ML20002C525
| ML20002C525 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/13/1975 |
| From: | Lamley R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20002C521 | List: |
| References | |
| NUDOCS 8101100412 | |
| Download: ML20002C525 (41) | |
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COMSUMERS POWER COMPANY l-LS Ts f..
l Docket No 50-155 Request for Change to the Technical Specifications License No DPR-6 For the reasons hereinafter set forth, the following changes to the Technical Specifications'of License No DPR-6 issued to Consumers Power Company on May 1, 1964 are requested:
1.
Changes Delete Section 6, " Administrative Controls," from the proposed Technical Specifications submitted June 7, 1974 and, in its place, insert the following:
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- 2 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Superintendent shall be responsible L overall plant operation and shall delegate in writing the succession to this responsibility during his absence.
6.2 ORGANIZATION OFFSITE P'
6.2.1 Ti:e offsite organization for plant management and technical support shall be as shown on Figure 6.2-1.
PLANT STAFF 6.2.2 The plant organization shall be as shown on Figure 6.2-2 and:
Each on-duty shift shall be composed of at least the minimum shift crew composition shown a.
i in Table 6.2-1.
b.
At least one licensed Operator shall be in the control room when fuel is in the reactor.
c.
At least two licensed Operators shall be present in the control room during r actor start-up (approach to critical), scheduled reactor shutdown and during recovery from reactor trips, d.
All core alterations after the initial fuel loading shall be directed by a licensed Reactor Operator or Reactor Operator Limited to Fuel Handling who has no other concurrent responsi.
bilities during this operation and under the general supervinion of either a licensed Senior Reactor Operator or a Senior Reactor Operator Limited to Fuel Handling.
6.3 PLANT STAFF QUALIFICATIONS 6.3.1 Each member of the plant staff shall meet or exceed the minimum qualificati ns of ANSI N18.1-1971 i
for comparable positions.
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CONSUMERS POWER COMPANY 0FF-SITE ORGANIZATION i
CHAIRMAN OF THE BOARD AND PRESIDENT
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EXECUTIVE VICE PP.E.S IDENT SAFETY AND AUDIT 4
VICE PRESIDENT REVIEW BULK POWER OPERATIONS BOARD MANAGER OF BULK DIRECTOR MANAGER
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POWER PRODUCTION QUALITY TECHNICAL SERVICE
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ASSURANCE ELEC PRODUCTION l
SUP - NUCLEAR QA ADMINISTR4-N'}
PLANT NUCLEAR 1
SUPERINTENDENTS TRAINING (NUCLEAR)
ADMINISTRATOR FIGURE 6.2-1 w
m CONSUMERS POWER ' COMPANY BIG ROCK POINT Organization QA-ADMIN OFF-SITE I
PLANT SUPERINTENDENT i
PLANT REVIEW COMMITTEE QA ENGINEER CLERKS OPERATIONS ENGINEER TECHNICAL ENGINEER MAINTENANCE cNGINEER.
SECURITY REACTOR ENGINEER IEC SUPERVISORi ENGINEER TRAINING S.S.,
SUPERVISOR I
N INEER PLANT TECHNICIANSi MAINTENANCE SUPERVISORi PLANT SHIFT SUPERVISORS ASST MAINTENANCE SUPERVISOR i
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l OPERATORS I
I CHEM & RAD PROT SUPV ENGINEEP.
REPAIRMEN l
ST0cxxEN
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CHEM &R$bROTSUPVR
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CHEM & RAD PROT T EC HN I C I ANS
FIGURE 6.2-2 6-3 u-n i
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TABLE 6.2-1 Minimum Shift Crew Composition 1
Operating 1 Shift Supervisor - SOL 1 Crerator - OL 7',
t 2 Operators - Nonlicensed Cold Shutdown 1 Shift Supervicer - SOL 1 Operator - OL 1 Operator - Nonlicensed Refueling Operations 1 Shift Supervisor - SOL 1 Operator - OL 2 Operators - Nonlicensed SOL - Senior Reactor Operating License OL - Reactor Operating License (1)During control rod motion associated with reactor start-up (approach to critical), one licensed Operator shall observe the control rod manipulation to ensure established control rod withdrawal Procedures are adhered to.
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6.h TRAINING 6.4.1 A retraining and replacement training program'for the plant staff shall be maintained'under.
the direction of the Nuclear Training Administrator and shall meet or exceed the requirements-and recommendations of Section 5 5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR,'Part'55
.9.
1 6.5 REVIEW AND AUDIT 9;
6.5 1 PLANT REVIEW COMMITI'EE (PRC)
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6.5 1.1 FUNCTION i
The Plant Review Committee (PRC) shall function to advise the Plant Superintendent on all matters related to nuclear safety.
6.5 1.2 COMPOSITION The PRC shall be composed of the:
J Chairman: Plant Superintendent Member:
Operations Engineer Member:
Technical Engineer Member:
Maintenance Engineer Member:
Plant Instrument and Control Supervisor Member:
Reactor Engineer Member:
Chemistry and Radiation Protection Supervisor a
j Member:
Quality Assura.Me Engineer Member:
Shift Supervisc? (One)
Member:
Engineer With a.
Least One-Year Plant Experience Member:
Training Supervisor 6.5 1.3 ALTERNATES 1
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Alternate. members shall be appointed in writing by the PRC Chairman to serve'on a. temporary i
basis; however, no more than two alternates shall partic pate:in PRC activities at any one time.
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6.5.1.4 MEETING FREQUENCY The PRC shall meet at lea once per calendar month with special PRC meetings as required.
6.5.1 5 QUORUM i
A quorum of the PRC shall consist of the Chairman and'four members (including alternates).
6.5.1.6 RESPONSIBILITIFE The PRC shall be responsible for:
Review of (1) all procedures required by Specification 6.8 and changes thereto, a.
(2) any other proposed procedures or changes thereto as determined by the Plant Superintendent to affect nuclear safety, b.
Review of all proposed tests and experiments that affect nuclear safety.
c.
Review of all proposed changes to the Technical Specifications.
d.
Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.
e.
Investigation of all violations of the Technical Specifications. A report shall be prepared and forwarded covering evaluation and recommendations to prevent recurrence to the Electric Production Superintendent-Nuclear and to the Chairman of the Safety and Audit Review Board (SARB).
f.
Review of plant operations to detect potential safety hazards.
g.
Performance of special reviews and investigations and reports thereon as requested by the Chairman of the SARB.
h.
Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Emergency Plan to the Chairman of the SARB.
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l 6.5 1.7 AUTHORITY The PRC'shall:
a.
Recommend to t' e Plant Superintendent written approval or disapproval of items considered under 6.5 1.6(a) through (d) above.
t b.
Render determinations in writing with regard to whether or not each item considered-under 6.5 1.6(a) through (e) above constitutes an unreviewed safety question.
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c.
Provide immediate written notification to the Electric Production Superintendent-Nuclear' and the Chairman of SARB of disagre. ment between the PEC and the Plant Superintendent.
However, the Plant Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above.
6.5.1.8 RECORDS The PRC shall maintain written minutes of each neeting and copies shall be provided to the Electric Productfon Superintendent-Nuclear and the Chairman of SARB.
6.5.2 SAF M AND AUDIT REVIEW BOARD (SARB) 6.5.2.1 RESPONSIBILITIES SARB is responsible for maintaining a continuing examination of designated plant activities.
In all cases, where a matter is formally considered by SARB, its findings and recommendations are communicated in writing to the Vice President - Bulk Power Operations (BPO) and other appropriate levels of management. A written charter is prepared and approved by the Vice President - BPO which designates the membership, authority and rules for conducting the meetings. Board membership, qualifications, meeting frequency, quorum, responsibilities, authority and records are in accordance with the nuclear plant Technical Specifications and ANSI N18.7-1972.
6Property "ANSI code" (as page type) with input value "ANSI N18.7-1972.</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..5 2.2 FUNCTION The SARB shall function to provide independent review or designated activities affecting safety-related components, systems and structures designated on the plant's Safety-Related Quality List contained in the Consumers Power Company Quality Assurance Program.
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m 6.5 2.3 COMPOSITION AND QUALIFICATIONS Collectively, the personnel appointed for the_SARB by the Vice President - BPO shall be competent to conduct reviews and technical audits.in the following areas:
a.
Nuclear power plant operations.
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b.
Nuclear engineering.
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Chemistry and radiochemistry.
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d.
Metallurgy.
e.
Instrumentation and control.
f.
Radiological safety.
g.
Mechanical and electrical engineering.
h.
Qv111ty Assurance practices.
An individual appointed to the.SARB may possess expertise in more than one of the above specialties. He should, in general, have had professional experience at or above,the senior engineer level in his specialty.
6.5 2.3 ALTERNATE MEMBERS
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Alternate members may be appinted by the Vice President - EPO to act in place of members during any legitimate and unavoidable absences including a conflict-of-interest determination.
The qualifications of alternate.n.mbers shall be similar to those members for whom they will substitute.
6.5 2.h CONSULTANTS Consultants shall be utilized as determined by the CARB members and/or chairman'to provide expert advice to the SARB. SARB members are not restricted as to sources of technical input and may call for separate investigation from any competent source.
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i 6.5 2 5 MEETING FREQUENCY-The SARB shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at ler.st once every six months thereafter, i
6.5.2.6 QUORUM A quorum of SARB shall consist of the Chairman or his designated alternate and four (h)
- r. embers or their alternates. No more than a minority of the quorum shall have line responsibility for operation of the facility.
It is the responsibility of the Chairman to ensure that the quorum convened for a SARB meeting contains appropriately qualified
, f" members or has at its disposal consultants sufficient to carry out the review functions
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required by the meeting agenda.
6.5 2 7 REVIEW The SARB shall review:
a.
Proposed tests or changes to procedures, e'tuipment, systems which are deemed to involve an unreviewed safety question as defined in 10 CFR 50 59 b.
Proposed changes in Technical Specifications or licenses.
Significant operating abnormalities or deviations from normal and expected performance c.
of plant equipment that affect nuclear safety.
d.
ABNORMAL OCCURRENCES, as defined in Section 6.9.1.2.a of these Technical Specifications and other violations of applicable statutes, codes, regulations, orders, Technical Specifica-tions, license requirements or of internal procedures or ' instructions having nuclear safety significance.
e.
Reports and meeting minutes of-the PRC.
f.
Operational and major modification Quality Assurance Program audit reports, g.
Technical audit reports.
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.The ste.tus of deficiencies identified by the Quality' Assurance Program,' including t'he
-i effectiveness of the corrective actions completed and implemented, at least once. every six (6) months.
6.5.2.8 AUDITS 4
Audits of safety-related facility activities.during operations we ' performed by the ' Quality' Assurance Department - BPO in accordance with the pol'cies and procedures of the. Consumers Power Company Quality Assurance Program. Quality assurance audit reports are sent to SARB for review.
In addition, technical audits are the responsibility of the Technical Services Depart-O ment and shall be reviewed by SARB. These technical audits encompass:
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The conformance of facility operation to all provisions contained within the Technical'-
Specifications and applicable license conditions at least once per year.
b.
The performance, training and qualifications of the entire facility staff at least once per year.
c.
The facility Site Emergency Plan and implementing procedures at least once per two-years.
d.
Any other area of facility operation considered appropriate by SARB or the Vice-President - BPO.
6.5.2.9 AUTHORITY SARB shall report to and advise the Vice President - BPO on those areas of responsibility
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specified in Sections 6.5.2.7 and 6.5.2.,;.
6 5 2.10 RECORDS Records of SARB activities shall be prepared and distributed"as indicated below:
a.
Minutes of each SARB meeting shall.be prepared and forwarded.to the Vice President - BPG within fourteen (lk) days following eachsmeeting.
b.
If not included in SARB meeting minutes, reports of reviews encompassed by 'Section 6.5 2.T(e),
j (f), (g) and (h) above, shall-be prepared and forwarded to the Vice President - BPO within l
fourteen (14) days following completion of the review.
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c.
Audit reports encompassed by Section u.5.2.8 above, shall be forwarced to the-management positions responsible for the areas audited within thirty (30) days after completion of the audit.
3 6.6 ABNORMAL OCCURRENCE ACTION 1
6.6.1 The following actions shall be taken in the event of an ABNORMAL OCCURRENCE (AO):
a.
The Comission shall be notified and/or a report submitted pursuant to the requirements
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of Specification 6.9 t
b.
Each A0 shall be reviewed by the PRC. The results of the PRC review shall be submitted
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(either by PRC minutes or by separate report) to SARB and the Electric Production Superintendent-Nuclear.
6.7 SAFETY LIMIT VIOLATION 6.T.1 The following actions shall be taken in the event a Safety Limit (refer to Section~ 2.0 Column titled " Safety Limits") is violated:
The reactor shall be shut down immediately and not restarted until Commission approval a.
is received [10 CFR 50 36(c)(1)(1)].
b.
The Safety Limit violation shall be immediateIy reported to the Commission in accordance with Section 6.9, the Electric Production Superintendent-Nuclear and to SARB Chairman or Vice-Chairma:.-
c.
A report shall be prepared in.accordance with Section 6.9 The Safety Limit violation and the report shall be reviewed by,the PRC.
d.
The report shall be submitted to the Commission (in accordance with the requirements of l
Section 6.9), SARB and the Electric Production Superintendent-Nuclear.
i 6.8 PROCEDURES 6.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the requirements and recommendations of Sections 51 and 5 3' of ANSI N18.7-1972 6-11 j
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(Rev 1, Draft 3 dated November 12,1973).and Appendix'"A" of.USAEC. Regulatory Guide '1 33 (Nov 1972, excluding Plant Security Procedures) except as provided.in 6.8.2' and 6.8.3 below.
i 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 implementation.
t 6.8.3 Temporary changes to procedures of 6.8.1 above may be ~made provided:
a.
The intent of the original procedure is not altered.
y-b.
The change is approved by two members of the PRC, et least one of whom holds a Senior c
Reactor Operator's License.
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The change is documented, reviewed by the PRC at the next regularly scheduled meeting and approved by the Plant Superintendent.
6.9 REPORTING REQUIREMEIES 6.9 1 ROUTINE AND ABNORMAL OCCURRENCE REPORTS i
Information to be reported to the Commission, in addition to the reports r quired by Title 10, Code of Federal Regulations, shall be submitted to the Director, Directorate of Regulatory.
i Operations, Region III. The reports submitted shall be as follova:
6.9 1.1 ROUTINE REPORTS a.
Start-Up Report A summary report of plant start-up and power escalation testing shall' be submitted fol--
loving (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different' design or has been manufactured by La different fuel supplier, and (h) modifications that may have significantly ' altered the nuclear, thermal, or hydraulic performance of the plant.- The i
report shall address each of the tests conducted and shall include rm -. ription of the neasured values :of the operating conditions or characteristics obtained during the test i
program and a comp *
~on of these values with design predictions and specifications..Any j
corrective' actions chat were required to obtain satisfactory opernion shall also be deseribed.
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Start-up reports shall be submitted within (1) ninety (90) days followin6 completion of the start-up~ test program, (2) ninety (90) days following resumption or commencement of commercial' power operation, or (3) nine (9) months following initial' criticality, which-ever is earliest.
If the start-up report.does not cover all three events.(ie,Linitial t criticality, completion of start-up test program, and resumption or commencement of com-mercial power operation), supplementary reports shall be submitted' at least every three.
(3) months until all three events have been' completed.
b.
Annual Operating Report Routine operating reports covering the operation of the unit. during the previouc calendar.
l year shall be submitted prior to March 1 of each year. The initial report shall be sub-'
mitted prior to March 1 of the year following initial criticality.
The annual operating reports made by licensees shall provide a comprehensive summary of the operating experience. gained during the year, even though some repetition of previously reported information may be involved.
References in the an.1ual operating report;to pre-viously submitted reports shall be clear.
Each annual operating report shall include:
(1) A narrative senmary of operating experience during the report period ' relating'to.
safe operation of the facility.
4 (2) For each outage or forced reduction in power" of over 20 (15 MWe) percent of design i
power level where the. reduction is for greater than four hours.
(a) The proximate cause and the system and major component involved'(if the outage or forced reduction 'in power involved equipment malfunction);
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(b) A brief discussion of (or reference to reports of) any A0s pertaining to the outage or power reduction; "The term " forced reduction in power" is normally defined in the electric power industry as the occurrence of a component. failure or other condition which requires that the.lcad on the unit be reduced for cor-rective action immediately or up to and including the very next weekend. Note that routine preventive maintenance, surveillance, and calibration activities requiring power reductions are not covered by this section.
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(c) Corrective action'taken to reduce the probability of recurrence, if appropriate; (d) Operating time lost as a result of the outage or power reduction (for forced outages,5 use the generator off-line hours; for forced reductions in power, use.
the approximate duration' of operation at reduced power);
..1 (e) A description of major safety-related corrective maintenance performed during the outage or power reduction, including the system and component. involved and identi-fication of the critical path activity dictating the length of the outage or power reduction; and P
(f) A report of any unusual releases of radioactivity or unusual radiation exposures specifically associated with the outage.
(3) A tabulation (supplementing the requirements of 5 20.h07 of 10 CFR, Part 20) of the number of personnel receiving ' exposures greater than 100 mrem in the reporting period according to duty function, eg, routine plant maintenance, special plant maintenance (describe maintenance), routine fueling operation, special refueling operation (de-scribe operation), and other job-related exposures.
Estimates of the dose assignment t
to various duty functions shall be based on pocket dosireters, TLD or film badge-measurements. Small exposures totaling less than 20 percent of the total dose need not be individually accounted for.
However, in the aggregate, at least 80 percent of the total whole body dose received from external sources shall be assigned to specific work functions.
(h) A report on fuel performance, as follows:
(a) A tabulation on a vetkly basis of off-gas data reported as the sum bf the six principal fission gas nuclides (Xe-133, -135, -138, Kr-85m, -87, -88) sampled at the steam jet air ejector and corrected 'for holdup time prior to release.
The reactor power level at the time of the sampling should be recorded with the sample result.
5The term " forced outage" is normally 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 correc-tive action immediatt.y or up to and including the very next weekend.
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(b) A tabulation of primary coolant sample results for. iodines following any short-term (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less) increase in the steam jet air ' ejector off-gas rate when the base exceeds 5,000 pCi/s and the spike causes at least a factor of 2 increase.-
Such samples should be' obtained within h hours after the time of determination that the spike occurred.
1 (c) All significant findings. from failed fuel examinations.
(5) Effluent Release Data A report on effluent date, summarized quarterly (except in. instances when more-detailed data are needed) as follows:
(a) Gaseous Effluents 1.
Gases a.
Total curies of fission and activation gases released.
b.
Average release rates (pCi/s) of fission and activation gases for the quarterly periods covered by the report.
c.
Percent of Technical Specifications limit for releases of fission and activation gases.
d.
Quarterly sums of total curies for each of the radionuclides deter-I mined to be released, based on analyses of fission and activation gases.
ii.
Iodines a.
Total curies of each of the isotopes, Iodine-131, Iodine-133, and Iodine-135 determined to be released.
b.
Average release rate (pCi/s) of Iodine-131.
c.
Percent of Technical Specifications limit for Iodine-131.
iii. Particulates a.
Total curies of radioactive mat trial in particulate form with half-lives greater than 8 days det(211ned to be released.
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Average release rate (pCi/s) of radioactive material in perticulate form with half-lives greater than 8 days.
c.
Percent of Technical Specifications limit for radioactive material in particulate form with half-lives greater than 8 dcys.
d.
Total curies for each of the radionuclides in particulate form 1
determined to be released based on analyses performed.
e.
Total curies of gross alpha radioactivity determined to be released.
iv.
Tritium Total curies of tritium determined to be released in gaseous ' effluents. ~
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a.
b.
Average release rate (pCi/s) of tritium, c.
Percent of appropriate Technical Specifications or MPC limits for tritium.
(b)
Liquid Effluente 1-1.
Mixed Fission and Activation Products a.
Total curies of radioactive material determined to be released in liquid effluents (not including tritium, dissolved and entrained ca.ses, and alpha-emitting material).
b.
Average concentrations (pCi/ml) of mixed fission and activation a
products released to unrestricted areas, averaged over the quarterly periods covered by the report.
Percent of applicable limit of average concentrations released to c.
unrestricted areas.
d.
Quarterly sums of total curies for each of the radionuclides determined -
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to be released in liquid effluents, based on analyses performed.
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Tritium Total curies of tritium detemined to be released in liquid effluents.
a.
b.
Average concentrations (pCi/ml) of tritium released in liquid efflu-i ents to unrestricted areas, avera6ed over the quarterly periods covered
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by the report.
i c.
Percent 'of applicable limit of average concentrations released.to unre-stricted areas.
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Dissolved and Entrained Gases Total curies of gaseous radioactive material detencined to be released a.
in liquid effluents.
b.
Average concentrations (pCi/ml) of dissolved and entrained gaseous ~
3 radioactive material released'to unrestricted areas, averaged over i
the quarterly periods ' covered by the report, c.
Percent of Technical Specifications limit of average concentrations released to unrestricted areas.
d.
Total curies for each of.the radionuclides determined to be released as dissolved and entrained gases in liquid effluents.
iv.
Alpha Radioactivity v
Total curies of gross alpha-emitties material determined.to be released in liquid effluents.
v.
Volumes Total measured volume (liters), prior to dilution, of. liquid
-a.
effluent released.
b.
Total determined volume, in liters, of dilution water used during the period of the report.
(c) Solid Wastes
- i. The total quantity in cubic meters and'the total estimated radioactivity in curies for the categories or types of waste.
a.
Spent resins, filter sludges, evaporator bottoms; b.
Dry compressible waste, contaminated equipment, ete; c.
Irradiated components, control rods, etc;~
d.
Other (furnish description).
ii. The disposition of solid waste shipments.
(The number of shipments, the mode of transport, and the destination. )
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e 111. The disposition of irradiated fuel shipments.
(Identify the number of shipments, the mode of transport, and the destination. )
(d) Radiological Impact on Man I
Potential doses to individuals and populations will be calculat'ed using mea-sured effluent and averaged meteorological data.
1.
Total body and significant organ doses (greater than 1 millirem) to indi-viduals in unrestricted areas from receiving water-related exposure pathways.
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The maximum offsite air doses due to beta and gamma radiation at locations near ground level from gaseous efflue'nts, iii. Organ doses (greater than 1 millirem) to ir.dividuals in unrestricted areas from radioactive iodine and radioactive material 'in particulate form from the major pathways of exposure.
iv. Total body doses (greater than 1 manRem) to the population 'and average doses (greater than 1 millirem) to individuals in the population from receiving water-related pathways.
v.
Total body doses to-the population and average doses to individuals in the population from gaseous effluents to a distance of 50 miles from the site.
(6) Radiological-Environmental Monitoring Data (a) Content Descriptive material covering the offsite environmental surveys performed during the reporting period shall include information on:
1.
The number and types of samples taken, eg, airborne dust, milk, lake water.
ii. The number and types of measurements made, eg, gross beta.
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111. Locations of the sample points' and monitoring. stations.
iv.
The frequency of. the surveys.
v.
A summary of survey results.
(b) Significant Levels of Radioactivity.
If samples or measurements averaged over appropriate data base indicate
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statistically significant levels of radioactivity above established or concur-rent backgrounds, the following information shall be provided:
- i. The type of analysis peri'rmed, eg, alpha, beta, gamma and/or. isotopic.
ii. The minimum senritivity of the monitoring system.
iii. The measured radiation level of umple concentration.
iv.
The specific times when samples were taken and measurements were made.
v.
An estimate of the likely resultant exposure to the public.
c.
&nthly Operating Report Routine reports of operating statistics and shutdown experience should be submitted on a monthly basis. The report formats set forth in Appendices "C," "D" and "E" to Regulatory Guide 1.16, Revision 2, shall be completed.in accordance with the instructions provided.
The completed forms shall be sent' to s the Director of Regulatory Operations, US Atomic Energy Commission, Washington, DC 20545, with a copy to Region III Directorate of Regu-latory Operations, to arrive no later than the tenth of each month following the calendar month covered by the report.
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l 6.9 1.2 ABNORMAL OCCURRENCF' REPORTS a.
Pronpt Notification With Written Follow-Up The types of events listed belov shall be reported as expeditiously as possible, but within
- 1 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director, Directorate of Regulatory Operations, Region III, or his designate, no later than the first working day following the event.. A written follow-up report shall be sub-mitted within two (2) weeks of the event. The written report shall include, as a minimum, a completed copy of the transcription sheet (see Appendix "A" to Reeulatory Guide 1.16, Revision 2) used for entering data into the AEC's computer-based file of information con-cerning k).
Information provided on the transcription sheet shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circum-stances surrounding the event.
(1) Failure of the reactor protection system to trip, as required, by the time a monitored parameter reaches the set point specified as the' limiting safety system setting in the Technical Specifications.
Note:
Instrument drift discovered as a result of testing need not be 1eported under this item (but see Items 6 91.2.a(5) and (6), and 6.91.2.b(1) belov].
(2) Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative-aspect of the limiting condition for operation established in the Technical Speci-fications.
Note: If specified action is taken when a system is found to be operating between the most conservative and the 'least conservative. aspects of a limiting condition for operation listed in the Technical Specifications, the limiting condition for operation is not considered to have been violated and no report need be submitted under this section [but see Item 6.91.2(b)(2) belov].
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(3) Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment.
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(h) Reactivity Anomalies (a) Discovery of disagreer.ent with predicted value of reactivity balance greater.
than or equal to 1% ok/k.
S (b) A projection of a reactivity balance that would threaten the ability to attain required shutdown margin.
(c) Short-term reactivity increases that cor*espond to a reactor period of less than 5 seconds, or it suberitical, an unplanned reactivity insertion of more than 0 5% ak/k.
t (5) Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to function to cope with accidents analyzed in the Final Hazards Summary Report (FHSR).
(6) Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to function to cope with accidents analyzed in the FHSR.
Note: For Items 6.9 1.2.u(5) and (6), reduced redundancy that does not result in loss of system function need not be reported under this section [but see Items 6.9 1.2.b(2) and (3)].
(7) Conditions arising from natural or man-made events that, as a direct result of the event, require plant shutdown, operation of safety systems, or other protective mea-sures required by Technical Specifications.
(8) Errors in the transient or accident analyses or in the methods used for such analyses as described in the FilSR or in the bases for the Technical Specifications that have or could have permitted reactor operation in a manner lesr conservative than assumed in the analyses.
(9) operation in a manner less conservative than assumed in the FESR or Technical Specifi-cations (including bases) or discovery during plant life of conditions not considered in the FHSR or Technical Specificstions that require remedial action or corrective measures to prevent the existence of an unsafe condition.
6-21
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g b.
Thirty-Day Written Reports The AO discussed below have lesser immediate importance than those described under "a" above. Such events should be the subject of written reports (transcription sheets, see Appendix "A" to Regulatory Guide 1.16, Revision 2) to the Director, Directorate.of Regu-latory Operations, Region III.
A0 reports submitted in.accordance with this section shall be submitted within thirty (30) days of discovery.
(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 fulfillment of the functional requirements of affected '
- r systems (but see Items 6.91.2.a(1) and (2) above].
(2) Conditions leading to operation in a degraded mode permitted by a limiting condition of operation, or plant shutdown required by a limiting condition of operation.
o Note: Routine surveillance testing, instrument calibration or preventative mainte-nance which require system configurations as described in Items 6.91.2.b(1) and 6.91.2.b(2) above need not be reported except where test results are not satisfactory.
(3) Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protec-tion systems or engineered safety feature systems [but see Item 6 91.2.a(6) above).
6.9 2 SPECIAL REPORTS 6.9 2.1 Special reports shall be submitted to the Director, Directorate of Regulatory Operations, Region III within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
I a.
Containment leak rate tests (90 days).
b.
Inservice inspection reports (150 days).
i c.
Materials radiation surveillance specimen reports (90 days after completion of specimen testing).
6-22 4
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l' 6.9 3 EVENT OF POTENTIAL PUBLIC INTEREST The types of events listed below are frequently of high. public interest; while some' of the events ~
may not be reportable as defined in these Technical Specifications, every effort should be made to keep the Directer, Directorate of Regulatory Operations, Region III, or his designate informed.
i of such events' by telephone as soon as possible after the event has been' discovered:
a.
An event that causes damage to pmperty or equipment, when such damage affects the power production capability of the facility.
g.
b.
Radiation exposure to licensee personnel or members of the public in excess of
- r.
- applicable exposure limits set forth in 10 CFR 20.
c.
Natural or man-made conditions that may require action that need not be reported
. under Section 6 9 1.2.a(6) above.
l d.
Discovery of significant radiological event offsite occurring during transport of t
material for which the licensee was shipper or intended receiver.
e.
Unscheduled shutdowns. expected to last more than one week regardless of cause.
f.
Unusual low-level releases of radioactive material from the site boundary not reportable under other requirements.
l t
g.
Failure of or damage to safety-related equipment which need not be reported under i
Section 6 91.2.a above, if the time of the repair is likely to exceed the time allowed by the Technical Specifications.
6.10 RECORD RETENTION (Records not previously required to be retained shall be retained as required below l
commencing January 1, 1976.)
r 6.10.1 The following records shall be retained for at least five years; l
a.
Records and logs of facility operation covering time interval-at each power level.
1 6-23
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b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c.
A0 reports.
d.
Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
e.
Records of reactor tests and experiments.
f.
Records of changes made to Operating Procedures.
g.
Records of radioactive shipments.
h.
Records of monthly facility radiation and contamination surveys.
- i. Reecrds of training and qualification for current members of the plant staff.
J.
Records of sealed source leak tests and results.
k.
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 modifications made to systems a.
and equipment described in the FHSR.
b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
Records of radiation exposure for all individuals entering radiation control areas.
c.
d.
Records of gaseous and liquid radioactive material released to the environs.
Records of transient or operational cycles for those facility components designed for a e.
limited nudber of transients or cycles, a
6-24+
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i f.
Records of inservice inspections performed pursuant to these Technical Specifications.
g.
Records of Quality Assurance activities required by the QA Manual.
h.
Records of reviews performed for changes made to procedures or ' equipment or reviews of -
tests'and experiments pursuant to 10 CFR 50 59 i.
Records of meetings of the PRC and the SARB.
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 6.12.1 ALLOWANCE Pursuant to 10 CFR 20.103(c)(1) and (3), allowance may be made for the 'use of' respiratory pro-l tective equipment in conjunction with activities authorized by the operating license fbr this facility in determining whether individuals in restricted areas are exposed to concentrations i
l in eccess of the limits specified in Appendix "B," Table I, Column 1, of 10 CFR 20, subject' to the following conditions and limitations:
l a.
The limitc riove led in Section 20.103(a) and (b) shall.not be exceeded.
I b.
If the radioactive material is of such form that intake through the skin or other j
additional route is likely, individual exposures to radioactive material shall be I
l controlled so that the radioactive content of any critical organ from all routes of w
l intake ave:? aged over seven (T) consecutive days does not exceed that which would result from inhaling such radioactive material for forty (h0) hours at the pertinent concentra-tion values provided in Appendix "B," Table I, Column 1, of 10 CFR 20, For radioactive materials designated "Sub" in the " Isotope" co] amn of Appendix "B," Table I, c.
j Column 1, of 10 CFR 20, the concentration value specified shall be based upon exposure to t
i 6-25 I
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the material as an external radiction cource.
Individual exposures to these materials shall be accounted for as part of the limitation on individual dose in 5 20.101.
These
=aterials shall be subject to applicable process and other engineering controls.
6.12.2 In all operations in which adequate limitation er 'he inhalation of radioactive material by the use of process or other engineering controls is in;racticable, the licensee may permit an indi-vidual in a restricted area to use respiratory erotective equipment to limit the inhalation of airborne radioactive material, provided:
The limits specified in 6.12.1 above are not exceeded.
a.
f Respiratory protective equipment is selected and used so that the peak concentrations of b.
airborne radioactive material inhaled by an individuel wearing the equipment do not exceed the pertinent concentration values specified in Appendix "B," Table I, Column 1, of 10 CFR 20.
For the purposes of this subparagraph, the concentration of radioactive material that is in-haled when respirators are worn may be determined by dividing the ambient airturne concen-tration by the protection factor specified in Table 6.12-1 for the respirator protective equipment worn.
If the intake of radioactivity is later determined by other measurements to have been different than that initially estimated, the later quantity shall be used in evaluating the exposures.
The licensee advises each respirator user that he may leave the area at any time for relief c.
from respirator use in case of equipment malfunction, physical or psychological discomfort, or any other condition that might cause reduction in the protection afforded the wearer.
d.
The licensee maintains a respiratory protective program adequate to assure tha*. the require-ments above are met.
Such a program shall include:
(1) Air sampling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper selection of respiratory protective equipment.
(2) Written procedures to assure proper training of personnel using such protective equipment.
(3) written procedures to assure the adequate fitting of respirators; snd the testing of respiratory protective equipment for operability immediately prio) to use.
6-26
^
.m (h) Written procedures for maintenance to assure full effectiveness of respiratory protec-tive equipment, including issuance, cleaning, decontamination, inspection, repair and storage.
(5) Written operational and administrative procedures for proper use of respiratory protective equipment.
(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 US Bureau of Mines under its appropriate
^~
e.
Approval Sched.nies as set forth in Table 6.12-1.
Equipment not approved under US 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 performanie characteristics of the equipment are at least equal to those afforded by US Bureau of Mines 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.
6.12 3 REVOCATION The specifications of Se tion 6.12 shall be revoked in their entirety upon adoption of the proposed change to 10 CFR 20, Section 20.103, which would make such provisions unnecessary.
6-27
m TABLE 6.12-1 Protection Factors for Respirators 2
Protection Factors Particulates and Guides to Selection of Equipment Vapors and Gases Bureau of Mines Approval ' Schedules 4
Except Tritium For Equipment Capable of Providing at l
Oxide 3 Least Equivalent Protection Factors Description Modes I.
AIR-PURIFYING RESPIRATORS Facepiece, Half-Mask" t N?
5 21B 30 CFR S 14.h(b)(h) 7 Facepiece, Fu11 NP 100 21B 30 CFR 5 14.h(b)(5); 1hF 30 CFR 13 7
II.
ATMOSPHERE-SUPPLYING RESPIRATOR 1.
Air Line Respirator Facepiece, Half-Mask CF 100 19B 30 CFR 5 12.2(c)(2) Type C(i)
Facepiece, Full CF 1,000 19B 30 CFR 5 12.2(c)(2) Type C(i)
Facepiece, Full D
100 19B 30 CFR 5 12.2(c)(2) Type C(ii) 7 Facepiece, Full PD 1,000 19B 30 CFR 5 12.2(c)(2) Type C(iii) 5 6
6 2.
Self-Contained Breathing Apparatus (SCBAl Facepiece, Ful17 D
100 13E 30 CFR S l's.h(b)(2)(i)
Facepiece, Full PD 1,000 13E 30 CFR S 11.h(b)(2)(ii)
Facepiece, Full R
100 13E 30 CFR 5 11.h(b)(1)
III.
COMBINATION RESPIRATOR Any Combination of Air-Protection Factor for 19B CFR S 12.2(e) or Applicable Purifying and Atmosphere-Type and Mode of Opera-Schedules as Listed Above Supplying Respirator tion as Listed Above 1,
2, 3,
4, 5,
6, 7(These notes are on the following pages.)
- 0r Schedule Superseding for Equipment of Type Listed 6-28
m.
A
,~
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a TABLE 6.12-1 (Contd)
ISee the following symbols:
1 CF: Continuous Flow D:
Demand NP: Negative Pressure (ie, Negative Phase During Inhalation) ~
PD: Pressure Demand (ie, Always Positive Pressure)
R:
Recirculating (Closed Circuit) 2(a) For purposes of this specification the protection factor is a measure _of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radioactive material outside the respiratory protective equipment to that inside the equipment (usually insi~; the face-piece) under conditions of use.
It is applied to the ambient airborne concentration to estimate the concentration inhaled by the wearer according to the following fonsula:
i
^#
Concentration Inhaled = ^"
Protection Factor (b) The protection factors apply:
(i) Only for trained individuals wearing properly fitted respirators used and maintained under supervision in a well-planned respiratory protective program.
(ii) For air-purifying respirators only when high efficiency (above 99 9% removal efficiency by 4
US Bureau of Mines type dioctyl phthalate (DOP) test) particulate filters and/or sorbents appropriate to the hazard are used in atmospheres not deficient in oxygen.
1 (iii) For atmosphere-supplying respirators only when supplied with adequate respirable air.
6-29
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.Mb-TABLE 6.12-1 (Contd)
Excluding radioactive contaminants that present an absorption or submersion hazard..For tritium oxide, 3
approximately half of the intake occurs by absorption through the skin Oo that an overall protection factor.
of not more than approximately 2 is appropriate when atmosphere-supplying respirators are used to protect 1
See against tritium oxide. Air-purifying respirators are not recommended for use against tritium oxide.
also Fo3tnote 5, below, concerning supplied-air suits and hoods.
"Under chin type only.
Not recommended for use where it might be possible for the ambient airborne-concen-(
tration to reach instantaneouc values greater than 50 times the pertinent values in Appendix "B," Table I, J
Column 1 of 10 CFR, Part 20.
Appropriate protection factors must be determined taking account of tse design of the suit or hood and its-1 5
permeability to the contaminant under conditions of use. No protection factor greater than 1,000 shall be
)
used except as authorized by the Commission.
6No approval schedules currently available for this equipment. Equipment must be evaluated by testing or on basis of available test information.
70nly for shaven faces.
NOTE 1: Protection factors for respirators, as may be approved by the US Bureau of Mines according to approval schedules for. respirators to protect against airborne radionuclides, may be used to the extent that they do not exceed the protection factors listed in this table. The pro-tection factors in this table may not be a,7ropriate to circumstances where chemical or other respiratory hazards exist in addition to radst.ccive hazards. The selection and use.of respirators for such circumstances should take into account approvals of the US Bureau of Mines in accordance with its applicable schedules.
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i II.
Discussion By letter dated October 21,197h, the Regulatory Staff requested that Consumers Power Company submit a proposed change to the section on Administrative Controls in the Big Rock Point Plant Technical Specifica-tions. Justification for any departure from the standard section on Administrative Controls transmitted by the October 21, 1974 letter or Regulatory Guide 1.16, Revision 2, was requested.
In several areas of the proposed Technical Specifications Change, we have deviated from the standard Administrative Controls and Regulator y Guide 1.16, Revision 2.
Justification for these deviations follows.
The existing Big Rock Point Plant Technical Specifications were developed in 1962 and are not in the format of more recently developed Tech-nical Specifications. The standard section titled " Administrative Controls" has been developed to complement more recently developed Technical Speci-fications. The implementation of the standard format requires major re-vision to the existing Big Rock Point Plant Technical Specifications.
Such a revision is under way, but all differences between the Directorate of Licensing and Consumers Power Company have not yet been fully reconciled.
The existing Technical Specifications do not explicitly contain
" Safety Limits," " Limiting Conditions for Operation," etc. The new section on Administrative Controls requirer, explicit definition. Purther, the existing section on Administrative Controls contains many requirements other than Adrinistrative Control requirements.
Incorporation of these requirementr. in other sections of the existing Technical Specifications would require a major modification and, as stated previously, an effective modification has already been proposed.
We have modified Section 6.2.c for clarification of the existing words " reactor start-up" by inserting after them "(approach to critical)."
We have taken exception to the requirement in the Technical Standard Administrative Controls numbered 6.2.d which states, "An indi-vidual qualified in radiation protection procedures shall be on site when fuel is in *ne reactor." We believe that this statement would require a Radiation Protection Technician to be on site at all times when fuel is
(
in the reactor. This would unnecessarily require additional manning of f
e s
2
(
the Big Rock Point Plant. On backshifts, no Radiation Protection Techni-cians a v on site unless they have been specially called in for a specific task. The individuals at the plant during normal backshift conditions have qualified as radiation work permit exempt workers. In addition, the Shift Supervisor, under the Site Emergency Plan, is qualified to assume the responsibility of the Site Emergency Director. He Is qualified and j
licensed to perform this function. He is always on site when fuel is in
.the reactor. He also has the autho r./ to call any plant personnel back to the site during normal operations et the situation may require. These j
personnel include Radiation Protection Technicians. Therefore, we' do not believe a requirement to have additional staffing in the form of continu-s 1
ous around-the-clock Radiation Protection Technicians at the plant site is necessary or justified.
Item 6.2.2.e requires that all core alterations be directly supervised by a licensed Senior Reactor Operator. We have reworded this for clarification to require direction by a licensed Reactor Operator under the general supervision of a licensed Senior Reactor Operator. A licensed Reactor Operator is qua.ified and tested on his knowledge and skills which
-include the cafety considerations asse iated with refueling as well as actual equipment operatic.
We believe that devoting the entire ~atten-tion of a licensed Reactor Operator to the task of core alterations is justified and that a senior licensed operator should function in a general supervision capacity such that he is aware of the core altera-tions that are taking place as well as other evolutions in the plant that may have an effect on the core alterations.
i In addition, core alterations are performed using detailed procedures. These procedures are required to be reviewed and approved by the Plant Review Committee (PRC) before they are utilized. The'se step-by-step procedures are required to be followed precisely by the l
licensed Reactor Operator. This provides additional assurance and provides additional detailed technical knowledge in the conduct of operations associated with core alterations.
Item 6.51.2 composition of the PRC has been modified to reflect f
actual job titles at the Big Rock Point Plant. Several additional members have been specified beyond that suggested in the standard administrative 9
~-
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g 3
i s
control. ~These personnel include: The Quality Assurance Engineer, one Shift Supervisor, the Training Supervisor and an Engineer with at least one year's plant experience. For those positions which could be filled by more than one individual, any individual (but'not more than one) meeting l
the requirements could, at any' time, fill the position as a member of the t
committee. We believe these additions provide a greater depth of knowledge of the plant'as well as. greater flexibility to the PRC. This flexibility.
is needed because of the PRC workload. We have found that, using a com-mittee organization with very similar responsibilities, a large amount of time is required on the part of all members.
It is necessary to keep I
abreast of other responsibilities ta perform effectively as a PRC member.
2 j
The time requirement tends to detract from the performance of other duties 3
assigned. The addition of the three members will tend to reduce the work-
~
load slightly on key managment personnel and allow more time to be spent 4
on other assigned responsibilities.
We have deleted requirenents for the PRC to review and approve I'
the Plant Security Plan and Implenenting Procedures. During the prepera-tion of the Plant Security Plan and Implementing Prceedures, we addressed the problem of review of these types of procedures. At that time, we concluded that it was more meaningful to provide for review of changes in the Plant Security Plan or procedures by a combination of a represen-f tative from the Company Security Department as well as a knowledgeable member of the plant staff. We believe that requiring review by the PRC
~
would be time-consuming and offer no advantage over the present methods used. We note that neither the PRC nor Safety and Audit Review Board (SARB) has any personnel on the comr'.ttee with expertise in Security Operations.
i Section 4.5 2 " Safety and Audit Review Board" is proposed in accordance with directions provided in the box text in the standard section titled " Administrative Controls." Several differences exist between our proposd and the example provided other than those differences necessary to facilitate our organization. The function of the SARB is 4
more explicitly defined to clarify requirements. We have also defined areas of expertise and qualifications for SARB members and alternates rather i
I-
'a
~
.g b,
h than specifying position titles. This provides flexibility for. minor or-ganizational changes without requiring Technical Specifications changes.
The provisions of section concerning consultants have been expanded to allow SARB members to use consultants as they see fit. A quorum is defined as four members including alternates plus the chairman rather than a majority. Our proposal also includes the requirement that the g
quorum include appropriately qualified members or use of consultants-We believe this offers some flexibility to SARB while being more effec-tive than requiring a simple majority.
Further, with regard to SARB review, we have eliminated Items 6.5 2 7.a and 6.5 2.7.h of the standard Technical Specifications. Item "a" is fulfilled by review of the reports and meeting minutes of ti e PRC l
and item 6by review of " Abnormal Occurrences."
In addition, several new items have been added under "Reviev."
These are due to organization differences between Consumers Power Company i
and the assumed " typical" organization. Our Quality Assurance Department k
and Technical Services Department are independent and do not have line responsibility for the operation of the plant. As such, they are assigned 4
auditing functions as described under " Audits." The audits are reviewed by SARB.
In addition, under " Audits" we have deleted requirement of insurir.g that the QA program meets the requirements of Appendix B.
This respor.sibility is otherwise assigned as indicated in the description of QA submitted under Docket No 50-255 This program is being refined and 1
vill be implemented at both plants. Further, the provisions for audit of the Security Plan are contained in the Security Plsn. For the rea-sons set forth earlier in this discussion, we do not believe it is ap-propriate for SARB to evaluate the effectiveness of ;he Security Plan.
Under " Records," we have removed the requitement that the minutes and reports of reviews be approved within 14 days. We believe a
it is unnecessary to get the SARB m unbers together for an additional meeting to approve the minutes of a previous meeting within the 1b-day time requirement. The meeting minutes vill be routed individually to i
- (:
j SARB members for approval and, if necessary, be brought up,at the next regular meeting. Any differences will be resolved which will require reporting to the Vice President - Bulk Power Operations (BPO).
.V.
o
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5
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Secion 6.6.1.b - The words proposed broaden the scope of the PRC review and clarify what we believe tc be the intent of this section of the standard Technical Specifications. We do not believe that the review of the Abnormal Occurrence (AO) performed by the PRC should be limited to the report submitted alone. We further believe that the re-sults of the PRC review should be submitted to the SARB and the Electric Production Superintendent - Nuclear. In addition, the change in the wording clarifies whether the review requirement is at a prereport requirement or a postreport requirement. We believe this change is cont..atent with the reporting philosophy of Regulatory Guide 1.16, Revision 2.
Section 6.T l.a through d - This section concerning safety limit violatior.s has been rewritten to clarify the intent and make it compatible with Regulatory Guide 1.16, Revision 2.
We believe the wording fulfills the intent of the standard Technical Specifications and is much clearer than that which appears in the standard Technical Specifications.
(
In Section 6.8.2, we have deleted the phrase, "and periodically as set forth in each document." We believe that this phrase is so general in nature that it has very little value.
Section 6.8.3.b - The term " Plant Management Staff" is not clearly defined in the standard Technical Specifications nor is it meaningful in terms of the plant organization. We have substituted PRC in its place; we feel this is appropriate as the original procedures were required to be reviewed by the PRC.
Section 6.8.3.c - Has been modified to allow PRC review at the next regularly schedaled meeting. The requirement to have a special mceting to provide PRC review of a temporary change is time-consuming.
The temporary changes can be efficiently reviewed at regularly scheduled meetings. We believe that this requirement vill result in a maximum of approximately 30 days between the implementation of a temporary change and the review of the said change. We believe this is reasonable in that at least two members of the PRC are required to approve the tempo-rary change.
(
Section 6.91 in the standard Technical Specific 9 ons included ti a phrase 2ncorporating Regulatory Guide 1.16, Revision 2, by reference.
r
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c s
6
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We. feel it its more desirable to specifically repeat the reporting require-ments in the Technical Specifications. This saves the inconvenience of having-to use two documents to determine whether an item is reportable and how it should be treated. If the Regulatory Guide were incorporated by reference, the Technical Specifications would have to be checked also to determine if exception were taken to these requirements. Regulatory i
Guide 1.16, Revision 2, as modified by telephone conversation with a
. member of the Directorate of Licensing staff on November 19,197h hac been incorporated except as described in the following paragraphs.
The requirements of the annual operating report have been modified to include effluent release data, solid radioactive vaste data and radiological environmental monitoring data. The Big dock Point Plant does not yet have an Appendix "B" to the Technical Specifications.
Several exceptions were taken to Regulatory Guide 1.21 These exceptions include:
1.
We have specified that t a requirements of Reguir. tory Guide i i 1.21 be reported annually to be consistent with the annual -
report of the standard Technical Specifications. Regulatory Guide 1.21 states they should be reported semiannually.
With regard to the solid waste report, we have deleted, "an 2.
estimate of the major nuclide composition." In the past, we have usually repo*ted the activity as mixed fission product and see no value in the change. This would be both time-consuming and expensive.
3 The requirements for reporting with regard to radiological impact on man have been modified so that only doses calculated to be greater than 1 mrem need be reported.
4.
We have changed the wording with regard to radiological impact on man for, " total body and skin doses to individuals exposed at the point of maximum off-site ground level concentration" to "aaximum off-site air dose due to beta and gamma radiation, etc," because the gamma (total body) dose would not be a maximum at the point of maximum groand concentration due to
(
the tall stsch at the Big Rock Point Plant.
/
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)
Q' T
.(
5 Again, with regard to the radiological impact on man, we have
. deleted the requirement " direct radiation from the facility" because any such dose would be measured by environmental TLDs (plus plume dose). The direct radiation doses are less than 1 mrem per year. The calculat, lens associated with these doses are complex and time-consuming. We believe it is tetter practice to measure the doae rather than rely on time-cortuming complex calculations.
6.
The meteorological reporting requirements have been deleted.
The meteorological tower at the Big Rock Point Plant was re-moved several years ago prior to the issuance of Regulatory Guide 1.21.
As the Big Rock Point Plant releases small amounts r
of radioactivity via the plant stack essentially continuously, we believe the use of averaged meteorology is justified and do not feel it necessary to reinstall a meteorological tower fcr the sole purpose of conformance with suggested reporting re-quirements in a Regulatory Guide.
T.
Under environmental data ve have changed the wording "if a particular sample or measurement indicates statistically significant levels and, etc," to "if samples or measurements averaged over appropriate data base indicate, etc."
This will eliminate the need for detailed reporting on random high readings from TLDs and air samplers.
8.
The weekly tabulation of off-gas data required has been modified for purposes of clarification.
It now clearly indicates that the sample is taken at the steam jet air ejector and corrected for holdup time prior to release.
9 The requirement for a tabulation of primary coolant sample results following short-term increases in off-gas release rates has been modified such that it is meaningful at Big Rock Toint. The installed instrumentation at the plant is sufficient to detect trends but not sufficiently precise l
to perform as required if the Regulatory Guide words were I
specified. We see no need to upgrade this equipy nt because r
l.
)
s 8
i releases are based on sample results. The words have been modified such that they are achievable at Big Rock Point and -
still attempt to fulfill as closely as possiL~ the data re-quirement of the Regulatory Guide.
Under " Abnormal Occurrence Reporte" (6.9 1.2.a(4)) we have-modified the reactivity values considered abntrmal by the regulatory guide. Based on our operating and core physics experienec, we believe, that for a BWR, differences of $1.00 and $.50 are insignificant and not unusual. Therefore, we have proposed 1% dk/k and
.5" dk/k, respectively, as anamolies of this magnitude begin to be significant. The only other alternative we see is to better define " reactivity balance."
Under " Abnormal Occurrence Reports" (6.9 1.2.a(h)) we have altered the words suggested in the Regulatory Guide to clarify the in-tent. We believe that the Regulatory Guide definition could be con-strued to requile reporting of all maintenance performed on essentially all safety-related systems. This is not consistent with the intent of f
the prompt notification Abnormal Occurrence Reports.
The time requirement for the submittal of the 30-day reports has been modified to require submittal within 30 days of discovery of the event classified as an AO.
The section on record retention has been modified slightly. We have clearly specified that records not previously required to be retained are not required to be retained until January 1, 1976. This is specified to clarify why some of these records were not retained in the past and to provide time to Onplement any change necessary in the records retention system.
Two record retention requirements have been changed from the life of the facility to fite years. This is the requirement for training r*
'ds. In our operator training program, which was approved by the
, ate of Licensing last year, we specify retention for two train-in This is nearly five years. We see no need to escalate this u
requiremeo The second is with regard to retention of facility contam-ination su.vey records. We do not believe it is necessary to maintain these records for longer than five years. Further, this exejeeds the two-year retention requirement of 10 CFR 20.
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9 III. Conclusion Based on the foregoing, both the Big Rock Point PRC and SARB have concluded that these proposed changes do not involve an ur.aviewed safety question.
CONSl4ERS POWER COMPANY By
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R. A. Lamley, Vice Prerddent Sworn and subscribed to before me this 13th day of January 1975 Sc d
_ 41(,
Sylvic /B. Ball, Notary Public Jackson County, Michigan PJ comission expires May 18, 1976.
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