ML20070V523
| ML20070V523 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 02/07/1983 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20070V511 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 GL-82-16, NUDOCS 8302170385 | |
| Download: ML20070V523 (5) | |
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6.0 ADMINISTRATIVE CONTROLS i
6.1 RESPONSIBILITY 6.1.1 The Plant Superintendent shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.
6.2 ORGANIZATION l
OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1.
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FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:
l Each on duty shif t shall be composed of at least the a.
minimum shif t crew composition shown in Table 6.2-1.
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b.
At least one licensed Operator shall be in the control room when fuel is in the reactor.
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-l c.
At least two licensed Operators shall be in the control
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room during reactor start-up, scheduled reactor shutdown.
and during recovery from reactor trips.
d.
An individual qualified in radiation protection procedures shall be onsite when fuel is in the reactor.
e.
ALL CORE ALTERATIONS af ter the initial fuel loading shall be l
directly supervised by either a licensed Senior Reactor 3
Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this ope ra tion.
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A Fire Brigade of at least 5 members shall be maintained on I
site at all times. The Fire Brigade shall not include 3 members of the minimum shif t crew necessa ry for safe shutdown l
of the unit or any personnel required for other essential lj functions during a fire emergency.
g.
Administrative procedures shall be developed and implemented to limit the working hours of unit staf f who perform safety-a rela ted functions; senior reactor operstors, reactor operators, l l radiation control technicians, auxiliary operators, meter and I
control repairman, and all personnel actually performing work i
on safety related equipment.
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6-1 8302170385 830207 PDR ADOCK 05000325 PDR p
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ADMINISTRATIVE CONTROLS (Continurd) 1 The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant is opera-ting. However, in the event that unforseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the follow-ing guidelines shall be followed:
a.
An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.
b.
An individual should not be permitted to work more than t
16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> I
in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any seven day period, all excluding shif t turnover time.
c.
A break of at least eight hours should be allowed between work periods, including shift turnover time.
d.
Except during extended shutdown periods, the use of over-time should be considered on an individual basis and not
'l for the entire staff on a shif t.
Any deviation from the above guidelines shall be authorized by the Plant Superintendent or predesignated alternate, or higher levels of management. kathorized deviations to the working hour guidelines shall be documented and available 4
f or NRC review.
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ADMINISTRATIVE CONTROLS according to work and job functions, 2 e.g., reactor operations and m2rveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), weste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measu rements. Small exposures totalling less than 20 percent of the individual total dose need not be accounted for.
In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions.
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,i b.
Documentation of all challenges to the pressurizer power operated relief valves (?ORVs) or pressurizer safety valves.
4 MONTHLY OPERATING REPORT 1'
6.9.1.6 Routine reports of operating statistics and shutdown experience in shall be submitted on a monthly basis to the Director, Office of Manage-
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ment Information and Program Control, U. S. Nuclear Regulatory Commission,
't Washington, DC 20555, with a copy to the Regional Of fice, submitted no later than the 15th of each month following the calendar month covered by the report.
REPORTABLE OCCURRENCES 6.9.1.7 The REPORTABLE OCCURRENCES of Specifications 6.9.1.8 and 6.9.1.9 below, including corrective actions and measures to prevent -
l recurrence, shall be reported to the NRC.
Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original report date.
PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.8 The types of events listed below shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> be telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designate no later than the first working day following the event, with a i
written followup report within 14 days. The written followup report i
shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form j
shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
a.
Failure of the reactor protection system or other systems subject to limiting safety-system settings to initiate - the required protective fenetion by the time a monitored parameter 2
~ This tabulation supplements the requirements of 20.407 of 10 CFR Pa r t 20.
BEAVER VALLEY - UNIT 1 6-14 PROPOSED WORDING 3
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i ADMINISTRATIVE CONTROLS' a
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Performance of structures, systems, or components that requires 1
remedial action or corrective measures to prevent operation in a 1
manner less conservative than that assumed in the accident analyses in the safety analysis repor or technical specifications bases;
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or discovery during plant life of conditions not specifically con-sidered 'in the safety analysis report or technical specifications 3a that require remedial action or corrective measures to prevent the existence or development of an unsafe condition.
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Failure of the pressurizer PORV's or pressurizer safety valves.
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THIRTY DAY WRITTEN REPORT I
6.9.1.9 The types of events listed below shall be the subject of written reports to the Director of the Regional Of fice within 30 days of occurrence of the event. The written. report shall include, as a minimum, a completed copy of a licensee event report form.
Information provided on. the licensee j'
event report form shall be supplemented, as needed, by additional narrative
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material to provide complete explanation of the circumstances surrounding q
the event.
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a.
Reactor protection system of engineered safety feature l
instrument settings which are found to be less conservative
'l than those established by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems.
i b.
Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.
c.
Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems i
or engineered safety feature systems.
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Abnormal degradation of systems other than those specified lI in 6.9.1.8.c above, designed to contain radioactive material
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resulting from the fissSn process.
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l RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.10 The F limit for Rated Thermal Power (F
) shall be provided to the I
xy j;
Director of the Regional Of fice of Inspection and Enforcement, with a copy to the Director, Nuclear Reactor Regulation, Attention Chief of the Core Performance
,j Branch, U. S. Nuclear Regulatory Commission, Washington, DC 20555 for all core
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planes containing bank "D" control rods and all unrodded core planes at least '
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60 days prior to cycle initial criticality. In the event that the limit would i
be submittted at some other time during core life, it will be submitted 60 days
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f fl B8 AVER VALLEY - UNIT 1 6-16 PROPOSED WORDING i
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'i ATTACHMENT B i
e Proposed Change Request No. 84 revises the Beaver Valley Power
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Station, Unit No.1 Technical Specifications, Appendix A to incor-porate the applicable NUREG-0737 changes recommended by Generic Letter 82-16.
The change to Section 6.2.2 " Facility Staff" is an administrative
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j requirement to limit the working hours of plant personnel who perform j
safety related functions. This change is recommended by NUREG-0737, Item I.A.l.3.
i The change to Section 6.9.1.5 " Annual Reports" incorporates an administrative requirement to include all challenges to the pressurizer j)
Power Operated Relief Valves (PORV's) or Pressurizer Safety valves in
'l the annual report. The change to Section 6.9.1.8 " Prompt Notification l
With Written Followup" incorporates an administrative requirement to i
report any failure of the pressurizer PORV's or pressurizer Safety valves i
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and provide a written followup within 14 days. These changes are recommended by NUREG-0737, Item II.K.3.3.
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1 The Technical Specification BASES were reviewed, and it was deter-mined that these changes will not reduce the margin of safety for any l
specification.
1 The Updated Final Safety Analysis Report (UFSAR), Sections 4.2.2.7, 7.2,14.1.7 and 14.1.15, was reviewed and it was determined that these changes will not increase the likelihood of a malfunction of safety related equipment, increase the consequences of a malfunction previously analyzed, nor create the possibility of a malfunction different than previously evaluated and do not constitute an unreviewed safety question.
These proposed changes to Sections 5.2.2, 6.9.1.5 and 6.9.1.8 are administrative and do not involve physical changes to any plant safety related systems or components.
The OSC and ORC have reviewed this proposed change, and based on the above safety evaluation, it is concluded there is reasonable assurance that the public health and safety will not be endangered by operation in the proposed manner.
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