ML19317G867
| ML19317G867 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/03/1975 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19317G866 | List: |
| References | |
| NUDOCS 8004020528 | |
| Download: ML19317G867 (26) | |
Text
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RANC110 SECO UNIT 1 gg,3,7f,
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TECHNICAL SPECIFICATIONS DEFINITIONS D.
All automatic containment isolation valves are operable or closed in the safety features position.
E.
The containment leakage satisfies Specification 4.4.1, and no kncun changes have. occurred.
1.8 ABNORMAL OCCURRENCE Any of the following:
(1)
Failure of the Reactor Protection System or other systems subject to limiting safety-system settings to initiate the required pro-tective function by the time a monitored parameter reaches the setpoint specified as the limiting safety-system setting in the Technical Specifications or failure to complete the required protective function.
(2)
Operation of the unit or affected systers 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 Specifica-1 tions.
(3)
Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.
(4)
Reactivity anomalies involving disagreement with the predicted value of reactivity briance under steady state conditions greater than er equal to 1.0%AK/K; a calculated reactivity balance 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 suberitical, an unplanned reactivity incertion of more than 0.5%AK/e; or occurrence of any unplanned criticality.
(5)
Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR.
(6)
Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional require-ments of system (s) required to cope with accidents analyzed in the SAR.
(?)
Conditions arising from natural or manmade events that, as a direct result of the event, require plant shutdown, operation of safety systems, or other protective measures required by Technical Specifications.
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RANCllO SECO UNIT 1 TECllNICAh SPECIFICATIONS DEFINITIONS-(8)
Errors discovered in the transient or accident' analyses or in the methods used for such analyses as described in the safety analysis
. report or in the bases for the Technical Specifications that have or could have permitted reactor operation in a manner less con-servative than assumed in the analyses.
(9)
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 analyses in the safety analysis report or Technical Specification bases; or discovery during plant life of conditions not specifi-cally considered in the safety analysis report or Technical Speci-fications that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.
1.9 TIME PERIODS 1.9.1 Shif t - A time period covering 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, three shif ts equally dividing a 24-hour time period.
1.9.2 Daily - A time period within a 24-hour span from 0000 to 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> from Monday through Sunday.
1.9.3 Weekly-A time period from Monday through Sunday spaced to occur 52 times a year.
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RANC110 SECO UNIT 1 TECilNICAL SPECIFICATIONS 4.34 SAFETY SYSTEMS 11YDRAUT.1C SNUBBERS App 11cability Applies to all hydraulic snubber,s installed on safety'related systems.
Objective HTo verify that the hydraulic snubbers will perform their design function.
Specification l.
Those hydraulic snubbers accessible during plant operation shall be inspected for proper reservoir oil level and evidence of Icahage monthly.
Those hydraulic snubbers not accessibic during plant operation l
shall be inspected for proper reservoir level and evidence of lenhage quarterly.
Any units showing evidence of oil leakage shall be dis-l assembled and cause of Icakage determined.
If the cause is determined to be seal degradation, all units exposed to a similar or harsher atmosphere shall be disassembled and scals and fluid replaced in accordance with the manufacturer's instructions.
2.
After the fifth year of operation, at least one unit containing othelene propylene seals and one unit containing polyurethanc seals, which are exposed to'the most adverse environment (radiation and temperatures),
shall be disassembled and seal materi:.1 physical properties evaluated.
if physical properties noted indicate seals are approaching end of useful 1.if e, as recommended by snubber manufacturcr, all scals in units in a similar atmosphere thall be replaced not later than the next refueling interval.
In addition, if seal replacement is necessary, per above requirceents, sample units from areas having less harsh environment shall be disassembled and need for seal replacement determined.
3.
If current test programs or operating experience proves that the continua-tion of this inspection program is not required, this section of the Technical Specifications (4.14) will be deleted.
Justification for this action will be submitted to the AEC in a special report.
Basis This The useful life of the seals and fluids being used is not now known.
of surveillanceprogramwillmonitorsealconditionandverify@abii 0
I D snubbers to perform intended design function.
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o-6.0 ADMlKISTRATIVE CONTROLS 6.1 RESPONSTHTLITY 6.1.1 The Plant Superintendent' chall be responsibic for overall facility operation and shall delegate in writing the succession to this responsibility during _his absence.
6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical _ support shall be as shown on' figurc 6.2-1.
FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figurc 6.2-2 and:
Each on duty shift shall be composed of at least the minimum a.
shift crce cocpacition clean in Tabic 6.2-1.
b.
At Icast one licensed Operator shall be in the control room when fuel is in the reactor.
At least two licensed Operators shall be present in the control c.
room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.
d.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor, ALL CORE ALTERATIONS af ter the initial fuel loading chall be c.
directly supervised by either a licensed Senior Reactor Operator -
or Senior Reactor Operator limited to Fuci Handling who has no other concurrent responsibilitics during this operation.
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DOAftO Or Ollit.CTORS SECitE1 AltY Tite AStlitt f t ACCOUN1 ANT a
MANAGIMfNT fflDCPI f4DEf!T CONSUL T ANTS PLfttODIC EAFFTY AUOli GENE RAL MAN AGCit sal L1Y ItrVILW COMr.u TT E E OF rat 4CHO SECO Ort R AllONS I.
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I PURLIC REL ATIONS PERSONNEL LEGAL STAFF AND IN F ORM AllON DEPARTMENT AS$1ST ANT,
ASSISTANT A%S!STANT ASSISTANT GENERAL MANAGER GENERAL MANAGER GENERAL MAf;AGER GENER AL MAN AGER CONTROLLER TREASURER OPCitATIOf4S CHIEF LNGINEER DISitt:CUTION GENERATION ENGINL E RING ACCOUNilNG PURCHASING OPE R ATIONS
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' DISTRICUTION TRICUTION DATA ARKETING OL ANf ING ENGINEL RING PROCESSING DISTRICU110N NUCLEAR GENERAL FINANCE CONS 1 RUC1 TON /
OPE RATIONS SE RVICES MAINTL f 4 Af4CE CUSTOME R HYDRO IN1EItN AL NSURANCE SERVICES OPE RATIONS '
AUDITOR
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LAfJD ASSURAt:CE SYSTEMS AND AND l'ItOCEDUrtCS REGUL ATIONS l
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PLANT SUPERINTENDENT TECHNICAL ADMINISTRATIVE ASSISTANT
- STAFF ASST. SUPT ASST. SUPT.
MAINTENANCE NUCLEAR PLANT TECilNICAL SUPERVISOR '
OPERATIONS ISL).
SUPPORT SlitFT MAINTE N ANCE SUPERVIRORS (SL)
PERSONNEL SR. CONT ROL ROOM OPERATORS (L)
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CONTROL ROOM OPE R ATORS (L)
CHEMISTRY RADIATION INSTRUMENT-CONT ROL PERSONNEL PERSONNEL 1
AUXILI ARY
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OPE RATORS SL-SENIOR ALC LICENSE L-AEC LICENSE EOUtPMENT
- ROUTINE REPORTING REQUIREMENTS ATTENDANTS ON PERSONNEL Cl' ANGES D
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v FIGURE G.2-2 PLANT OltGANIZATION Cli AllT RANCHO SECO UNIT 1 It ANCllO SECO UNIT 1 TECilNICAL SPECIFICATIONS
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e' RANC110 SECO UNIT 1 TECllNICAL SPECIFICATIONS TABLE 6.2-1 SilIFT CREW PERSONNEL AND LICENSE REQUIREMENTS ***
REACTOR MODE RANC110 SECO JOB TITLE COLD OTilER TilAN SilUTDOUN COLD SilUTDOWN Shift Supervisor 1 - SL 1 - SL Sr. Control Room Operator or Control Room Operator 1-L 2 - L*
Auxilia'ry Operator or Equipment Attendant 1
1 Equipment Attendant or 1
Power Plant Helper Minimum Total Personnci 3
5**
- 0nc licensed operator when the reactor is shutdown gr cater than 1%
AK/K.
- In the event that any mcther of a minimum shif t crew is absent or incapacita*.cd due to illness or injury, a qualified replacement shall be designated to report ensite within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
SL - AEC Senior Licensed Operator L - AEC Licensed Operator
- Until such a time as the startup tests and varranty run are completed, and suf fi:icnt " hot" operator license examinations have successfully completed to provide the shift staffing as indicated on figurc 6.2-1, cach operating shif t shall censist of at least five persons, includ-ing at least one licensed senior reactor operator, one licensed reacter operator, and one or more senior staff nembers from cither the Plant Staff, qualified members of the Generation Engineering Staff, or NSSS vendor's staff or consultants, who, by virtue of their training and experience can provide competent technical support for the startup and power ascension program.
This staff :ncmber will not be requir::d when the reactor in shutdown by more than 1% AK/K nor when the full compliment of three licensed operators is availabic.
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r AlmINISTRATIVE CONTROLS 6.3 FACILITY STAFF OUALTFICATIONS 6.3.1 Each member of the operating staff shall meet or exceed the minimuta qualifications of ANSI N18.1-1971 for comparable positions.
6.4 TRAINING 6.4.1 A retraining and replacement training program for the operating staff shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements and recommendations of Section 5.5 of ANS1 N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.5 REVIEW AND AUDIT 5
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g 6.5.1.1 The Plant Review Committee shall function to advise the Plant Superintendent on all matters related to nuclear safety, t
C0liPOSITION 6.5.1.2 The Plant Review Com:nittee shall be composed of the:
Chairman:
Technical Assistant
?!cmber:
Assistant Superintendent Nuclear Plant Operations
}!cmber:
Assistant Superintendent Technical Support
}!cmber:
}!aintenance Supervisor
}iember:
Chemical and Radiation Supervisor Other members as the Plant Superintendent may appoint from time to time.
6-3
ADMINTSTRATIVE CONTROLS ALTERNATES 6.5.1.3 Alternate members shall be appointed in writing by the Plant Super-intendent to serve on a temporary basis; however, no more than two alternates shall participate in PRC activitics at any one time.
MEETING PREOUENCY 6.5.1.4 The PRC shall meet at least once per calendar month and as convened by the PRC Chairman.
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QUORUM 6.5.1.5 A Quorum of the PhC shall consist of the Chairman and two members including alternates.
RESPONSIBILITIES 6.5.1.6 The Plant Revicu Committee shall be responsibic for:
Pcview of 1) all procedurcs required by Specification ~6.8 and a.
changes thereto, 2) any other proposed procedures or changes thereto as determined by the Plant Superintendent to affect nucicar safety.
b.
Review of all proposed tests and experiments that affect nucicar safety.
Review of all proposed changes to the Technical Specifications.
c.
Review of all proposed changes or modifications to p'lant systeds d.
or equipment that affect nuc1 car safety.
ti Investigation of all violations of the Technical Specifications c.
and shall preparc and forward a report covering evaluation and recommendations to prevent recurrence to the Plant Superintendent and'to the Chairman of the Management Safety Review Committec.
f.
Review of facility operations to detcet potential safety hazards.
Performance of special reviews and investigmtions and reports g.
thercon as requested by the Chairman of the Management Safety Review Committee.
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o AD!!TfliSTRATIVC CONTR07.S RESPONSTHT1.TTTES (Continued) h.
Review of the Plant Security Plan and impicmenting procedures and shall submit recom:nended changes to the Chairman of the 11anagement Saf ety Review Committee.
i.
Review of the Emergency Plan.and implementing procedures and shall submit recommended changes to the Chcirman of the
!!anagement Safety Review Committee.
UTih0RITk' 6.5.1.7 The Plant Review Committee shall:
Recommend to the Plant Superintendent written approval or a.
disapproval of items considered under 6.5.1.6(a) t rough h
(d) above.
b.
Render determinations in writing with regard to whether or not cach item considered under 6.5.1.6(a) through (c) above constitutes an unreviewed safety question.
Provide immediate written notifjcation to the Chairman of tha c.
between the llanagement Safety Review Committee of disagreement PRC and the Ilant Superintendent; however, the Plant Superin-tendent shall have responsibility for resolution of such dis-agreements pursuant to 6.5.1.1 above.
RECORDS _
The Plant, Review Committee shall maintain written minuter. of each 6.5.1.8 raceting and copics shall be provided to the Plant Superintendent and the Chairman of the 14anagement Safety Revicu Committee.
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t!ANAGE!-1ENT SAFETY REVTEU CO!21TTTEE (11SRC) 6.5.2 MHICTIO!!^
The llanagement Safety Review Committec chall function to provide 6.5.2.1 independent rc.ricw and audit of designated activitics in the areas of:
nuclear power plant operations a.
b.
nucicar engineering
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chemistry and radiochemistry d.
metallurgy instrumentation and control c.
[.
rad'iological safety mechanical and cicetrical engineering g,
h.
quality assurance practices e
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a ADMTHISTRATIVE CONTROLS COMPOSITION 6.5.2.2 The MSRC shall be composed of the:
Assistant concral Manager and Chief Engineer - Chairman Technical Assistant - Secretary Manaccr Engineering Department Assistant Cencral Manager, Operations
}!anager, Nucicar Operations Supervising Nucicar Engineer Manager, Cencration Engineering Qual'sy Assurance Director Quality Assurance Engineering ALTERNATES 6.5.2.3 Alternate members shall be appointed in writing by the MSRC Chairman to s,crve on a temporary basis; however, no more than two alternatives shall participate in MSRC activitics at any one time.
CONSULTANTS 6.5.2.4 consultants shall be utilized as d'termined by the MSRC Chaircan e
to provide expert advice to the MSRC.
!!EETINC PREOUE"CY 6.5.2.5 The MSRC 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 thercafter.
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_UORUM Q
A quorum of MSRC sh'll consist of the Chairman or his designated 6.5.2.6 a
alternate and a majority of the liSRC members including alternates.
No more
.than a minority of the quorum shall have line responsibility for operation of the facility.
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. REVIEW 6.5.2.7 The MSRC shall review:
The safety evaluations for 1) changes to procedurcs, equipment n.
or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems which involve an unrevicued safety question as defined in Section 50.59, 10 CFR.
Proposed tests or experiments which involve an unreviewed c.
safety question as defined in Section 50.59, 10 CFR.
d.
Proposed changes in Technical Specifications or licenses.
c'.
Violations of applicabic statutes, codes, regulations, orders, Technical Specifications, licence requirer..cnts, or of internal procedures or instructions having nucicar safety significance.
f.
Significant operating abnormalitics or deviations from normal and.cxpected performance of plant; equipment that affect nucicar snfety.
ABNORMAL OCCURRENCES, as defined in Section 1.8 of these g.
Technical Specifications.
h.
Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components, i.
Reports and meeting _ minutes of the Plant Review Committec.
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9 e-o ADMTNTSTRATTVE CONTROL.S A11DITS 6.5.2.8 Audits of facility activitics shall be performed under the cognizance of the MSRC.
These audits shall encompass:
a.
The conformance of facility operation to all provisions contained within the Technical Specifications and applicable license conditions at 1 cast once per year.
b.
The performance, training and qualifications of the entire facility staff at least once per year.
c.
The results of all actions taken to. correct deficiencies occurring in facility equipment, structures, systems or method of operation that affect nuclear safety at least once per six months.
d,.
The performance of all activitics' required by the quality Assuran:c Program to meet the criteria of Appendix "B",
10 Crit 50, at least once per two years.
The Facility Emergency Plan and implementing procedures at
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least once per two years.
f.
tlc Tacility Security Plan and implementing procedures.at 1 cast once per tuo years.
g.
Any other area of facility operation considc' red appropriate by the MSRC or the General Manager.
A11TITORITY 6.5.2.9 The Mfnc shall report to and advice the General Manager on those areas of responsibility specifica in Section 6.5.2.7 and 6.5.2.8.
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C-ADMinISTnATivn CONTROLS RECORDS 6.5.2.10 Records of MSRC activitics shall he prepared, approved, and distributed as indicated below:
Minutesofeach$SRCmeetingshallbeprepared, approved a.
and forwarded to the General Manager within 14 days following each meeting.
b.
Reports of reviews encompassed by Section 6.5.2.7 c, f, g, and h above, shall be prepared, approved and forwarded to the General Manager within 14 days following completion of the review, Audit reports encompassed by Section 6.5.2.8 above, shall c.
be forwarded to the General Manager and to the management positions responsible for the areas audited within 30 days after completion of the audit.
6.6 ABNORMAL OCCURRENCE ACTION 6.6.1 The following actions shall be taken in the event of an ABNORMAL OCCURRENCE:
The Commission shall be notified and/or a report submitted
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pursuant to the requirements of Specification 6.9.
b.
Each Abnormal Occurrence Report submitted to the Commission shall be reviewed by the PRC and submitted to the MSRC and the Plant Superintendent.
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, ADMINISTRATIVE CO!TTROI,S 6.7 SAFi'TY Lit!IT V101.ATTON 6.7.1 The following actionn shall be taken in the' cvent a Safety Limit in violated:
The provisions of'10 CFR 50.36 (c) (1) (i) shall be complied a.
with immediately.
b.
The Safety Limit violation shall be reported to the Plant Sup -
intendent, the Chairman of the MSRC and to the Commission immed-iately.
A Safety Limir Violation Report shall be prepared.
The report c.
shall be revievnd by the PRC.
This report shall describe (1) applicabic circumstances preceding the violation, (2) effects of the violation upon facility components, systems or structurcs, and (3) corrective acetion taken to prevent recurrence.
d.
The Safety Limit Violation Report shall be submitted to the
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Commission, the MSRC and the Plant Superintendent within 10 days of the violation.
6.8 PROCEDURES 6.8.1 Urftten procedures and administra'tive policies shall be established, implemented and maintained that meet or exceed the requirements and eccommen-dations of Section 5.1 and 5.3 of ANS1 N18.7-1972 and Appendix "A" of USAEC Regulatory Guide 1.33 except as provided in 6.8.2 and 6.8.3 below.
6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PRC and approved by the Plant Superintendent prior to imp 3cmentation and periodically as set forth in each document.
Temporary changes to procedures of 6.8.1 above may be made provided:
6.8.3 The intent.of the original procedure is nat altered.
a.
b.
The change is approved by two members of the plant management staff, cc least one of whom holds a Senior Reactor Operator's License on the unit affccted.
The change is documented, reviewed by the PRC, and approved c.
by the Plant Superintendent within 7' days of implementation.
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ADM1111STPf.TIVF. CONTROLS 6.9 REPORTING REQUIRE!!ENTS ROUTINE AND ADN0iO!AL OCCURRhaCE REPORTS 6.9.1 Information to be reported to the Commission, in addition to the reports required by Titic 10, Code of Federal Regulations, shall be in accordance with the Regulatory Position in Revision 3 of Regulatory Guide 1.10, " Reporting of Operating Information - Appendix "A" Technical Specifi-cations."
SPECIAL REPORTS 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 activitics identified below pursuant to the requirements of the applicabic reference specification:
A.
Startup Report A summary report of unit startup and pouer escalation testing and an evaluation of the results from these test programs shall be submitted uithin 60 days following commenecnent of commercial power operation.
The test results shall be compared with design predictions and specifications.
B.
A Reactor Building structural integrity report shall be submitted within 90 days of completion of each of the follouing tests covered by Technical Specification 4.4.2 (the integrated Icak rate test is covered in Technical Specification above).
.1 Annual Inspection
.2 Tendon stress surveillance
.3 End anchorage concrete surveillance
.4 Liner plant surveillance D
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Inservice inspection Program.
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ADMINISTRATIVE CONTR01.S 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
Records and logs of facility operation covering time interval a.
at each power level.
b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nucletr-safety, c.
ABNORMAL OCCURRENCE Reports.
d.
Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
c.
Records of reactor tests and experiments.
f.
Records of changes uade-to Operating Procedures, g.
Records of radioactive shipments.
h.
Records of sealed souce leak tests and results.
i.
Records of annual physical inventory of all source material of record.
6.10.2 The following records shall be retained for the duration of the Facility Operating License Record and drawing changes reflecting facility design modifi-a.
cations made to systems and equipment described in the Final Safety Analysis Report.
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ADMINTSTRATThE CONTR0!.S' b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
c.
Records of facility radiation and contamination surveys.
d.
Records of radiation exposure for all individuals entering radiation control areas.
Records of gaseous and liquid radioactive material released to c.
the environs.
f.
Records of transient or operational cycles for those facility components designed for a limited nuniber of transients or cycles.
g.
Records of training and qualification for current members of the plant operating staff.
h.
Records of in-service inspections performed pursuant to these Technical Specifications.
i.
Records of Quality Assurance activitics required by the QA Manual.
j.
Records of reviews perforced for changes made to procedures or equipment or reviews of tests and~ experiments pursuant to 10 CFR 50.59.
k.
Records of meetings of the PRC an'd the MSRC.,
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6.11 RADINfION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and chall 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 activitics authorized by the operating license for this facility in deter-mining whether : individuals in restricted arcas arc exposed to concentrations in excess of the limits specified in Appendix B, Tahic I, Column 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 execeded.
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AtefINISTRATTVE CONTn01.S ALLOUANCE (Continued) b.
If the radioactive materini is of such form that intake through
.the skin or other additional route is likely, individual exposuren to radioactive matqrial shall be controlled so that the radioact ive content of any critical organ from all routes of intake averaged over 7 consecutivo 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 per-tinent concentration values provided in Appcudix B, Tabic I, Co.lumn 1,.of 10 CFR 20.
For radioactive materials designated i'Sub" in the " Isotope" column c.
of Appendi,x B, Tabic 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 mat-crials shall be accounted for as part of the limitation on individ-ual dose in 520.101.* These materials shall be subject to applicable process and other engineering controls.
PROTECTION PROGRAM In all operations in which adequate limitation of the inhalation of 6.12.2 radioactive material by the use of process or other engineering controls is impracticable, the liceimcc may permit an individ6a1 in a restricted area to use respiratory protective equipment to limir: the inhalation of airborne radioactive material, provided:
The limits specified in 6.12.1 above, are not exceeded.
a.
Respiratory protective equipment is selected and used so that b.
the peak concentrations of airborne radioactive material inhaled by an indi'idual wearing the equipment do not execed'the pertin-v ent concentration values specified in Appendix B, Tabic I, Column 1, of 10 C1'R 20.
For the purposes of this subparagraph, the concentration of radioactive material that is inhaled when res-pirators arc' worn may be determined by dividing the ambient air-borne concentration by the protection-factor specified in Tabic If the 6.12-1 for the respirator protective equipment worn.
intake of radioactivity is later determined by other measure-ments to have been different than that initially estimated, the later quantity shall be used in evaluating the exposures.
The licensee adviscs cach respirator user that he may leave the arca_at any time for relief from respirator use in case of equip-c.
ment 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 t.hc requirements above are m @
porates V
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- *s ADt!INTSTRATIVE CONTHof.S e
PROTECTIOM PROGRAM (Continued) practices for respiratory protection consintent 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 exposurcs, and to permit proper selection of reapiratory protective equipment.
. '2 Written procedures to assure proper selection, supervision, and training of personnel using such protective equipment.
.3 Uritten procedures to assure the adequate. fitting of res-pirators; and the testing of respiratory protective equip-ment-for operability immediately prior to use.
.4 Written procedures for maintenance to assure full effective-ness of respiratory protective equipment, including issuance, c1 caning and decontamination, inspection, repair, and storage.
.5 Uritten operational and administrative procedures for proper use of respiratory protective equipment including provisions for planned limitations on working times 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.
The licensee shall use equipment approved by the U.S. Bureau of c.
j Mines under its appropriate Approval Schedules as set forth.in i
Table 6.12-1.
Equipment not approved under U. S. Bureau of-Mines Approval Schedules shall be used only if the licensee has evaluated the equipment and can demonstrate by testing, or on the basis of reliable test information, that the material and performance characteristics of the equipment are at least equal to those afforded by U. S. Bureau of !!ines approved equipment of the same type, as specified in Table 6.12-1.
f.
Unicas otherwise authorized by the Commission, the licensee shall not assign protection factors in-excess of those specified in Tabic 6.12-1 in celecting and using respiratory protective equipment.
REVOCATION 6.12.3
- The specifications of Section 6.12 shall be revoked in their entirety
- pon' adoption of-the proposed change to 10 CFR 20, Sectioni.
Ify3p d h would u
0 make such provisions unnecessary.
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ADM1HISTRATIVE COTrROLS 6.13 11TGli RADTATION AREA 6.13.1 In lieu of the " control device" or " alarm signal" required by para-graph 20.203 (c) (2) of 10 C1'R 20:
Each liigh Radiation Area in which the intensity of radiation in a.
greater than 100 mrem /hr but icss than 1000 mrem /hr shall be barricaded and connpicuously posted as a liigh Radiation Arca
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and entrance therer:o shall be controlled by issuance of a Radia-tion Work Permi t a-td 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 liigh Radiation Area in which the intensity of radiation is greater than 1000 mrem /hr shall be subject to the provisions of 6.13.1(a) above, and in addition locked doors shall be provided to prevent unauthorized entry into such area and the keys shall be maintained under the adrainistrative control of the Shif t Supervisor on duty.
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PROTECTION FACTORS FOR RESPIRATORS PROTECTION FACTORS 4 GUIDES TO SELECTION OF EQUIPMENT PARTICULATES AN'D BUREAU OF MINES APPROVAL SCHEDULES
- VAPORS AND GASES FOR EQUIPMENT CAPABLE OF PROVIDING AT EXCEPT TRITIUM LEAST EQUIVALENT PROTECTION FACTORS 3
- or schedule superseding for I
OXIDE DESCRIPTION MODES equipment of type listed AIRFUkIFYINGRESPIRATORS I.
Facepiece, half-mask"'7 NP.,
5 21B 30 CFR S 14.4(b)(4) 23B 30 CFR 514.4(b)(5); 14F 30 CFR 13 Facepiece, full NP 100 7
II. AT:!0 SPHERE-SUPPLYING RESPIRATOR 1.
Airline respirator -
19B 30 CFR 512.2(c)(2) Type C (i)
Facepiece, half-mask CF 100 Facepiccc, full CF 1,000 19B 30 CFR 512.2(c)(2) Type C (i) 7 D
100 19B 30 CFR 5 12.2(e)(2) Type C (ii)
Facepiccc, full 19B 30 CFR S 12.2(c)(2) Type C (iii)
Facepiece, full PD 1,000 5
6 Hood CF 5
6 Suit CF 2.
Self-contained breathing apparatus (SCBA) 13E 30 CFR S ll.4(b)(2)' (i)
Facepiece, full D
100 7
Facepiece, full PD 1,000 13E 30 CFR 5 11.4(b)(2) (ii)
Facepiece, full R
100 13E 30 CFR 5 ll.4(b)(1)
III. COMBINATION RESPIRATOR Protection factor 19B CFR 5 12.2(e) or applicable e
Any combination of air-for type and mode Schedules as listed above purifying and atmosphere-of operation as supplying respirator listed above 1,
2, 3,
4, 5,
6, 7
[These notes are on the followinG pages}
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TABl.E 6.9-1 (Continued)
D l See the following symbols:
%0gn q
U A
'@6 MLl d-CF:
continuous flow.
D:
demand NP: negative pressure (it c., negative _ phase during inhalation)
PD:
pressure demand (i.c., always positive pressure)
. R:
rceirculating (closed circuit).
2 (a)' For purposes of this specification the protection factor.is a measure of the degree of protection afforded by a respirator, defined as the ratio of the concentration of~ airborne rad 4.,-
active material outside the respiratory protective equipment l
to that inside the-equipment (usual 3y inside the facepiece) under conditions of ucc.
It is applied to the ambient air-borne cencentration to estimate the concentration inhaled by the wearer according to the following formula:
Concentration Inhaled = An icnt Airborne Concentration
~
l-Protection Factor i
(b) The protection factors apply:
l I
(i) only for trained individuals wearing properly fitted respirators used and maintained'under supervision in,
a well-planned respiratory protective prograu.
(ii) for air-purifying respirators only when high efficiency
[above 99.9% removal efficiency by U. S. Bureau of. Mines type df. octyl phthalate (DOP) test] particulate filters.
and/or'sorbents appropriate to the hazard are used in atmosphercs 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 sub-mersion 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..
Under chin type only.
Not recommended for use where it might be possibic
. for the ambient airborne concentration to reach instantaneous values igrcntor-than 50 times the pertinent values in Appendix B, Tabic 1, Column 1 of 10 CFR Part 20.
6-19.
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TABT.E 6.9-1 (Continued) 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 usc.
No protection factor greater then 1,000 sha_1 be used except as authorized by the Commission.
6 No approval schedules current,1y availabic for this equipnent.
Equip-ment must be evaluated by testing'or on basis of availabic test infor-
,mation.
7 Only for shaven faces.
NOTE 1:
Protection factors for respirators, as may be approved by the U.S. Burcou of Mines recording 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 Tabic.
The protection factors in this Table may not he 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. B arcau of Mines in accor-dance with its applicabic schedules.
NOTR 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.
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TECllNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.2.7 The control rod drive patch pancis shall be locked at all times wit h i fmited accesstobeauthorizedbythesuperintendentorhisdeafgnatedrepresentative.@
Bases The power-imbalance envelope defined in figure 3.5.2-2 is based on LOCA analyses which have defined the maximum linear heat rate (see figure 3.5.2-3) such that the maximum clad terperature will not exceed the Interim Acceptance Criteria.
Operation outside of the pouer imbalance envelope alone does not constitute a situation that would caunc the Interim Acceptance Criteria to be exceeded should a 1.0CA occur.
The power imbalance envelope reprocents the boundary of operation limited by the Interim Acceptance Criteria
.only if the control rods are at the position limits as defined by figure 3.5.2-1 and if a 4 percent quadrant power tilt exists.
Additional conservatism is in*roduced by epplicatioa of:
1 A.
Nuclear uncertainty factors.
B.
Thermal calibration uncertainty.
C.
Fuel densification effects.
D.
Hot rod manufacturing tolcrance factors.
'The 30 percent overlap between successive control r'od' groups is allowed since the worth of a rod is Iv.cr at the uppc' and lower port of the strohc.
Control rods are arranged in groupe or banks defined as follows:
Group Function mo' D
ts es 1
Safety l
F rgn}(
2 Safety 7 p 3
Safety 4
Safety 6
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5 Regulating 6
Regulating Regulating 7
ApSR (axial power chaping bank) 8 Control rod groups are withdrawn in sequence beginning with group 1.
Group 5 is The normal overlapped 25 percent with groups 6 and 7, which operate in parallel.
position at power is for groups 6 and 7 to be partially inserted.
The minimum-available rod worth provides for achieving hot shutdown by reactor trip at any time assuming the highest worth control rod remains in the full out position. (1)
Inserted rod groups during power operation will not contain single rod worths greater than 0.65 percent ak/k.
This value has been shown to be safe by the safety analysis of the hypothetical rod ejection accident. (2)
A singic inserted control rod worth of 1.0 percent Ak/k at beginning of life, hot,-zero power would result in the same trannient peak thermal power and therefore the same rated environmental conscquences as a 0.65 percent 6k/k ejected rod worth at power.
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6 TABLE 3.5.1-1 (Concluded)
(
INSTRGENTS OPERATING CONDITIONS
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(A)
(3)
Minimum Operable Minimum Degree Operator Action if Conditions of Functional Unit Channels of Redundancy Colurns A and 3 Cannot be Met Safety Features
- 4.
Reactor Building spray valve a.
Reactor Building ^ pres-2 1
Bring to hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> g
sure instrument channel m
2 1
Bring to hot shutdown.
O_
b.
Manual Pushbutton within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
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N.A.
Bring to hot shutdown
[.
1.
Pressurirer Level within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> gg oo 9
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- 1f minimum conditions are not met within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after hot shutdown, the unit shall be placed in h -i a cold shutdown condition within ar. additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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