ML19210A609

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Tech Spec Change Request 5,Amend 3 Supporting Licensee Request to Change DPR-50,App a Re AEC to NRC Name Change, Util Organization Changes,Section 6 Consistency & Reg Guide 1.16 Revision.Certificate of Svc Encl
ML19210A609
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 10/03/1975
From: Arnold R
METROPOLITAN EDISON CO.
To:
Shared Package
ML19210A607 List:
References
RTR-REGGD-01.016, RTR-REGGD-1.016 NUDOCS 7910300645
Download: ML19210A609 (30)
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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY and PENNSYLVANIA ELECTRIC COMPAhT THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 5 Amendment No. 3 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1.

METROPOLITAN EDISON COMPANY By  ! /

Vice President-Generation Sworn and subscribed to me this 3rd day of October , 1975

$' Nl nu- ~

Notary Public RITA ?.1. POWERS tiotarf M.c. ?/ah;enberg Tag., Berks Ca.

My Ccmm:ss on Expires Septemter 30,1978 1492 123 2910300 (9s

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~:,, ~,._,m MElHOPOLII AN EDISON COMPANY POST OFFICE BOX 542 READING, PENNSYLVANI A 19603 TELEPHONE 215 - 929-3601 October 3, 1975 GQL 1579 Mr. Charles P. Hoy, Chairman Board of County Commissioners of Dauphin County Dauphin County Cor,: House P.O. Box 1295 Harrisburg, Pennsylvania 17120

Dear Mr. Hoy:

Enclosed please find one copy of Technical Specification Change Request No. 5 Amendment No. 3 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1.

This request was filed with the U. S. Nuclear Regulatory Commission on October 3, 1975.

Very truly yours,

/s/ R. C. Arnold R. C. Arnold Vice President RCA:CWS:tas File: 20.1.1 / 7.7.4.3.3.1 1492 124

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METHOPOLil AN EDISON COMPANY POST OFFICE BOX 542 READING. PENNSYLVANI A 19603 TELEPHONE 215 - 929 3601 October 3,1975 GQL 1579 Mr. Weldon B. Archart, Chairman Board of Supervisors of Londonderry Township R. D. #1, Geyers Church Road Middletown, Penasylvania 17057

Dear Mr. Archart:

Enclosed please find one copy of Technical Specification Change Request No. 5 Amendment No. 3 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1.

This request was filed with the U. S. Nuclear Regulatory Commission on October 3, 1975.

Very truly yours,

/s/ R. C. Arnold R. C. Arnold Vice President RCA:CWS:tas File: 20.1.1 / 7.7.4.3.3.1 (492 125

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 5 Amendment No. 3 to Appendix A of the Operating License for Thre e Mile Island Nuclear Stacion Unit 1, dated October 3,1975, and filed with the U. S. Nuclear Regulatory Commission on October 3,1975, has this 3rd day of October been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:

Mr. Weldon B. Archart, Chairman Mr. Charles P. Hoy, Chainnan Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 P.O. Box 1295 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY By /s/ R. C. Arnold Vice President-Generation 1492 126

v' Metropolitan Edison Company (Met-Ed)

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Docket #50-289 Operating License #DPR-50 Technical Specification Change Request #5 Amendment 3 Licensee requests that the enclosed pages be substituted for the existing pages comprising Technical Specificatic,n Section 6 except that the enclosed pages 1-6, 1-7, and 1-8 should replace page 1-6 of the existing Technical Specifications Section 1.

Reason for Amendment 3 This amendment to change request 5 is necessary in order to:

a) incorporate the AEC to NRC name change; b) reflect recent Met-Ed administrative and organizational changes; c) make section 6 consistent with other sections of the Technical Specifications; d) and incorporate Regulatory Guide 1.16 revision 4 August 1975 into the Technical Specifications.

Note that items 2.a(8) and 2.a(9) of Reg. Guide 1.16 are changed to 30 day reporting requirements since the Licensee does not believe that 14 days is sufficient to compile an adequate report on items of this nature.

Safety Analysis Justifying Amendment This proposed amendment does not involve any unreviewed safety questions, in that a) it does not alter those responsibilities that need to be S1 filled to ensure safe operation of the unit but only serves to make Section 6 stated titles of management personnel consistent with the most recent orgaairational changes and serves to make Section 6 consistent with other portions of the existing Technical Specifications; b) it utilizes an approved NRC guideline to establish reporting requirements for Appendix A of the Technical Specifications.

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1.8 14-DAY REPORTABLE OCCURRENCE A 14-day reportable occurrence is defined as any one of the folloving conditions

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 parsmuter reaches the setpoint specified as the limiting safety-system setting in the technical specifications or failure to complete the required protective function.

Note: Instrument drif t discovered as a result of testing need not be reported under this item but may be reportable under items 1.8e,1.8f, or 1.9c.

b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation 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 techneial 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 1.9d.

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

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

d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions greater than or equal to 1% Ak/k; a calculated reactivity balan:e 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 sub-critical, 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 requirements of systems (s) used to cope with accidents analyzed in the Final Safety Analysis Report (FSAR).

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f. Perscanel error or procedural inadequacy which prevents or could prevant, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the FSAR.

Note: For items 1.8e and 1.8f reduced redundancy that does not result in loss of system function need not be reported under this section but may be reportable under items 1.9d and/or 1.9e.

g. Conditions arising from natural or man-made events that, as a direct result of the event, affect or threacen to affect the safe operation of the Unit, and require plant shutdown, operation of safety systems, or other protective measures required by technical specifications.

The following are examples:

(1) Threatened civil disturbances requiring plant shutdown.

(2) Significant damage to the facility caused by fire, flood, earthquake, or other similar occurrences.

1.9 30-Day Reportable Occurrence A 30-day reportable occurrence is defined as any one of the following conditions.

a. 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 boses for the technical specifications that have or could have permitted reactor operation in a manner less conserva-tive than assumed in the analyses,
b. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than that assumed in the accident analysis 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.
c. Reactor protection system ar engineered safety feature instrument settings which are found to be less conservative than those established by the technical specifications but which do not prevent the fulfillment of the functionai requirements of affected systems.
d. Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.

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e. Abnormal degradation of systems other than those specified in item 1.8c above designed to contain radioactive material resulting from the fission process. For example, a through-wall leak in a liquid waste storage tank.

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

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6.0 ADMINISTRATIVE CONTROLS 6_.1 RESPONSIBILITY 6.1.1.a. The Unit Superintendent shall be responsible for the overall safety of plant operations and shall ensure that:

1. All proposed changes to procedures, equipmenc, or systems are evaluated to determine if they constitute a change to the facility ,r procedures as described in the Final Safety Analysis Report.
2. All proposed changes to procedures, equipment, or systems which constitute a change of the f acility or procedures as described in the Final Safety Analysis Report are evaluated to determine that they do not involve an unreviewed safety question as defined in paragraph 50.59 (c), Part 50, Title 10, Code of Federal Regulations.
3. All proposed tests and experiments, not described in the Final Safety Analysis Report, are evaluated to determine that they do not involve an unreviewed safety question as defined in paragraph 50.59 (c), Part 50, Title 10, Code of Federal Regulations.
4. Records are kept: a) of changes to procedures, equipment or systems completed under the provisions of paragraph 50.59 (b),

Part 50, Title 10, Code of Federal Regulations; b) of tests and experiments conducted in accordance with those provisions; and c) of the written safety evaluation used as a basis for determining that such changes, tests and experiments do not involve an unreviewed safety question.

5. Copies of evaluations conducted pursuant to 6.1.1.a.2 and 6.1.1.a.3 above are forwarded to the Plant Operations Review Committee, the Manager-Generation Engineering, and the General Office Review Boara Secretary.
b. The Unit Superintendent shall have the authority to:
1. Make a determination that proposed changes to procedures, equipment, or systems do not involve a change to the procedures or facility as described in the Final Safety Analysis Report.
2. Make a preliminary determination that propased changes to procedures, equipment or systems are descr.5cd in the Final Safety Analysis Report, or that proposed tests or experiments not described in the Final Safety Analysis Report do not constitute an unreviewed safety question; however, such a determination must be based upon a formal written evaluation.
3. Direct the Plant Operations Review Committee to review:
a. Evaluations of proposed changes to procedures, equi'pment or systems; 6-M92 13I
b. Proposed tests and experiments, and to make an initial determination that "a" and "b" above do not constitute an unreviewed safety question.

NOTE: The Unit Superintendent shall report directly to the Manager-Generation Operations-Nuclear and is responsible to him for the administration, operation and maintainance of Three Mile Island Nuclear Station Unit 1.

6.2 ORGANIZATION OFFSITE 6.2.1 The organization of the Met-Ed Corporate Technical Support staff for Station management and technical support shall be as shown in Figure 6-1.

FACILITY STAFF 6.2.2 The organization within the station for operations, technical support, and maintenance shall be functionally as shown in Figure 12-1 of the Final Safety Analysis Report.

a. Each on-duty shif t shall as a minimum be composed of the following shift crew:

Shift Supervisor or Shift Foreman (See Notes 1 & 3) 1 Control Room Operator (See Notes 2 & 3) 2 Auxiliary Operator (See Note 3) 2 Men / Shift 5

b. At least two licensed Reactor Operators shall be at the station, one of whom shall be in the Control Room at all times when there is fuel in the reactor vessel. One of these operators shall hold a Seniar Reactor Operator's License,
c. At least two licensed Reactor Operators shall be present in the Control Room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.
d. At least one member of each operating shif t shall be qualified to implement necessary radiation protection procedures.
e. A licensed Senior Reactor Operator with no other concurrent operational duties shall directly supervise: (a) irradiated fuel handling and transfer activities onsite, and (b) all unirradiated fuel handling and transfer activities to and from the Reactor Vessel.

NOTES:

1. The Shif t Supervisor, or the Shif t Foreman if a Shift Supervisor is not assigned, shall have an NRC Senior Reactor Operator's License.

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2. Only ene licensed Control Room Operator shall be required per shif t during cold shutdown or refueling operations.
3. Shift Supervisor, Control Room Operator and Auxiliary Operator refer to functions that are to be performed and do not refer to the title of the individual. These functions may be performed by any individual possessing the necessary licenses and qualifications.

6.3 STATION STAFF QUALIFICATIONS 6.3.1 Comprising the station staff shall be supervisory and professional personnel encompassing the qualifications described in Section 4 of ANSI-N18.1 (1971), " Selection and Training of Nuclear Power Plant Personnel."

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Supervisor of Training and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.

6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)

FUNCTION 6.5.1.1 The Plant Operations Review Committee shall function to advise the Unit Superintendent on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The Plant Operations Review Connittee shall be corposed of:

a) Unit Superintendent b) Supervisor of Operations c) Supervisor of Maintenance d) Unit Electrical Engineer e) Unit Mechanical Engineer f) Unit Nuclear Engineer g) Unit Instrument and Control Engineer h) Supervisor of Radiation Protection and Chemistry i) PORC Chairman j) Other plant engineers assigned by the Unit Superintendent The Unit Superintendent shall designate the members, the Chairman, and the Vice Chairman of the Plant Operations Review Commit tee .

ALTERNATES 6.5.1.3 Alternate members shall be appointed in writing by the Unit Superintendent to serve on a temporary basis. For purposes of this specification, a designated alternate shall be considered to have the 6

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same responsibility and authority as a member when attending a committee meeting in place of a member.

MEETING FREQUENCY 6.5.1.4 The Plant Operations Review Committee shall meet as required on call by the Unit Superintendent, the Chairman of the Committee or the General Office Review Board, but not less frequently than once per month.

QUORUM 6.5.1.5 A quorum shall consist of four members, at least one of whom shall be either the Chairman or Vice Chairman of the Committee. A quorum shall not take credit for more than one alternate member.

RESPONSIBILITIES 6.5.1.6 The Plant Operations Review Committee shall be responsible for:

a. 1) Review of procedures and changes thereto in accordance with the requirements of Section 6.8, and
2) review of evaluations of proposed changes to procedures to make an initial determination as to whether or not such proposed changes involve an unreviewed safety question when so directed by the Unit Superintendent.

NOTE:

Initial determinations that proposed changes to procedures, equipment or systems, and tests and experiments did not involve an unreviewed safety question shall be subsequently reviewed by the Manager-Generation Engineering to verify that the initial determination was. correct. This review by the Maaager-Generation Engineering shall be documented, b.

1) Review of p roposed tests and experiments, when directed by the Unit Superintendent, to make an initial determination as to whether or not such tests or experiments may involve an unreviewed safety question as defined in 50.59 (c), Parc 50 Title 10, Code of Federal Regulations, and
2) review of the results of all tests and experiments conducted pursuant to paragraph 50.59 (a), Part 50, Title 10, Code of Federal Regulations.

c.

Review of proposed changes to these Technical Specifications or licenses.

d.

Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety as determined by the Unit Superintendent.

e.

1) Review of reportable occurrences under 6.6.2c and any violations of these Technical Specifications or Operating License DPR-50, ir.cluding a report to the >kt-Ed Manager-Generation Operations-Nuclear to the Chairman of General Office Review Board, and to the Unit Superintendent covering evaluation and recommendations to prevent recurrence , a d 6L

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24 review of viola 61ons of applicable federal statutes, codes, regulations and internal station procedures and instructions having nuclear safety significance,

f. Evaluating plant operations for and providing assistance in planning future activities to the Unit Superintendent.
g. Perform special reviews and investigations and submit reports theceon as directed by the Manager-Generation Division, the Manager-Generation Operations-Nuclear or Unit Superintendent.
h. Review of the Plant Security Plan and implementing procedures as they relate to nuclear safety and shall submit recommended changes to the Unit Superintendent.
1. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Unit Superintendent.

AUTHORITY 6.5.1.7 The Plant Operations Review Committee shall:

a. Recommend to the Unit Superintendent written approval or disapproval of items considered under 6.5.1.6(a) through (d) above.
b. If requested by the Unit Superintendent for 6.5.1.6(a) through (d) and at all times for 6.5.1.6(e), render determinations with regard to whether or not each item considerei constitutes an unreviewed safety question.
c. Provide immediate written notification to the Manager-Generation Operations Nuclear of any unresolvable disagreements between PORC and the Unit Superintendent as they may relate to nuclear safety; however, the Unit Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.

Note: The Plant Operations Review Committee shall be advisory to the Unit Superintendent. Nothing herein shall relieve the Unit Superintendent of his responsibility for overall safety of plant operations including taking immediate emergency actions .

RECORDS 6.5.1.8 The Plant Operations Review Committee shall maintain at the station written minutes of each meeting and copies shall be provided to the Unit Superintendent, Managas-Ccreration Operations-Nuclear, Manager-Generation Engineering, and the General Office Review Board Secretary.

6.5.2.A MET-ED CORPORATE TECHNICAL SUPPORT STAFF ORGANIZATION 6.5.2.A.1 The organization of the Met-Ed Corporate Technical Support Staff is as shown on Figure 6-1 and consists of the Manager-Generation Operations Nuclear, Manager-Generation 6-5 M92 05

Engineering, h oger-Generation Maintenance, Manager-Operational Quality Assurance and their staff. The Corporate Technical Support Staff shall collectively have the competence required by ANSI-N18.7-1972, Standard for Administrative Controls for Nuclear Power Plants, Section 4.2.2 or the Manager-Generation Division shall insure that deficiencies can be readily compensated for through the use of outside groups such as GPU Service Corporation staff, consultants, or vendors.

RES PONSIBILITY 6.5.2.A.2 In its concern with the more detailed issues (rather than the broad issues) of nuclear safety, it shall be the responsibility of the Met-Ed Corporate Technical Support Staff to:

a. Review evaluations of proposed changes to procedures, equipment or systems and tests and experiments (including their results) which were accomplished pursuant to 6.1.1.a.2 and 6.1.1.a.3 to verify that an unreviewed safety question was not involved,
b. Control of design changes to equipment or systems having nuclear safety significance as defined in Section 2.2.19 of ANSI-N18.7-1972, including verifying that such proposed changes do not constitute unreviewed safety questions or Technical Specification changes.
c. Specifying tests that must be performed following a design change to demonstrate that safety related structures, components, and systems meet Technical Specification requirements,
d. Review of proposed changes to these Technical Specifications and Operating License DPR-50.
e. Review of violations of applicable federal statutes, codes, regulations, orders, and internal station procedures and instructions having nuclear safety significance.
f. Review of reportable occurrences, and violations of these Technical Specifications and Operating License DPR-50.

g.

Review of station performance records of significant operating abnormalities or deviations from normal and expected performance of plant equipment.

h. Review of indicatic.s of an unanticipated deficiency in some aspect of design or operation of nuclear safety related structures, components or systems, including confirmation of determinations regarding whether they involve unreviewed safety questions, or reportable occurrences.
1. Review of events covered under 6.5.2. A.2.d, e, f, and g shall include reporting to the Manager-Generation Division, Unit Superintendent, and other appropriate members of management on the results of investigations and recommendations to prevent or reduce the probability of recurrence.

J. Development, diraction and overall coordination of Operational Quality Assursace activities.

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k. Periodically audit the areas listed below to verify compliance with the Three Mile Island Cperating Quality Assurance Plan, internal rules and procedures, federal regulations, and operating license provis1ons:
1) The 18 Critiera of 10CFR50, Appendix B
2) Normal Station Operation
3) Inservice Inspection
4) Refueling
5) Radiological Controls
6) Station Maintenance
7) Technical Specificationa
8) Training and Qualifications of Station Staff
9) Emergency Plan
10) Industrial Security Program In performing these audits, uritten procedures and/or checklists shall b e us ed . As a minimum, each area shall be audited at least once every two years.

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AUDITS 6.5.2.A.3 Audits shall periodically be conducted under the direction of the Manager-0perational Quality Assurance to verify compliance of plant operations with aspects of the Three Mile Island Operating Quality Assurance Plan, including verification of compliance with applicable internal rules and procedures; federal regulations and operating license provisions; training qualifications and performance of operating staff. Audits of the Industrial Security Program and the Emergency Plan shall also be conducted at periodic intervals not to exceed two years. In performing these audits, written procedures and/or check lists shall be used and written reports of such audits shall be issued.

AUTHORITY 6.5.2.A.4 The Met-Ed Corporate Technical Support Staff was approved by the Company President. The Company President has assigned to the Manager-Generation Division responsibility for the overall effectiveness of the corporate technical support and plant organizations and the Three Mile Island Operating Quality Assurance Plan. The Mancger-Generation Division fulfills this responsibility by delegating the appropriate authority to the Met-Ed Corporate Technical Support S taf f . The Manager-Generation Division shall issue instructions and procedures which delineate the responsibilities and ithority of the various managers who report to him.

REPORTS TO MANAGEMENT AND THE GENERAL OFFICE REVIEW BOARD 6.5.2.A.5 Reports shall be made to management and the General Office Review Board as follows:

a. The Manager-Generation Division shall report to the Company President any problems identified by the Generation Division staff which require the President's administrative corrective action, together with appropriate recommendations.
b. Any reportabJe occurrence or item involving an unreviewed nuclear safety question which is identified by the Corporate Technical Support Staff review shall be brought to the attention of the Manager-Generation Division, and the General Office Review Board if it has not been previously reported by the Plant Operations Review Committee or Unit Superintendent.
c. Written reports of audits performed pursuant to 6.5.2.A.3 shall be submitted to the Manager-Generation Division and the Chairman, General Office Review Board.

6.5.2.B GENERAL OFFICE REVIEW BOARD (GORB)

FUNCTION 6.5.2.B.1 In its concern with the broader issues (rather than the detailed issues) of nuclear safety, it shall be the primary responsibility of the General Office Review Board to:

a. Foresee potentially significant nuclear and radiation safety problems and to recommend to the President of Met-Ed how they may be avoided.

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b. Periodicall, review the Generation Division .dit program to assure that audits are being accomplished in acco'rdance with requirements of Technical Specifications and ANSI-18.7-1972 Standard for Administrative Controls for Nuclear Power Plants.

COMPOSITION 6.5.2.B.2 a. The Chairman and Vice Chairman shall be appointed by the Company President.

b. The Chairman shall designate a minimum of is ar additional members. No more than a minority of the committee shall have line responsibility for day-to-day operation of Three

>dle Island Nuclear Station.

c. Members of the General Office Review Board shall possess extensive experience in their individual specialties and collectively have the competence set forth in ANSI-N18.7-1972, Standard for Acministrative Controls for Nuclear Power Plants, Section 4.2.2.2.

ALTERNATES 6.5.2.B.3 Alternate members shall be appointed in writing by the Chairman or Vice Chairman of the General Office Review Board to serve on a tetporary basis; however, no more than two alternates shall participate in Review Board activities at any one time.

CONSULTANTS 6.5.2.B.4 Consultants shall be utilized as determined by the Chairman and Vice Chairman of the General Office Review Board to provide expert advice to the Review Board.

MEETING FREQUEFCY 6.5.2.B.5 The General Office Review Board shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.

QUORUM 6.5.2.B.6 A quorum for formal meetings shall have no less than a majority of the principals or duly appointed alternates and shall include the Chairman or Vice Chairman. No more than a minority of the quorum shall hold line responsibility for day-to-day operations of the Three Mile Island Nuclear Station. A quorum shall not take credit for more than two alternate members.

REVIEW 6.5.2.B.7 The General Office Review Board shall review as is consistent with its responsiollities:

a. Proposed changes to procedures, equipment or systems referred to the Committee by the Plant Operations Review Committee, the Unit Superintendent, the Manager-Generation Engineering, or the Manager-Generation Operations-Nuclear.

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b. Proposed tests
  • experiments referred to the c nittee by the Plant Operations Review Committee, the Unit Superintendent, the Manager-Generation Engineering, or the Manager-Generation Operations Nuclear.
c. Proposed changes in and violations of these Technical Specifications or Operating License DPR-50.
d. Operating abnormalities and deficiencies in some aspect of design or operation of nuclear safety related equipment which involves an unreviewed nuclear safety question.
e. 14-Day and 30-Day Reportable Occurrences,
f. Adequacy of the Plant Operations Review Ccmmittee's and the Ebt-Ed Corporate Technical Support Staff's determinations concerning unreviewed safety questions.
g. Audits and audit program of the Generation Division.
h. Adequacy of Plant Operations Review Committee minutes.

AUDITS 6.5.2.B.0 The General Office Review Board shall perform periodic reviews of the Operational Quality Assurance audit program to insure that audits are being accomplished in accordance with the requirements of these Technical Specifications and ANSI-18. 7-1972, " Standard for Administrative Controls for Nuclear Power Plants." Special reviews, audits and investigations shall also be conducted as requested by the Company President er as deemed necescary to confirm the adequate functioning of the station and corporate technical staffs.

AUTHORITY 6.5.2.B.9 The General Office Review Board shall be advisory to the Company President .

Written administrative procedures for committee operation shall be prepared and maintained. These procedures shall describe the requirements for submittal and content of presentations to the committee, provisions for use of subcommittees, review and approval by members of written committee evaluations and recommendations, dissemination and approval of minutes, and other appropriate matters.

RECO RDS 6.5.2.B.10 Records of General Office Review Board activities shall be prepared, approved and distributed as indicated below:

a.

Minutes shall be recorded and approved for all meetings of the General Office Revfew Joard. Copies of the minutes shall be forwarded to the members , Company President , Manager-Generation Division, Unit Superintendent, the Chairman of the Plant Operations Review Committec, the Manager-Generation Operations-Nuclear, and such others as the Chairman may designate.

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b. As appropriate, Lne Chairman of the General Office Review Board shall by letter to the Company President within 14 days following completion of the review:
1) Recommend actions that should be taken on proposed changes to these Technical Specifications or Operating License DPR-50.
2) Recommend actions that should be taken on proposed tests, facility changes, procedure changes, or operating abnormalities which they have reviewed by referral or upon their own initiative.
3) Recommend to the Company President appropriate action to prevent recurrence of reportable occurrences or to improve the effectiveness of the plant and corporate organization.

6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken in the event of an 14-day reportable occurrence:

a. Each occurrence shall be reported immediately to the Manager-Generation Operations-Nuclear, Unit Superintendent, and the Manager-Generation Division and shall be reviewed promptly by the Plant Operations Review Committee. This committee shall prepare a separate report for each occurrence which shall include an evaluation of the cause of the occurrence and recommendations for appropriate action to prevent or minimize the probability of a repetition of the occurrence. Copies of all such reports shall be submitted to the Unit Superintendent, General Office Review Board, and the Manager-Generation Operations-Nuclear.
b. The Nuclear Regulatory Commission Regional Office shall be notified by telephone within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and confirmed by telegram, mailgram, or facsimile transmission no later than the first working day following the occurrence.
c. A written report shall be submitted to the Director, Division of Reactor Licensing and the Director of the Regional Office, of the Nuclear Regulatory Commission within two kicks following the occurrence.

6.6.2 The fcilowing actions shall be taken in the event of a 30-day reportable occurrence:

a. Any 30-day reportable occurrence shall be reported promptly to the Unit Superintendent, Manager-Generation Operations-Nuclear, and the Manager Generation Division. A seperate written report for each occurrence shall be prepared and shall include a description of the occurrence, the cause of the occurrence, (and appropriate corrective action to) prevent or minimize the probability of repetition of the occurrence.

Copies of all such reports shall be submitted to the Unit Superintendent, Manager-Generation Operations-Nuclear, and the Manager Generation Division.

b. Written reports required under' 6.6.2a shall be prepared by the Plant Operations Review Committee for occurrences categorized under Technical Specification 1.9c through 1.9e, and prepared by the Met-Ed Corporate Technical Support Staff for occurrences categorized under Technical Specification 1.9a and 1.9b.

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c. A written report shall be submitted to the Director, Division of Reactor Licensing and the Director of the Regional Office, of the Nuclear Regulatory Commission with 30 days following the occurrence.

6.7 OCCURRENCES INVOLVING A SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a safety limit is violated:

a. The reactor shall be shut down and operation shall not be resumed until authorized by the Nucleae Regulatory Commission.
b. An immediate report shall be maai to the Unit Superintendent, Manager-Generation Operations-Nuclear to the Manager-Generation Division, and to the General Office Review Board, and the occurrence shall be promptly reported to the Nuclear Regulatory Commission in accordance with Technical Specification, 6.6.lb.
c. A complete analysis of the circumstances leading up to and resulting from the occurrence shall be performed by the Plant Operations Review Committee and a report prepared. This report shall include analysis of the effects of the occurrence and recommendations concerning operation of the unit and prevention of a recurrence. This report shall be submitted to the Unit Superintendent, Manager-Generation Operations-Nuclear, the General Office Review Board, and the Manager-Generation Division. Appropriate analysis of reports will be submitted to the Nuclear Regulatory Commission in accordance with Technical Specification, 6.6.lc.

}kh2 6-10a

6.8 PROCEDURES 6.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the requirements and recocmendations of Sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix "A" of USNRC Regulatory Guide 1.33 November 1972 except as provided in 6.8.2 and 6.8.3 below.

6.8.2 Each nuclear safety related procedure had administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the Plant Operations Review Committee and approved by the Unit Superintendent prior to implementation and periodically as may be set forth in each document.

6.8.3 Temporary changes to procedures of 6.8.1 above may be madt provided:

a. The intent of the original procedure is not altered.
b. 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.
c. The change is documented, reviewed by the Plant Operations Review Committee and approved by the Unit Superintendent within 7 days of implementation.

6.9 REPORTING REQUIREMENTS ROUTINE, 14-DAY REPORTABLE OCCURRENCE, AND 30-DAY REPORTABLE OCCURRENCE REPORTS 6.9.1 Information to be reported to the Commission, in addition to the reports required by Title 10, Code of Federal Regulations, shall be the occurrences defined by Technical Specifications 1.8 and 1.9.

In addition, the Annual Operating Report shall include information on aircraft movements at the Harrisburg International Airport. This additional information shall include the total number of aircraft movements. (takeoffs and landings) at the Harrisburg International Airport for the previous twelve-month period. Also included shall be the :otal number of movements of aircraft larger than 200,000 pounds, based on a current percentage estimate provided by the airport manager.

6.9.2 Special reports shall be submitted to the Director of the Regulatory Operations Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below:

Tests Submittal Dates

a. Containment Structural Integrity Test
1. Tendon Surveillance Program Within 3 months after performance of surveillance program.
2. Ring Girder Inspection Within 3 months after Program performance of each inspection.

6-1492 14T

b. Containment In, grated Leak Within 6 wonths after Rate Test completion of test.
c. Inservice Inspection Program Within 6 months after five years of operation.

6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

a. Records of normal station operation including power levels and periods of operation at each power level.
b. Records of principal maintanance activities, including inspection, repairs, substitution, or replacement of principle items of equipment pertaining to nuclear safety.
c. Reports of abnormal occurrences and safety limits exceeded.
d. Records of periodic checks, tests , and calibration.
e. Records of reactor physics tests and other special tests pertaining to nuclear safety.
f. Changes to nuclear safety related operating procedurc6.
g. Records of solid radioactive shipments.
h. By-probact material inventory records and source leak test results.
1. S?ecial nuclear material inventory records.
j. Control Room Log Book.
k. Shif t Foreman's Log.

6.10.2 The following records shall be retained for the duration of Operating License DPR-50:

a. Record and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.
b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
c. Routine station radiation surveys and monitoring records.
d. Records of radiation exposure history and radiation exposure status of personnel, including all contractors and station visitors who enter radioactive material area.
e. Records of radioactive liquid and gaseous wastes released to the environment, and records of environmental monitoring surveys,
f. Records of transient or op.erational cycles for those nuclear safety relate.d f acility components designed for a limited number of transients or cycles as defined in the Final Safety Analysis Report.

6-12

g. Records of training and qualification for current members of the plant staff,
h. Records of in-service inspections performed pursuant to these Technical Specifications.
1. Recoriis of quality assurance activities required by the OQA Plan.
j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
k. Plant Operations Review Committee and General Office Review Board Minutes.

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 _

ALLOWANCE 6.12.1 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, Colu=n 1, of 10 CFR 20, subject to the following conditions and limitations:

a. The limits provided in Section 20.103(a) and (b) shall not be exceeded,
b. 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, of 10 CFR 20.
c. 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.

PROTECTION PROGRAM 6.12.2 In all operations in which adequate limitation of the inhalation of radioactive material by the use of process or other engineering controls 6-13 g492 145

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:

a. The limits specified in 6.12.1 above, are not exceeded.
b. Respiratory protective equipment is selected and used so that the peak concentrations of airborne radioactive material inhaled by an individual wearing 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 concentration 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 later quantity shall be used in evaluating the exposures.
c. The licensee advises each respirator user that he may leave the area at any time for relief from respirator use in case of equipment malfunction, physical or psychological discomfort, or any other condition that might cause reduction in the protection afforded the wearer.
d. 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-Z88.2-1969) .

Such a program shall include:

1. Air sampling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper selection of respiratory protective equipment.
2. Written procedures to assure proper selection, supervision, and training of personnel using such protective equipment.
3. Written procedures to assure the adequate fitting of respirators; and the testing of respiratory protective equipment for operability immediately prior to use.
4. Written procedures for maintenance to assure full effectiveness of respiratory protective equipment, including issurance, cleaning and decontamination, inspection, repair, and storage.
5. Written operational and administrative procedures for proper use of respiratory protective equipment including prceisions for planned limitations on working times as necessitated by operational cond1: ions .
6. Bioassays and/or whole body counts of individuals (and other surveys, as appropriate) to evaluate individual exposures and to assess protection actually provided.

6-14 1492 146

e. The licensee shall use equipment approved by the U. S. Bureau of Mines (or the National Institute of Occupational Safety and Health, as applicable) under its appropriate Approval Schedules as set forth in Table 6.12-1. Equipment not approved under U. S. Bureau of Mines (or National Institute of Occupational Safety and Health, as applicable)

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 characteristics of the equipment are at least equal to those afforded by U. S. Bureau of Mines (or National Institute of Occupational Safety and Health, as applicable) approved equipment of the same type, as specified in Table 6.12-1.

f. Unless otherwise authorized by the Commission, the licensee shall not assign protection factors in excess of those specified in Table 6.12-1 in selecting and using respiratory protective equipment.

6.13 HIGH RADIATION ATIA 6.13.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20:

a. Each High Radiation area (100 mrem /h or greater) in which the intensity of radiation is 1000 mrem /h or less shall be barricaded and conspicuously posted as a high radiation area, and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit. 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 rate 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.l(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 Radiation Protection Supervisor / Foreman or the Shift Foreman on duty.

1492 147 6-15

TABLE 6.12-1 .

PROTECTION FACTORS FOR RESPIRATORS PROTECTION FACTORS GUIDES 'IU SELECTION OF EQUIPMENT PARTICULATES AND BUREAU OF MINES (OR NATIONAL INSTITQTE VAPORS AND GASES OF OCCUPATIONAL SAFETY AND llEALTil, AS APPLICABLE) APPROVAL SCliEDULES* FOR DESCRIPTION MODES 1 EXCEPTTRI}IUM OXIDE EQUIPMENT CAPABLE OF PROVIDING AT LEAST EQUIVALENT PROTECTION FACTORS

  • or schedule superseding for equipment of type listed I. AIR-PURIFYING RESPIRATORS Facepiece, half-mask4 ,7 NP 5 21B 30 CFR 5 14.4(b)(4)

Facepiece , full 7 NP 100 21B 30 CFR 5 14.4(b)(5); 14F 30 CFR 1 II. ATMOSPHERE-SUPPLYING RESPIRATOR f 1. Airline respirator g Facepiece , half-mask CF 100 19B 30 CFR 5 12.2(c)(2) Type C(i)

Facepiece, full CF 1,000 19B 30 CFR 5 12.2(c)(2) Type C'1)

Facepiece, full D 100 19B 30 CFR 5 12.2(c)(2) Type C(ii)

Facepiece, full PD 1,000 19B 30 CFR S 12.2(c)(2) Type C(iii) llood CF 5 6 Suit CF 5 6

2. Self-contained breathing apparatus (SCBAl Facepiece, full' D 100 13E 30 CFR ? 11.4(b)(2)(1)

Facepiece, fuli PD 1,000 13E 30 CFR 5 11.4(b)(2)(ii)

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

III. COMBINATION RESPIRATOR Any combination of cir- Protection factor 19B CFR S 12.2(e) or applicable purifying and atmosphere- for type and mode schedules as listed above g supplying respiratoc of operation as lis ted above 4

CD 1,2,3,4,5,6,7 (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 dhe 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 concen-tration 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 (or National Institute of Occupational Safety and Health, as applicable) 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 submersion 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 I, 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 Nuclear Regulatory Commission.

6-17 1492 149

TABLE 6.12-1 (Continued) 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 (or the National Institute of Occupational Safety and Health, as applicable) 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 te 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 (or the Institute of Occupational Safety and Health, as applicable) 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.

14'92 150 6-18

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UNITED STATES OF AMERICA ,

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NUCLEAR REGULATORY C01DiISSION IN THE MATTER OF DOCKET "'). 50-289 -

LICE 7 E NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 5 Amendment No. 3 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, dated October 3,1975, and filed with the U. S. Nuclear Regulatory Commission on Ocecher 3,1975, has this 3rd day of October been served on the chief execucives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennt,ylvanis by deposit in the United States mail, addressed as follows:

Mr. Weldon B. Arehart, Chairman Mr. Charles P. Hoy, Chairman Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 P.O. Box 1295 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY By  !/

Vice Pres'ident-Generation 1492 152

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