technical specification revisions, contained in Attachment A to this Application, would effect certain administrative and editorial changes.

The proposed revisions to sections 3.1, 6.1, 6.2 and 6.5 would require that:

(a) the Department Manager rather than the Plant Manager be responsible for overall Unit No. 2 facility operation, (b) the Station Nuclear Safety Committee report to the Department Manager rather than the Plant Manager, and (c) the Nuclear Facilities Safety Committee report to and advise the President of the Company rather than the Senior Company Officer in Charge of Power Supply. The revised Figures 6.2-1 and 6.2-2 would reflect these and additional pending organizational changes. The proposed changes to section 6.9.1 would clarify the reporting requirement that major safety-related corrective maintenance be included in the monthly operating reports.

The proposed changes do not in any way alter the safety analyses performed for Indian Point Unit No. 2. The proposed changes have been reviewed by the Station Nuclear Safety Committee and the Consolidated Edison Nuclear Facili ties Safety Committee. Both committees concur that these changes do not represent a significant hazards consideration and will not cause any change in the types or increase in the amounts of effluents or any change in the authorized power level of the facility.

L__