ML19290A354
| ML19290A354 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/27/1976 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19290A352 | List: |
| References | |
| NUDOCS 7911060590 | |
| Download: ML19290A354 (38) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION W ASHING TON. D.
C.
20355 METROPOLITAN EDISCN CO::PANY JERSEY CENTRAL PCWER AND LIGHT COMPANY PENNSYL'.'A' IA ELECTRIC COMPANY DCCKET NO. 50-289 THREE '4ILE ISLAND NUCLEAR STATION, UNIT 1 ANENDMENT TO FACILITY OPERATING LICENSE Amendment No. 11 License No. DPR-50 1.
The Nuclear Rcgulatory Commission (the Commission) has found f. hat:
A.
The application for amendment by Metropolitan Edison Company, Jersey Central Power and Light Company, and Pennsylvania Electric Company (the licensees) dated December 23, 1975, complies 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.
'Ihe 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; 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; and E.
An environmental statement or negative declaration need not be prepared in connection with the issuance of this amendment.
2.
Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment.
1556 091 7911060 N O
r 4 3.
This license atendment becomes effective 30 days after the date of its issuance.
FOR THE NUCLEAR REGULATORY CC:G:ISSION D
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Robert W. Reid, Chief Operating Reactors Branch #4 Division of Reactor Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: M2 7 ';70 1556 092
r ATTACHMENT TO LICENSE AMENDMENT NO.11 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Remove Pages Insert Pages i-Y i - vi 1-5 and 1-6 1-5 6 6-13 6 6-25 Figure 6.1-1 Figure 6-1 Figure 15.5-1 1556 093 e
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Amendment No. 11 tit
I TABLE OF CONTE!.TS Section Page 6.9 REPORTING REQUIRD:E.',TS 6-12 6.9.1 ROUTI.NE REPCRIS 6-12 6.9.2 REPORTABLE OCCURRENCES 6-14 6.9.3 UNIQUE REPCRTING PIQUIRDiENTS 6-18 6.10 RECORD PITENTICN 6-19 6.11 RADIATION PROTECTION PROGRA.1 6-20
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?00ROR3NAL 1556 097 Amendment No. 11 iv
LIST OF T.GLES Pace Table Title 2.3-1 Reactor Protection Systc= Trip Setting Limits 2-9 3-29 3.5-1 Instruments operating Conditions 4.1-1 Instrument Surveillance Requirenents 4-3 4-8 4.1-2 Mininum Equipment Test Frequency 4-9 4.1-3 Mini.um Sampling Frequency 4.2-1 Instrunent Surveillance Progran 4-14 4.4-1 Selected Tendons and Corresponding Inspection 4-35a Periods 4.4-2 Tendons selected for Tendon Physical Condition 4-36 Test 4.4-3 Ring Girder Surveillance 4-36g 4.15-1 Radioactive Liquid Waste Sampling and Analysis 4-59 4.15-2 Radioactive Gaseous Waste Samaling and Analysis 4-63 6.12-1 Protection Factors for Respirators 6-23 1556 098 Amendment No. 11 y
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?00R BR%g 1-5
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, equip =ent, or syste=s are evaluated to deter =ine if they constitute a change to the facility or procedures as described in the Final Safety Analysis Report.
2.
All proposed changes to procedures, equipment, or syste=s which constitute a change of the f acility or procedur2s as described in the Final Safety Analysis Report are evaluated to deter =ine that they do not involve an unreviewed safe ty question as defined in paragraph 50.59 (c), Part 50, Title 10, Code of Federal Regulations.
3.
All proposed tests and experi=ents, not described in the Final Safety Analysis Report, are evaluated to deter =ine that they do not involve an unreviewed safety question as defined in paragraph 30.59 (c), Part 50, Title 10, Code of Faderal Regulations.
4.
Records are kept: a) of changes to procedures, equip =ent or systems ce=pleted under the provisions of paragraph 30.59 (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 deter =ining that such changes, tests and experiments do not involve an unreviewed safety question.
5.
Copies of evaluattens c:nducted pursuant to -6.1.1.a. 2 and 6.1.1.a.1 above are foruarded to the Plant Operations Review l
Co==ittee, the Manager-Generation Engineering, and the General I
Office Review Board Secretary.
b.
The Unit Superintendent shall have the authority to:
1.
Make a deter =ination that proposed changes to procedures, equipment, or syste=s do not involve a change to the procedures or f acility as described in the Final Safety Analysis Report.
2.
Make a preli=inary deter =ination that proposed changes to procedures, equipment or syste=s as described 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 detar=ination must be based upon a for=al written evaluation.'
3.
Direct the Plant Operations Review Co==ittee to review:
a.
Evaluations of proposed changes to procedures, equip =ent or syste=s ;
Amend =ent No. 11 1556 101
-c-;
f i
b.
Proposed tests and experiments, and to make.an initial deter =ination 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 functionally 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 showr 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 shif t crew:
Shif t Supervisor or Shif t 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 Senior Reactor Operator's License, c.
At leas t two, licensed Reactor Operators shall be present in the Control Room during reactor start-up, scheduled reactor shutdown and during recovery f rom 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 bjb duties shall directly supervise:
(a) irradiated fuel handling and transfer activities onsite, and (b) all unirradiated fuel q'.
handling and transfer activities to and from the Reactor Vessel.
NOTES:
1.
The Shift Supervisor, or the Shift Foreman if a Shift Supervisor is not assigned, shall have an NRC Senior Reactor Opers, tor's License.
6-2 102 Amendment No. 11
a 2.
Only one licensed Gontrol Roco Operator shall be required per shif t during cold shutdown or refueling operations.
3.
Shif t 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 f unctions cay be perfor=ed by any individual possessing the necessary licenses and qualifications.
6.3 STATION STAFF OUALIFICATIONS 6.3.1 Co=prising the station staff shall be supervisory and professional personnel enco: passing the qualifications described in Section 4 of ANSI U18.1-1971, " 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 ceet or exceed the requirements and recommendations I
of Section 5.5 of ANSI N18.1-19 71 and Appendix "A" of 10 CFR Part 55.
f
_6. 5 REVIEW AND AUDIT l
6.5.1 PLANT OPERATIONS PIVIEW C0:01ITTEE (PORC)
FUNCTION 6.5.1.1 The Plant Operations Revieu Committee shall function to advise the Unit Superintendent on all matters related to nuclear safety.
{
I COMPOS ITION 6.5.1.2 The Plant Oper,ations Review Coenittee shall be cc: posed of:
a)
Unit Superintendent l
b)
Supervisor of Operations c)
Supervisor of Maint3 nance d)
Unit Electrical Engineer e)
Unit Methanical Engineer f)
Unit Nuclear Engineer g)
Unit Instrument and Control Engincer h)
Supervisor of Radiation Protection and Chemis try 1)
PORC Chair =an j)
Other plant engineers assigned by the Unit Superintendent The Unit Superintendent shall designate the me=bers, the Chairman, and the Vice Chairman of the Plant Operations Review Commit tee.
ALTERNATES 6.5.1.3 Alternate ec=bers 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 Amendment No. 11 6-3 1556 103
sa=e responsibility and authority as a =e=ber when attending a co==ittee =eeting in place of a =e=ber.
MEETING FREOUENCY 6.5.1.4 The Plant Operations Review Cc==ittee shall =eet as required on call by the Unit Superintendent, the Chair =an of the Ccenittee or the General Of fice Review Board, but not less frequently than once per =onth.
l I
QUORCM
)
6.5.1.5 A quoru= shall consist of four =e=bers, at least one of who= shall be either the Chair =an or Vice Chair =an of the Co==ittee.
A quoru= shall not take credit for more than one alternate =e=b2r.
RESPONSI3ILITIES 1
6.5.1.6 The Plant Operations Review Co==ittee shall be responsible for:
a.
- 1) Review of procedures and changes thereto in accordance with the i
require =ents of Section 6.8, and l
l
- 2) review of evaluations of proposed changes to procedures to =ake l
an initial deter =ination as to whether or not such proposed changes involve an unreviewed safety question when so directed by the Unit Superintendent.
NOTE:
Initial deter =inations that proposed changes to procedures, equip =ent or systems, add tests and experi=ents did not involve an unreviewad safety question shall be subsequently l
reviewed by the Manager-Generation Engineering to verify that the initial deter =ination was correct. This review by the Manager-Generati:n Engineering shall be docu=ented.
b.
- 1) Review of proposed tests and experi=ents, when directed by the l
Unit Superintendent, to =ake an initial deter =ination as to whether or not such tests or experi=ents =ay involve an unreviewed safety question as defined in 50.59 (c), Part 50, Title 10, Code of Federal Regulations, and
- 2) review of the results of all tests and experi=ents conducted pursuant to paragraph 30.59 (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 =edifications to plant syste=s or equip =ent that affect nuclear safety if directed by the Unit Superintendent.
e.
- 1) Review of reportable occurrences under Section 6.6 and any violations of these Technical Specifications or Operating License DPR-50, including a report to the Met-Ed Manager-Generation Operations-Nuclear to the Chair =an of General Of fice Review Board, and to the Unit Superintendent covering evaluation and reco==endations to prevent recurrence, and A=end=ent No. 11 6-4 1556 104
~
- 2) review of violations of applicabic fec_ al stat.es, codes, i
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.
Perform special reviews and investigations and submit reports g.
thereon as directed by the Manager-Generation Division, the Panager-Generation Operations-Nuclear or Unit Superintendent.
h.
Review of the Plant Security Plan and i=plementing procedures as they relate to nuclear safety and shall submit recce= ended changes to the Unit Superintendent.
1.
Review of the E=ergency Plan and implementing procedures and shall submit recoc= ended changes to the Unit Superintendent.
AUTHORITY 6.5.1.7 The Plant Operations Review Cocsittee shall:
Reco==end to the Unit Superintendent written approval or a.
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 ti=es for 6.5.1.6(e), render determinations with regard to whether or not each item considered constitutes an unreviewed safety question.
Provide ic=ediate written notification to the Manager-Generation Operatiens c.
Nuclear of any unresolvable disagreements between FORC 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.
The Plant Operaticas Review Co==ittee shall be advisory to the a
Note:
y Unit Superintendent. Nothing herein shall relieve the
(,
4 Unit Superintendent of his responsibility for overall gj safety of plant operations including taking in:ediate energency actions.
I RECORDS the station 6.5.1.8 The Plant Operations Review Co=mittee shall maintain at written minutes of each meeting and copies shall be provided to the Unit Superintendent, Manager-Generation Operations-Nuclear, Manager-Generation Engineering, and the General Office Review Board Secretary.
6.5.2.A MET-ED COR"0 RATE TECHNICAL SUPPORT STAFF ORGANIZATION The organization of the Met-Ed Corporate Technical Support Staf f 6.5.2.A.1 is as shown on Figure 6-1 and consists of the Manager-Generation Operations Nuclear, Manager-Generation Amendment No. 11 6-5 1556 105
Engineering, Manager-Generation Maintenance, Manager-0perational Quality Assurance and their staff. The Corporate Technical Support Staff shall collectively have the cc=petence required by ANSI NlS.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 CPU Service Corporation staff, consultants, or vendors.
RESPONSIBILITY 6.5.2.A.2 In its concern with the = ore detailed issues (rather than the broad issues) of nuclear safety, it shall be the responsibility of the Met-Ed Corporate l' Technical Support S taff to:
Review evaluations of proposed changes to procedures, equipment a.
or syste=s and tests and experiments (including their results) l which were acco plished 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 equip =ent or syste=s having nuclear I
safety significance as defined in Section 2.2.19 of ANSI N18.7-1972, including verifying that such proposed changes do not f'
constitute unreviewed safety questions or Technical Specification j
changes.
c.
Specifying tests that cust be performed following a design change to de=ons trate that safety related structures, cc=ponents,
and systens =eet Technical Specification require =ents, d.
Review of proposed changes to these Technical Specifications and Operating License DPR-50.
Review of violations of applicable federal statutes, codes, e.
regulations, orders, and internal station procedures and instructions having nuclear safety significance.
f.
Review of reportable occurrences, and violations of these i
Technical Specifications and Operating License DPR-50.
I g.
Review of station performance records of significant operating abnor=alities or deviations from normal and expected performance of plant equipment.
h.
Review of indications of an unanticipated deficiency in sc=e aspect of design or operation of nuclear safety related structures, co=ponents or syste=s, including confir=ation 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, L' nit Superintendent, and other appropriate = embers of =anagement on the results of investigations and reco==endations to prevent or reduce the probability of recurrence, j.
Development, direction and overall coordination of Operational Quality Assurance activities.
A=ener nt :.c. ::
e-6 1556 106
f k.
Periodically audit the areas listed below to verify compliance with the Three Mile Island Operating Quality Assurance Plan, internal rulds and procedures, federal regulations, and operating license provisions:
1)
The 18 Critiera of 10CFR50, Appendix 3 2)
Normal Station Operation 3)
Inservice Inspection 4)
Refueling 5)
Radiological Controls 6)
Station Maintenance 7)
Technical Specifications 8)
Training and Qualifications of Station Staff 9)
Emergency Plan 10)
Industrial Security Program In perfor=ing these audits, written procedures and/or checklists shall be used. As a minimum, each area shall be audited at least once every two years.
Eb h0f Amendment No. 11 6-6a
AUDITS 6.5.2.A.3 Audits shall periodically be conducted under the direction of the Manager-Operational Quality Assurance to verify cocpliance 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 qualificatiens 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 perfor=ing 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 i
Company President. The Co=pany 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 Manager-Generation Division fulfills this responsibility by delegating the l
appropriate authority to the Met-Ed Corporate Technical Support
}
Staff. The Manager-Generation Division shall issue instructions g
and procedures which delineate the responsibilities and authority i
of the various managers who report to him.
l REPORTS TO MANAGEMENT AND THE CENERAL OFFICE REVIE;I BOARD 6.5.2.A.5 Reports shall be =ade to canage:ent and the General Of fice Review Board as follows:
l The Manager-Generation Division shall report to the Cocpany a.
President.any prcblems identified by the Generation Division staf f which require the President's alministrative corrective action,.together with appropriate reconmendations.
i i
b.
Any reportable occurrence or item involving an unreviewed nuclear safety question which is identified by the Corporate Technica' Support Staf f 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.
Written reports of audits perfor:ed pursuant to 6.5.2.A.3 c.
shall be subsitted to the Manager-Generation Division and the Chairman, General Office Review Board.
6.5.2.3 CENERAL OFFICE REVIE!? 30ARD (COP 3)
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:
Foresee potentially significant nuclear and radiation safety problets a.
and to reco==end to the President of Met-Ed how they cay be avoided.
Amendment No. 11 1556 108
b.
Per icall; aview the Generation D
!sion lit program to assure that audits are being acco=plished in accordance with requirements of Technical Specifications and ANSI 13.7-1972 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 Cc=pany President.
b.
The Chairman shall designate a =inimum of four additional me=bers. No = ore than a =inority of the co==ittee shall have line responsibility for day-to-day operation of Three idle Island Nuclear Station.
c.
Me=bers of the General Of fice Review Board shall possess extensive experience in their individual specialties and collectively have the co=petence set forth in ANSI NIS. 7-1972, Standard for Administrative Controls for Nuclear Power Plants, Section 4.2.2.2.
ALTE RNATE S f
6.5.2.B.3 Alternate ce=bers shall be appointed in writing by the Chairman or Vice Chair =aa of the General Of fice Review Board to serve on a te=porary basis.
CONSULTANTS 6.5.2.B.4 Consultants shall be utilized as deter =ined by tha Chairman and Vice Chair an of the General Of fice Review Board to provide expert advice to the Review Board.
g i
MEETING FREOCENCY l
6.5.2.B.5 The Cencral Of fice Review 3 card shall =cet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six conths thereaf ter.
l QUORUM l
6.5.2.B.6 A quorum for formal =eetings 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 =inority of. the quorum shall hold line responsibility for day-to-day operations of the Three Mile Island Nuclear S tation. A quorum shall not take credit for more than two alternate =e=bers.
\\
REVICJ 6.5.2.B.7 The General Office Review Board shall review as is consistent with its responsibilities:
a.
Proposed changes te procedures, equipment or syste=s referred to the Co==ittee by the Plant Operations Review Co==ittee, the Unit Superintendent, the Manager-Generation Engineering, or the Manager-Cencration Operations-Nuclear.
Amendment No, 11 6-8 1556 109
b.
Proposed tests and experiments referred to the committee by the l
Plant Ope re-' ons Re riew Co==ittee, the Unit cuperir endent, the Kanage enert
- n Engineering, or the ager
.ne ra tion Operations Nuclear.
Proposed changes in and violations of these Technical Specifications or c.
Operating License DPR-50.
I d.
Operating abnormalities and deficiencies in some aspect of design or operation of nuclear safety related equip ent which involves an unreviewed nuclear safety question.
e.
Reportable Occurrences.
f.
Adequacy cf the Plant Operations Review Cocsittee's and the Met-Ed i
Corporate Technical Support Staff's determinations concerning unreviewed safety questions.
I g.
Audits and audit program of the Generation Division.
h.
Adequacy of Plant Operations Review Co=mittee minutes.
AUDITS 6.5.2.B.8 The General Office Review Board shall perfocm periodic reviews of the Operational Quality Assurance audit progran to insure that audits are being accomplished in accordance ' tith the require ents of these Technical Specifications and ANSI N18.7-l?72, " Standard for Administrative Controls f or Nucicar Power Plants." Special reviews, audits and investigations shall also be conducted as requested by the Company "Tesident or as dee=ed necessary to confirm the adequate functioning of the station and corporate technical staffs.
AUTHORITY I
6.5.2.B.9 The General Office Review Board shall be advisory to the Company President.
Written administrative procedures for catmittee operation shall be prepared and maintained.
These procedures shall describe the l
requirements for sub=ittal and content of presentations to the co=mittee, provisions for use of subcommittees, review and approval by =echers of written co=mittee evaluations and recom=cadations,
j dissemination and approval of minutes, and other appropriate matters.
l 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 Review Board. Copies of the minutes shall be forwarded to the =cebers, Company President, Manager-Generation Division, Unit Superintendent, the Chairman of the Plant Operations Review ~ Committee, the Manager-Generation Operations-Nuclear, and such others as the Chairman may designate.
Amendment No. 11 6-9 1556 110
b.
As appropriate, the Chairman of the General Office Review Board shall by letter to the Company President within 14 days following completion of the review:
- 1) Reco==end actions that should be taken on proposed changes to these Technical Specifications or Operating License DPR-50.
- 2) Reco==end actions that should be taken on proposed tests, facility changes, procedure changes, or operating abnor=alities which they have reviewed by referral or upon their own initiative.
- 3) Recocmend to the Company President appropriate action to pr? vent recurrence of reportable occurrences or to improve the effectiveness I
of the plant and corporate organization.
6.6 REPORTA3tE OCCURRENCE ACTIO
6.6.1 The following actions shall be taken in the event of a reportable occurrence requiring proept notification with written follow-up:
Each occurrence shall be reported i==ediately to the Manager-Generation a.
Operations-Nuclear, Unit Superintendent, and the Manager-Generation Division and shall be reviewed promptly by the Plant Operations Review Cc=nittee. This coesittee shall prepare a separate report for each occurrence which shall include an evaluation of the cause of the occurrence and recom=endations 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 Of fice Review Board, and the Manager-Generation Operations-Nuclear.
1 i
b.
The Nuclear Regulatory Cot =ission shall be notified in accordance with the require =ents of Technical Specification 6.9.2.A.
6.6.2 The fc21owing actions shall be taken in the event of a reportable occurrence requiring a thirty-day written report.
Each such occurrence shall be reported promptly to the Unit a.
Superintendent, Manager-Generation Operations-Nuclear, and the Manager l
Generation Divisien. A separate written report for each occurrence j
shall be prepared and shall include a description of the occurrence, the cause of the occurrence, (and appropriate corrective action to) prevent or inimize the probability of repetition of the occurrence.
Copies of all such reports shall be submitted to the Unit Superintendent, j
Manager-Generation Operations-Nuclear, and the Manager Generation Division.
- b.
Such written reports shall normally be prepared by the Plant Operations Review Coesittee, but mr.y be prepared by the Met-Ed Corporate Technical Support Staf f if appropriate.
1556 111 Amendment No. 11 6-10
I A written report shall be sub=itted to the Nuclear Regulatory c.
Cocsissien in accordance with the require =ents of Technical Specification 6.9.2.B 6.7 OCCURPINCES INVOLVING A SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be takea in the event a safety limit is violated:
The reactor shall be shut down and operation shall not be resumed a.
until authorized by the Nuclear Regulatory Coc=1ssion.
b.
An ic=ediate report shall be made 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 Cc= mission in accordance with Technical Specification, 6.9.2.A A complete analysis of the circunstances leading up to and resulting c.
from the occurrence shall be perfer ed by the Plant Operations Review Co=sittee and a report prepared.
This report shall include analysis of the effects of the occurrence and reco==endations 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, 6.9.2.A.
56 ll2 Amendment No. 11 6-10a
f 6.8 PROCEDURES 6.8.1 Written procedures and administrative policies shall be established, implemented and =aintained that meet or exceed the requirements and reco==endations of Sections 5.1 and 5.3 of ANSI N18.7-1972 and Appendix "A" of USNRC Regulatory Guide 1.33 Noveeber 1972 except as provided in
- 6. 8.2 and 6.8. 3 below, t
6.8.2 Each nuclear safety related procedure and administrative policy of l
6.8.1 above, and changes thereto, shall be reviewed by the Plant Operations Review Co=sittee and approved by the Unit Superintendent i
prior to implementation and periodically as =ay be set forth in each document.
6.8.3 Te=porary changes to procedures of 6.8.1 above may be made provided:
The intent of the original procedure is not altered.
a.
b.
The change is approved by two members of the plant management staff, i
at least one of whcm holds a Senior Reactor Operator's License on the unit af fected.
e i
c.
The change is docu=ented, reviewed by the Plant Operations Review l
Co=mittee and approved by the Unit Superintendent within 7 days of i=plementation.
1556 113 Amendment No. 11 6-11
e s
k 6.9 REPORTING REOUIREMENTS 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.
6.9.1 Routine Reports A.
Startun Reoort. A su= mary 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 l
8 been manuf actured by a different fuel supplier, and (4) modifications that =ay 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.
l Any corrective actions that were required to obtain l
satisfactory operation shall also be described. Any additional specific de. tails required in license conditions based on other commitments shall be included in this report.
Startup reports shall be sub=itted within (1) 90 days following completion of the startup test progras, (2) 90 i
days following resumption or com=encement of co==ercial power operation, or (3) 9 =onths following initial criticality, whichever is earliest.
If the Startup Report does not cover all three events (i.e., initial criticality,
i completion of startup test progras, and resu=ption or co==encement of co==ercial power operation), supple =entary reports shall be sub=itted at least every three =onths until all three events have been completed.
Annual Operating Report.2/ outine operating reports covering R
i B.
the operation of the unit during the previous calendar year j
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 cade by licensees shall i'
provide a comprehensive su= mary of the operating experience gained during the year, even though some repetition of Amendment No. !'
6-12
. c._
6.9 REPORTI!!G REQUIRE:E';TS (cont'd) previously reported information may be involved.
References in the annual operating report to previously submitted reports shall be clear.
l.
I Each annual operating report shall include:
I I
(1) A narrative su==ary of operating experience during the report period relating to safe operation i
of the f acility, including safety-related mainte-nance not covered in item 6.9.1B(2)(e) below.
1 (2) For each outage or forced reduction in power !of over twenty percent of design power level where the reduction extends for greater than four hours :
(a) the proximate cause and the sys tem and maj or component involved (if the outage or forced i
reduction in power involved equipment malfunction);
(b) a brief discussion of (or reference to I
reports of) any reportable occurrences pertain-ing to the outage or power reduction; l
(c) corrective action taken to reduce the proba-bility of recurrence, if appropriate; (d) operating time lost as a result of the outage or power,r; eduction (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),
(e) a description of major safety-related corrective maintenance performed during the outage or power reduction, including the j
system and component involved and identifica-l tion of the critical path activity dictating the length of the outage or power reduction; and (f) a report of any single release of radio-activity or radiation exposure specifically associated with the outage which accounts for = ore than 10% of the allowable annual values.
1556 115 Amend ent No. 11 6-13
6.9 REPORTING REQUIRE:ENTS (cont'd)
(3) A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 = rec /yr and their associated man rem exposure according to work and job f unctions, 5/ e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), was te processing, and refueling. The dose assign =ent to various duty functions =ay be esttnates based on pocket dosimeter, TLD, or fil= badge =easure=ents. 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 =ajor work functions.
(4)
Indications of failed fuel resulting from irradiated fuel examinations, including eddy current tests, ultrasonic tests, or visual examinations completed
'during the report period.
6.9.2 Reportable Occurrences /
1 Reportable occurrences, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of an occurrence.
In case of corrected or supplemental reports, reference shall be made to the original report date.
1556 116 Amendment No. 11 6-14
6.9 REPORTING REOUIRE ENTS (cont'd)
A.
Prompt Notification With Written Follow-Up 1/ 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 confir=ed by telegraph, mailgram, or f acsimile transmission to the Director of the appropriate Regional Office, or his designate no later than the first working day following the event, with a written follow-up report within two weeks.
The written follow-up report shall include material to provide complete explanation, cause of the event, the circumstances surrounding the event, any corrective action, and component failure data.
(1) Failure of the reactor protection system or other systems subject to limiting safety sys tem settings to initiate the required protective function by the time a monitored para =eter reaches the setpoint specified as the limiting safety system setting in the technicel specifications or failure to complete the required protective function.
No te :
Instru=ent drift discovered as a result of testing need not be reported under this item but may be
' reportable under items 6.9.2A(5), 6.9.2A(6), or I
6.9.2.B(1) below.
f (2) Operation of the unit or affected systems when any para =eter 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 leas t 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 6.9.2.3(2) below.
(3) Abnormal degradation discovered in fuel cladding, reactor coolant, pressure boundary, or primary containuent.
Note:
Leakage of valve packing or gaskets within the limits for identified leakage set forth in technical spec-ifications need not be reported under this ites.
l Amendment No. 11 6-15
9 6.9 REPORTING REOUIREMENTS (cont'd) (4) Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to 1% ak/k; a calculated reactivity balance indicating a shutdown =argin 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 = ore than 0.5% dk/k; or occurrence of any unplanned criticality. l 1 (5) Failure or calfunction of one or more components which I prevents or could prevent, by itself, the fulfillment of the functional requirements of systes(s) used to i cope with accidents analyzed in the FSAR. (6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of syste=s required to cope with accidents analyzed in the FSAR. Note: For items 6.9.2A(5) and 6.9.2A(6) reduced redundancy that does not result in a loss of system function need not be reported under this j section but =ay be reportable under ite=s 6.9.2.B(2) and 6.9.2.3 (3). I (7) Conditions arising f rom natural or man-made events that, as a direct result of the event require plant shutdown, operation of safety systems, or other protective =easures required by technical specifications. I (8) Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the FSAR or in the bases for the Technical Specifications that have or could have permitted reactor operation in a i manner less conservative than assumed in the safety analyses. (9) 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 FSAR or Technical Specifications bases ; or discovery during plant life of conditions not specifically considered in the FSAR or Technical Specifications that require remedial action or corrective measures to prevent the exis t'ence or development of an unsafe condition. Amendment No. 11 6-16
6.9 REPORTING REQUIPIMENTS (cont'd) Note: This item is intended to provide for reporting of potentially generic problems B. Thirty Day Written Reports. 1/ 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 narrative =aterial 3 to provide complete explanation of the cause of the event, circu=- { stances surrounding the event, any corrective action, and component I failure data. (1) Reactor protection system or 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 fulfill =ent of the functional requirements of affected syste=s. j i (2) Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a liciting condition for operation. Note: Routine surveillance testing, instrument calibration, or preventative maintenance which require system l configurations as described in items 6.9.2.B(1) and 8, 6.9.2.B(2) need not be reported except where test results themselves reveal a degraded code as described above. (3) 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 syste=s. (4) Abnormal degradatica of syste=s other than those specified in item 6.9.2. A(3) above designed to contain radioactive material resulting from the fission process. Note: 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 3 i this item. 1556 l19 Amendment No. 11 6-17
6.9 REPORTING ?l0UIPI:G::TS (cont'd) 6.9.3 Unis Report.ag Requirecents A. A section in the Annual Operating Report shall include information on aircraf t movements at the Harrisburg International Airport. This additional infor=ation shall include the total nu=ber of aircraft =ovements. (takeoffs and landings) at the Harrisburg International Airport for the previous twelve-month period. Also in-cluded shall be the total number of cove =ents of aircraf t larger than 200,000 pounds, based on a current percentage estimate provided by the airport manager. B. Special reports shall be submitted to the Director of the Regulatory Operations Regional Office within the ti=e period specified for each report. These reports shall be submitted covering the activities identified below: Tests Submittal Dates (1) Contain=ent Structural Integrity Test (a) Tendon Surveillance Program Within 3 months after perfor=ance of surveillance program. (b) Ring Girder Inspection Within 3 months after Program performance of each inspectic: (2) Contain=ent Integrated Leak Within 6 months after Rate Test completion of test. (3) Inservice Inspection Program Within 6 conths after five years of operation. FOOTNOTES 1. These reporting requirements apply only to Appendix A technical specifications. 2. A single submittal may be made for a cultiple wait station. The submittal should combine these sections that are common to all units at the station. 3. The term " forced reduction in power" is normally defined in the electric power industry as dae occurrence of a ccaponent 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. 4. The ter " forced outage" is nor= ally defined in the electric power industry as the occurrence of a component failure or other. condition which requires that the unit be removed from service for corrective action t=nediately or up to and including the very next weekend. 5. This tabulation supple =ents the require =ents of 520.407 of 10 CFR Part 20. Amendment No. 11 6-18
6.10 FICORD RETENTION 6.10.1 The following records shall be retained for at least five years: a. Records of nor=al 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. Records of reportable occurrences and safety limits exceeded. d. Records of periodic checks, tes ts, and calibration. e. Records of reactor physics tests and other special tests pertaining to nuclear safety. f. Changes to nuclear safety related operating procedures. g. Records of solid radioactive shipments. I h. By-product material inventory records and source leak test results. 1. Special nuclear material inventory records. {
- j. Control Room Log Book.
k. Shif t Foreman's Log. j t 6.10.2 The following records shall be retained for the duration of Operating g License DPR-50: a. Record and drawing changes reflecting facility design =odifications made to syste=s and equiptent described in the Final Safety Analysis l Report. i b. Records of new and irradiated fuel inventory, fuel transfers and asse=bly burnup histories. c. Routine station radiation surveys and monitoring records. d. Records of radiation exposure history and radiation exposure status of personnel, including ali contractors and station visitors who ,\\ enter radioactive material area. e. Records of radioactive liquid and gaseous wastes released to the environment, and records of environ = ental conitoring surveys. f. Rec'ords of transient or operational cycles for those nuclear safety related facility components designed for a limited number of transients or cycles as defined in the Final Safety Analysis Report. hh l2l Amendment No. 11
Records of training and qualification for current members of the g. plant staff. h. Records of in-service inspections performed pursuant. to these Technical Specifications. 1. Records 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 cade 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 l linita tions : [ The limits provided in Section 20.103(a) and (b) shall not be exceeded. l a. b. If the radioactive =aterial is of such form that intake through the I skin or other additional route is likely, individual exposures to radioactive caterial shall be controlled so that the radioactive i' content of any critical organ f rom all routes of intake averaged over 7 consecutive days does not exceed that which would result from inhaling such radioactive material for 40 hours at the pertinent concentration values provided in Appendix 3, Table I, Column 1, of 10 CFR 20. c. For radioactive =aterials des ignated "Sub" in the " Isotope" column of Appendix 3, 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 =aterials 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 Amendnent No.11 6-20
is impracticable, the licensee may per=it an individual in a restricted l area to use respiratory protective equip =ent to limit the inhalation l of airborne radioactive =aterial, provided: a. The li=its specified in 6.12.1 above, are not exceeded. l b. Respiratory protective equipment is selected and used so that the peak concentrations of airborne radioactive caterial inhaled by an individual wearing the equipment do not exceed the pertinent concentration values specified in Appendix 3, Table I, Column 1, of 10 CFR 20. For the purposes of this subparagraph, the concentration of radioactive =aterial that is inhaled when respirators are worn =ay be deter =ined by dividing the a=bient airborne concentration by the protection f actor specified in Table 6.12-1 for the respirator protective equipment worn. If the intake of radioactivity is later deter =ined by other ceasurements to have been different than that initially estimated, the later quantity shall be used in evaluating the exposures. The licensee advises each respirator user that he =ay leave the area c. at any ti=e for relief from respirator use in case of equipment =alfunction, physical or psychological disco = fort, or any other condition that =ight cause reduction in the protection afforded the wearer. d. The licensee =aintains a respiratory protective progra= adequate to assure that the require =ents above are cet and incorporates practices for respiratory protection consistent with those reco== ended by the A=erican National Standards Institute (ANSI Z88.2-1969). Such a progra shall include:
- 1. Air sa=pling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper selection t
of respiratory protective equip =ent. I
- 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 equip =ent for operability i==ediately prior to use.
j t
- 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 ad=inistrative procedures for proper use of respiratory protective equipment including provisions for planned limitations on working ti=es as necessitated by operational conditions.
6. Bioassays and/or whole body counts of individuals (and other surveys, as appropriate) to evaluate individual exposures and to assess protection actually provided. Amendment No. 11 6-2 1556 123
e. The licensee shall use equiptent 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, th at the material and performance characteristics of l the equipment are at least equal to those afforded by U. S. Bureau of l Mines (or National Institute of Occupational Safety and Health, as l 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 HICH RADIATION AREA 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 crem/h or greater) in which the intensity of radiation is 1000 cres/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 conitoring 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 =res/hr shall be subject to the provisions of 3 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 con' rol of the Radiation Protection Supervisor / Foreman or the Shif t Foreman on duty. 1556 124 Amendment No. 11 6-22
TAllLE 6.12-1 !i O
Tl:CTION FAC1' ORS FOR RESPIRATORS PRO g
d r, N PROTECTION FACTORS CUIDES TO SELECTION OF EQlllPMENT ~ PARTICULATES AND BUllEAU OF MINES (OR NATIONAL INSTITUTE H VAPORS AND GASES OF OCCUPATIONAL SAFETY AND llEALTil, AS' EXCEPT T!tLTIUll APPLICAllLE) APPROVAL SCllEDUI.ES* FOR DESCRIPTION MODES OX1DF.3 EQUIPMENT CAPAllLE OF PROVIDING AT 1 s LEAST EQUIVALENT PitOTECTION FACTORS g
- or schedule superseding for w
equipment of type listed I. AIR-PURIFYING RESPIRATORS 4 Faceplece, half-mask,7 NP 5 21B 30 CFR 5 14.4(b)(4) 7 Facepiece, fu11 NP 100 2113 30 CFR S 14.4(b)(5); 14F 30 CFR 1 f II. ATMOSPIIERE-SUPPL.YING ICES PIRATOR J S' 1. Airline res pi rator ~ U' Faceplece, half-mask CF 100 191130 CFR S 12.2(c)(2) Type C(i) Facepiece, full CF 1,000 191130 CFR 5 12.2(c)(2) Type C(i) 7 Facepiece, full D 100 19B 30 CFR S 12.2(c)(2) Type C(ii) Faceplece, full PD 1,000 191130 CFR S 12.2(c)(2) Type C(111) 5 6 llood CF 5 6 q g Suit CF 2. Sel f'-contained b reathing appa rat us (Scil A). Facepiece, full' D 100 13E 30 CFR 5 11.4(b)(2)(t) Facepiece, full PD 1,000 13E 30 CFR S 11.4(b)(2)(ii) Faceplece, full It 100 13E 30 CFR S 11.4(b)(1) ~ L.r1 U III. COMBltlATION RESPIRATOR Any combination of air-Protection factor 1911 CFR S 12.2(e) or applicable purifying and atmosphere-for type and mode schedules as listed above N supplying respirator of operation as 01 lis ted above 1, 2, 3,4,5,6,7 (These notes are on the following pages)
a TABLE 6.12-1 (Continued) 1 See the following symbols: CF: continuous flow demand D: negative pressure (i.e., negative phase during inhalation) NP: pressure demand (i.e., always positive pressure) PD: R: recirculating (closed circuit) For purposes of this specification the protection f actor is a measure 2 (a) of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radioactive material outside i to that inside the equipment the respiratory protective equipment It is (usually inside the facepiece) under conditions. of use. applied to the a=bient airborne concentration to estimate the concen-tration inhaled by the wearer according to the following formula: Concentration Inhaled = Ambient Airborne Concentration Protection Factor I (b) The protection factors apply: only for trained individuals wearing properly fitted respirators (1) used and maintained under supervision in a well-planned respiratory protective prograi. 4 for air-purifying respirators only when high efficiency (ii) (above 99.97. removal ef ficiency by U.S. Bureau of Mines (or National Institute of Occupational Safety and Health, type dic ctyl phthalate (DOP) test) particulate as applicable) filters and/or sorbents appropriate to the ha a'rd are used in l at=ospheres not deficient in oxygen. l for at=osphere-supplying respirators only when supplied j (iii) with adequate respirable air. 3 Excluding radioactive contaminants that present an absorption or submersion For tritium oxide approximately half of the intake occurs by hazard. an overall protection f actor of not absorption through the skin so that more than approxi=ately 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 3,
- below, concerning supplied-air suits and hoods.
Not reco= ended for use where it might be possible 4 Under chin type only. for the ambient airborne concentration to reach instantaneous values greater than 50 times the pertinent values in Appendix 3, Table I, Colu=n 1 of 10 CFR Part 20. Appropriate protection factors =ust be determined taking account of the 5 under design of the suit or hood and its permeability to the contaminant No protection f actor greater than 1,000 shall be used conditions of use. except as authorized by the Nuclear Regulatory Commission. 1556 126_. Amendment No. 11 6-24
~~ Ti3LE 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. l 7 Only for shaven faces. l 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 e the extent that they do not exceed the protection factors listed in this Table. The protection factors in this Table =ay 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 (or the Institute of Occupational Safety and Health, as applicable) in accordance with its applicable j 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, li=itations on occupancy may have to be governed by external dose limits. 1556 127 ~ Amendment No. 11 ~ m
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