ML19341C851
| ML19341C851 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 02/11/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19341C847 | List: |
| References | |
| NUDOCS 8103040291 | |
| Download: ML19341C851 (20) | |
Text
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'o UNITED STATES E\\f, e ( ~,,g NUCLEAR REGULATORY COMMISSION j gj WASHINGTON, D. C. 20555 g/(
pV_Ql'NNE LIGHT COMPANY OHIO EDISON COMPANY PENNSYLVANIA POWER COMPANY DOCKET N0. 50-334 BEAVER VALLEY POWER STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 39 Licanse No. DPR-66
' l~.
The' Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Duquesne Light Company, Ohio Edison Company, and Pennsylvania Power Company (the licensees) dated September 17, 1980, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this afnendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, Facility Operating License No. DPR-66 is hereby amended by:
a.
Renumbering paragraphs 2.C.(4), 2.C.(5), 2.C.(8) and 2.C.(9) as 2.C.(3), 2.C.(4), 2.C.(5) and 2.C.(6), respectively.
b.
Adding the following paragraphs 2.C.(7), 2.C.(8) and 2.C.(9):
e 8103040)f{,
2-l (7) Systems Integrity Duquesne Light Company shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident ta as low as practical levels.
This program shall include the following:
1.
Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2.
Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
.(8)
Iodine Monitoring Duquesne Light Company shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.
This program shall include the following:
1.
Training of personnel, 2.
Procedures for monitoring, and.
3.
Provisions for maintenance of sampling and analysis equipment.
(9)
Backup Method for Determining Subcooling Margin Duquense Light Company shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin.
This program shall include the following:
1.
Training of personnel, and 2.
Procedures for monitoring.
3.
Additionally, the license is amended by changes to the Technigal Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Op. rating' License No. DPR-66 is hereby amended to read e
as follows:
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 39, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with t.he Technical Specifications.
e 0
6
3-4.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
)
g-ef Operating Reac Branch #1 Division of Lic ing
Attachment:
Changes to the Techni. cal Spe'cifications Date,of Issuance:
February 11, 1981 i
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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 39 TO FACILITY OPERATING LICENSE NO. OPR-66 DOCKET NO. 50-334 Revise Appendix A as follows:
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Remove Pages Insert Pages iv iv v
v 3/4 3-19a 3/4 3-24a 3/4 3-27a 3/4 3-31 a 3/4 3-50 3/4 3-51 3/4 3-52 1
3/44-7 3/4 4-7 3/4 4-31 B 3/4 3-3 B 3/4 3-3 B 3/4 4-2 B 3/4 4-2 B 3/4 4-10 B 3/4 4-10 6-4 6-4 6-5 6-5 l
i i
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'(l INDEX i
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LIMITING CONDITIONS FOR OPERATIC.N AND SURVEILLANCE RE00!?EME"T5 Pace 1.:
.c r. 10N
.i 3/4.2 POWER OISTRIBUTION LIMITS 1
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!3/4.2.1 Axi al Fl ux 0i f f e rence.....................................
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3/4 2-8 i
3/4.2.4 Qu a d ra n t F.cwe r T i 10. Ra t i c.................................
3/ 4 2-10
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iv Amendment No. 39 s.
I
v INDEX LIMITING CONDITIONS FOR OPER' ION AND SURVEILLANCE REQUIREMENTS Page SECTION 3/4.4.2 SAFETY VALVES - SHUTD0WN...............................
3/4 4-5 3/4.4.3 S AF ET Y V ALV ES - 0 P E RAT I NG..............................
3/4 4-6 3/4.4.4 PRESSURIZER............................................
3/4 4-7 3/4.4.5 STEAM GENERATORS.......................................
3/4 4-8 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems..............................
3/4 4-11 Operational Leakage....................................
3/4 4-13 3/4.4.7 CHEMISTRY..............................................
3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY......................................
3/4 4-18 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.................................
3/4 4-22 Pressurizer............................................
3/4 4-27 3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Com;onents..................
3/4 4-28 3/4.5 EMERGEt:CY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS...........................................
3/4 5-1 3/4 5-3 avg-350*F.........................
3/4.5.2 ECCS SUBSYSTEMS - T 3/4 5-6 3/4.5.3 ECCS SUBSYSTEMS - T,yg <
350"F.........................
3/4.5.4 BORON INJECTION SYSTEM Boron Injection Tank...................................
3/4 5-7 Heat Tracing...........................................
3/4 5-8 3/4.5.5 R E F U E L I N G W AT E R ST O RA GE T AN K... '......................
3/4 5-9 BEAVER VALLEY - UNIT 1 V-Amendment No. 39
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TAllLE 3.3-3 (Continued)
<n 9M EllGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION N
MINIMUM E
TOTAL NO.
ClfANNELS CllANNELS
-APPLICABLE O
FUNCTIONAL UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION g
7.
AUXILIARY FEE 0 WATER
[
a.
Steam. Gen. Water Level-Low-Low t
f.
Start Turbine i
A Driven Puinp 3/stm. gen.
2/sta. gen.
2/sta. gen 1, 2, 3, j
14 any stm. gen.
11.
Start Motor 7
y Driven Pumps 3/stm. gen 2/sta. gen.
2/sta. gen.
1, 2, 3 l
14 any 2 sim. gen.
l 9$
h.
UnderynitaDe-RCP'
' Start Turbine-Driven Pump (3)-1/hus 2
2 1
/
14 i
c.
S. I.
Start Motor-Driven Pumps See 1 above (all S.I. initiating functions and requirements) d.
Faergency P.us Undervoltage Start Hotor Driven R
Pumps 1/hus 1
1 1,2,3 18 5
3 e.
Trip of Main g
feedwater Pumps Start Motor-g Driven Pumps 1/ pump 1
1 1,2,3 18
(j lAlllE 3.3-4 (Continuedl SI ENGINEEPED SAFETY FEATURE ACIllATION SYSTEM INSTRilHENTATION TRIP SETPOINTS r-E FUNCTIONAL UNIT
' TRIP SETPOINT' ALLOWABLE VALUES h
7.
El a.
Steam Generator Water Level-low-low
> 12%
of narrow range
> lli of narrow raage Instroment span each Instrument span each steam generator steam generator b.
Undervoltage - RCP 1 2750 volts RCP bus voltage 1 2725 volts RCP bus volta 0e c.
5.1.
See 1 above (all SI Setpoints) i-k' l
4x d.
Emergency' Bus Undervoltage
< 3350 volts
< 3325 volts i
jf e.
Trip of Hain Feedwater N/A N/A Pumps a
Bn N
TABLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 11.
Steam Generator Water Level-Low-low a.
Motor-driven Auxiliary 60.0 Feedwater Pumps **
b.
Turbine-driven Auxiliary 60.0 Feedwater Pumps ***
12.
Undervoltage RCP a.
Turbine-driven. Auxiliary 60.0 Feedwater Pumps 13.
I:ergency 3us Unde: voltage a.
Motor-driven Auxiliary 60.0 Feedwater Pumps 14.
Trio of Main Feedwater Pumas a.
Motor-driven Auxiliary 60.0 Feedwater Pumps Note:
Response time for Motor-60.0 driven Auxiliary Feedwater Pumps on all S.I. signal starts
- "on 2/3 any Steam Generator
""on 2/3 in 2/3 Steam Generators O
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Amendment No. 39
[
BEAVE.R VALLEY - UNIT 1 3/4.3-27a
TAllt.E 4.3-2 (Continued) 9-ENGINEERED SAFETY FEAltlRE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REf)DIREMENTS CilANNEL MODES IN WilICil
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CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CllECK CALIBRATION TEST REQUIRED N
7.
Steam Generator Water S
R H
1, 2, 3 Level-Low-tow b.
Undervoltage"- RCP S
R M
1, 2
~
uy See 1 above (all SI surveillance requirements) c.
S.I.
I'u d.
Dnergency nus Undervoltage N/A R
R 1, 2, 3 Trip of Main Feedwater N/A N/A R
1,2,3 e.
, Pumps 5
a t
ae F
INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 4
I 3.3.3.8 The accident monitoring instrumentation channels shown in
' Table 3.3.11 shall be OPERABLE.
4 APPLICABILITY: MODES 1, 2 and 3.
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l ACTION:
j a.
With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.3.11, either restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except for the PORV(s) which may be isolated in accordance with Specification 3.4.11.a.
b.
With the number of OPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Table 3.3.11, either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
c.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE RE0UIREMENTS
- 4. 3.3. 8 Each accident monitoring instrumentation channel shall be demon-strated OPERABLE by performance of the CHANNEL CHECX and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.
4 t
9 1
BEAVER VALLEY ~ UNIT 1
, 3/4 3-50 Amendment No. 39
Tne Ri TABLE 3.3-11 n
ACCIDENT MONITORING INSTRUMENTATION f
I 2
TOTAL NO.
MINIMUM CllANNELS s.
gg INSTRUMENT _
0F CilANNELS OPERABLE 1.
Pressurizer Water Level (3)
(2) 2.
Auxiliary Feedwater Flow Rate (1) per steam gen.
(1) per steam gen.
3.
Reactor Coolant System Subcooling Margin Monitor (1)
(0) se 4.
PORV Accoustical Detector Position Indicator 2/ valve
- 1/ valve
+
y>
5.
PORV. Limit Switch Position Indicator' 1/ valve 0/ valve 6.
PORV Block Valve Limit Switch Position Indicator 1/ valve 0/va.lve i
'7.
Safety Valve Accoustical Detector Position Indicator 2/ valve
- 1/ valve i
Tl.
Safety Valve Temperature Detector Position Indicator 1/ valve 0/ valve
- One Detector Active, Second Detector Passive F
a i
an O
e-
. Ei2 93 TABLE 4.3-7 ff ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
=
CilANNEL CllANNEL
[5 INSTRUMENT CllECK CALIBRATION
--4 1.
Pressurizer Water Level M
R III 2.
Auxiliary Feedwater Flow Rate S/U R
3.
Reactor Coolant System Subcooling Margin Monitor M
R 4.
PORV Accoustical Detector Pcsition Indicator M
R Y'
10 5.
PORV Limit Switch Position Indicator M
R 6.
PORV Block Valve limit Switch Position Indicator M
R 7.
Safety Valve Accoustical Detector Position Indicator M
R 8.
Safety Valve Temperature Detector Position Indicator M
R b'(
(1) Channel check to be performed in conjunction with Surveillance Requirement 4.7.1.2.a.9 g
following an extended plant outage.
S i
REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with at least (150) kw of pressurizer heaters and with a steam bubble.
APPLICABILITY:
MODES 1, 2 and 3.
ACTION:
With the pressurizer inoperable due to less than 150 kw of heaters supplied by an emergency bus, be in at leas HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With the pressurizer otherwise inoperable, be in at leas: HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in the HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.4.1 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE it least once per 18 months by energizing the heaters supplied by the emergency bus.
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L 3/4 4-7 Amendment No. 39 BEAVCR VALLEY - UNIT 1
REACTOR COOLANT SYSTiM, 3/4 4.11 RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.11
( Two ) power operated relief valves (PORVs) and their associated block valves shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
a.
Withless than 2 PORV(s) operable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore two PORV(s) to OPERABLE status or close the associated block valve (s) and remove power.from the-block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve (s) to OPERABLE status or close the block valves (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT-DOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
4 SURVEILLANCE REQUIREMENTS 4.4.11.1 Each PORY shall be demonstrated OPERABLE:
a.
At least once per 31 days by performance of a CHANNEL CHECK of tne position indication, excluding valve operation and b.
At least once per 18 months by performance of a CHANNEL CALIBRATION.
4.4.11.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.
4.4.11.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through a complete cycle of full travel.
1 BEAVER VALLEY - UNIT 1 3/4 4-31 Amendment No. 39
I 1
TNSTRUMENTATION BASES A
3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room.
This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.
3/4.3.3.6 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that i
adequate waraing capability is available for the prompt detection of fires.
This capability is required in order to detect and locate fires in their early stages.
Prompt detection of fires will reduce the poten-tial for damage to safety related equipment and is an integral element in the overall facility fire protection program.
In the event that a portion of the fire detection instrumentation 4
is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.
3/4.3.3.8 ACCIDENT MONITORING INSTRUMENTATION I
The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.
This capability is consistent with the recommendations of Regulatory Guide 1.97,
" Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Con-ditions During and Following an Accident," Decemoer 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommenda-tions."
t i
BEAVER VALLEY - UNIT 1 B 3/4 3-3 Amendment No. 39 p
REACTOR COOLANT SYSTEM BASES relieve any overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop, con-nected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer. code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.
The c'ombined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor tr'p until the first Reactor Protective System trip set point is reached (i.e...no credit is.taken for a direct reator trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Code, dated July 1974.
3/4.4.4 PRESSURIZER The requirement that (150)kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.
3/4.4.5 STEAM GENERATORS One OPERABLE steam generator in a non-isolated reactor coolant loop provides sufficient heat removal capability to remove decay heat after a reactor shutdown.
The requirement for two OPERABLE steam generators, combined with other requirements of the Limiting Conditions for Operation ensures adequate decay heat removal capabilities for RLS temperatures greater than 350 F if one steam generator becomes inoperable due to single failure considerations.
Below 350 F, decay heat is removed by the RHR system.
BEAVER VALLEY - UNIT 1 B 3/4 4-2 Amendment No. 39
REACTOR COOLANT SYSTEM BASES vessel inside radius are essentially identical, the measured transition shift for a _>mple can be applied with confidence to the adjacent section of the reactor vessel.
The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule is differentfromthecalcul52[ dartNDT for the equivalent capsule radiation exposure.
The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been pro-vided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice leak and hydrostatic testing.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in
~
Table 4.4-3 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
The limitations i~
.ed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the g
pressurizer is operated within the design criteria assumed for the fatigue-analysis performed in accordance with the ASME Code requirements.
3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintainec at an acceptable level tnroughout the life of the plant.
These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessei Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g)(6)(i).
3/4.4.11 RELIEF VALVES The relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.
The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.
1 BEAVER VALLEY - UNIT 1 B 3/4 4-10 Amendment No. 39
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TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION #
SINGLE UNIT-FACILITY LICENSE CATEGORY APPLICABLE MODES QUALIFICATIONS 1, 2, 3 and 4 5 and 6 SRO*
2 1**
R0 2
1 Non-Licensed Auxiliary 2
1 Operator Shift Technical Advisor i
None Required
- Includes the licensed Senior Reactor Operator serving as the Shift Supervisor.
- Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE OPERATIONS.
- Shift crew composition may be one less than the minimum requirements for i
a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to witnin the minimum requirements of Table 6.2-1.
This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.
1 BEAVER VALLEY - UNIT 1 6-4 Amendment No. 39
ADMINISTRATXVE CONTROLS a 6. 3 FACILITY STAFF OUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiation Control Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and response analysis of the plant for transients and accidents.
6.' 4. TRAINING 6.4.1 A retraining and replacement training program for the facility 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 ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55.
6.4.2 A Training program for the Emergency Squad shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements of Section 27 of the NEPA Code-1976.
6.5 REVIEW AND AUDIT 6.5.1 ONSITE SAFETY COMMITTEE (OSC)
FUNCTION 6.5.1.1.The OSC shall function to advise the Plant Superintendent on all matters related to nuclear safety.
COMPOSITION 6.5.1.2 The OSC shall be composed of the:
4 Chairman:
Chief Engineer Member:
Operations Supervisor Member:
Radiation Control Supervisor Member:
Maintenance Supervisor Member:
Nuclear Engineering & Refueling Supervisor Member:
Results Coordinator Member:
Training Supervisor Member:
Office Manager Nuclear (Security Officer)
Member:
Senior Engineer - Emergency Planning and Fire Protection Member:
Technical Advisory Engineer ALTERNATES 6.5.1.3 All alternate members shall be appointeit in writing by the OSC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in OSC activities at any one time.
, BE, AVER VALLEY - UNIT 1 6-5 Amendment No. 39