ML20003J376
| ML20003J376 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/07/1981 |
| From: | SAN DIEGO GAS & ELECTRIC CO., SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML13330A327 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 TAC-44652, NUDOCS 8105110326 | |
| Download: ML20003J376 (44) | |
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DESCRIPTION OF PROPOSED CHANGE AND SAFETY ANALYSIS PROPOSED CHANGE NO. 102 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE DPR-13 This is a request to revise Appendix A Technical Specifications and to amend the license by adding three license conditions.
Reason for Proposed Change Subsequent to the accident at Three Mile Island, the NRC issued NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Reconnendations," July 1979, which required certain actions by light water reactor licensees, including modifications and additions of equipment. By letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water j
reactor licensees, in order to provide reasonable assurance that operation is maintained within the limits determined acceptable following the implementation of the TMI-2 Lessons Learned Category "A" items, the NRC transmitted model Technical Specifications. Proposed Change No.102 is in compliance with the guidance and scope of the July 2,1980, NRC letter.
Existing Specifications and License Conditions The existing specifications are as constituted in Appendix A Technical Specifications.
The existing license conditions are as constituted in section 3 of Provisional Operating License No. DPR-13.
Proposed Specifications and License Conditions The existing specifications and license conditions would be revised as indicated in the Enclosure to this Proposed Change. The added o. revised portions are identified by a bar in the margin.
Safety Analysis The Technical Specification changes discussed in the enclosure provide specifications for certain TMI-2 Lessons Learned Category "A" items in the areas of equipment and administrative requirements, including actions considered appropriate if a limiting condition for operation cannot be met.
Proposed Change No.102 involves:
1.
, Emergency Power Supply Requirements The pressurizer water level indicators, pressurizer relief and block valves, and pressurizer heaters are important in a post-accident situation. Adequate emergency power supplies ensure post-accident functioning of these components. The enclosed specifications will satisfy these requirements.
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2.
Valve Position Indication The installed system for indication of valve position is a diagnostic aid to the operator. Although the indicating system provides no automatic action, this system should be operable and periodic surveillance should be performed.
3.
Instrumentation for Inadequate Core Cooling 4
Containment Isolation These specifications include a Table of Containment Isolation Valves which reflect the diverse isolation signals which the design currently provides.
Limiting conditions for operation and associated surveillance are included.
5.
Auxiliary Feedwater Systems This proposed change treats both initiation and indication of auxiliary feedwater flow.
6.
Shift Technical Advisor The specification related to minimum shift manning is revised to reflect the augmentation of a Shift Technical Advisor.
7.
Containment Sphere Hydrogen Detection and Control This proposed change includes specifications which govern hydrogen monitors and hydrogen recombiners.
In a post-LOCA situation, these systems monitor hydrogen levels and maintain hydrogen concentration within containment below its flammable limit.
The limiting conditions for operation and the surveillance standards of f coposed Change No.102 will provide assurance of the reliability and availability of equipment which could be required to mitigate the consequences of an accident.
The Bases of the Proposed Change provide greater detail to Support and supplement this Safety faalysis.
The proposed license conditioris are related to a system integrity measurements program, an improved iodine measurements capability, and a backup method for determining the subcooling margin of the reactor coolant system. These programs represent a continuing commitment to safe operation of the facility.
Accordingly, it is concluded that (1) the proposed change does not involve an unreviewed safety question as defined in 10CFR50.59, nor does it present significant hazard considerations not described or implicit in the Final Safety Analysis, and (2) there is reasonable assurance that the health and Safety of the public will not be endangered by the proposed change.
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ENCLOSURE 1 0F PROPOSED CHANGE N0. 102 Portions of Proposed Change No. 83, submitted as Amendment No. 84 on September 12, 1979, are included herein to provide continuity in the numbering system of the Technical Specifications, and to submit an addition to Table 1.1, FREQUENCY NOTATION.
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TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY (1)
S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
BW At least once per 14 days.
l M
At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days.
R At least once per 18 months.
S/U Prior to each reactor startup.
P Completed prior to each release.
N.A.
Not applicable.
1 (1) For each frequency, the allowable extension is 25%. The total allowable extension for three consecutive frequency intervals is 3.25 times the interval.
This page was submitted as part of Proposed Change No. 83.
The notation for "BW" is added here.
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To App:ndix A, Sp:cification 3.5, Instrumentation and Control, th2 following
' Specifications 3.5.3 and 3.5.4 will be added:
3.5.3 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION 3.5.4 RADI0 ACTIVE GASE0US PROCESS AND EFFLUENT MONITORING INSTRUMENTATION In Appendix A, Specification 4.1, Operational Safety Items, will be revised as follows:
4.1 EQUIPMENT, INSTRUMENTATION AND SAMPLING 4.1.1 Operational Safety Items Tables 4.1.1 and 4.1.0 will be changed to tables 4.1.1.1 and 4.1.1.2 Speci fi-cations 4.1.2, Radioactive Liquid Effluent Instrumentation, and 4.1.3, Radioactive Gaseous Process and Effluent Monitoring Instrumentation, will be added as follows:
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l FOR REFERENCE ONLY l
This page contains excerpts from Proposed Change No. 83.
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l Itcms.6, 7 and 9 of Table 3.5.1 will be deleted.
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TA81.E 3.5.1 155tRUMENT OPERATING CONDIT;005 COLUMN 1 COLtNel 11 Col.tMEI III Miniense operettesel Nielman Redundenere pequired Operating Action if Channene Required Coltuun I or Colmen 11 Functionet Unit caneet be Met j
l 1.
Nuoteer Feuer-Critteel 3
For 3-Chennel Neletate hotetahdby.eandittete.
Operation -l For 4 Channel operation --i
-86eritteel 3
1 Maintain hd dtandbytt A least i
one oeuree and see intermediate eheneet are ovellablet otherwise maintele 105 Ak/k shutdeun i
morsia.
1 2.
Pressuriser Vertebte 2
1 Notetete leed below 10$ F..F.
tav fressure l
3.
Pressuriser Fised utch pressure a'
1 Maintele hetstandiw eendittees.
j 4.
Pressuriser High Level a
1 Mdiatete hetstahdby esadittees.
O 5.
neestor Coolant Flow -
3-tmop operation 3
188/2ees Metetete toed below 105 F. P.
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6.
Dbleted.s
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C ye modundener le defined as N-M. uhere 2 is the ember er ebennels le operation, and M is the sunber of 1
ohannele is operekton uhtoh when tripped._ vill eeuse en_sutomette shutdown.
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@ ee For operation at (505 of full power 3
j g oes For operettoa et F50$ of full peuer.
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TABLE 3.3.1 (eentinued)
IIISTRt# err CFElutTIto COIIDIT10R8 4
i CotANEI III C0tA#et I C0tANGI11 Dequirse Operating Aeties if lO Mintm. operettemet nimisima nea eemera n ions 1 or c.1= = II
_ tueettomal Unit Chaamela Reentred Commo% be Met p
t' T.
Deleted.
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8.
Itaistain helstandby eendittees.
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Jbleted.
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- 10. Steam Feed-anter Fler Mienateh 3
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Operator shall essene eaetlemodo i
serveillanee ama eeteste emeum1 I
eeren SF required.
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O N
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i 8 Meeuseancr le defined no B-M ehere N is the saber of channels _la operat:,ee. ese N is the number of channele la eserstlee obich, when tr., peed m;,11 cause as setomatie j
g ebutdows.
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3.5.5 Containment Isolati@n Instrumentation Applicability: Applies to instrumentation which actuates the containment sphere isolation valves, containment sphere purge and exhaust valves, and containment sphere instrumentation vent header valves.
Objective:
To ensure reliability of the containment sphere isolation provisions.
Specification:
A.
The instrumentation channels shown in Table 3.5.5-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.5.5-2.
B.
With an instrumentation channel trip setpoint less conservative than the Allowable Values column of Table 3.5.5-2, declare the channel inoperable and apply the applicable Action requirement of Table 3.5.5-1 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint Value.
C.
With an instrumentation channel inoperable, take the action shown in Table 3.5.5-1.
Basis:
The operability of these instrumentation systems ensure that
- 1) the associated action will De initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available from diverse parameters.
The operability of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.
References:
(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.
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TABLE 3.5.5-1 CONTAINMENT ISOLATION INSTRUMENTATION MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION
_C_ontainment Isolation (Valves listed in Table 3.6.2-1) a) Manual 2
1 2
1,2,3,4 D
b) Containment Pressure-High 3/ train 2/ train 2/ train 1, 2, 3 8
c) Sequencer Subchannels 2/ sequencer 2/ sequencer 2/ sequencer 1, E, 3, 4 A
d) Safety injection
- 1) Containment Pressure-High 3/ train 2/ train 2/ train 1, 2, 3 E
- 2) Pressurizer Pressure-Low 3/ train 2/ train 2/ train 1, 2, 3 E
Purge and Exhaust Isolation (POV-9,POV-10,CV-10,CV-40,CV-116) a) Manual 1
1 1
1,2,3,4 C
b) Containment 1
1 1
1,2,3,4 C
Radioactivity-High
TABLE 3.5.5-1 (Continued)
TABLE NOTATION ACTION STATEMENTS ACTION A - With the number of OPERABLE channels one less than the Total Number of Channels, be in HOT STANDBY within 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 />; however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.1.4.
ACTION B - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is demonstrated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; one addit.onal channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.1.4.
ACTION C - With less than the Minimum Channels OPERABLE, operation may continue provided the containment purge and exhaust valves (POV-9 &
POV-10) are maintained closed.
ACTION D - With the number of OPERABLE Channels one less than the Total Number of Channels, restore the inoperable channel 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 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 />.
ACTION E - With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
b.
The Minimum Channels OPERABLE requirements are met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.1.4.
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TABLE 3.5.5-2 CONTAINMENTISdLATIONINSTRUMENTATIONTRIPSETPOINTS
]
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES l
Containment Isolation a) Manual Not Applicable Not Applicable
{
b) Containment Pressure-High 61.4 psig 62.0 psig c) Sequencer Subchannels, Not Applicable Not Applicable d) Safety Injection j
- 1) Containment Pressure-High 61.4 psig 6 2.0 psig
- 2) Pressurizer Pressure-Low 21685 psig 21675 psig 1
Purge and Exhaust Isolation 1
a) Manual Not Applicable Not Applicable b) Containment i
Radioactivity-High 52 x inservice reading
$2.5 x inservice reading 9
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l Item 17 cf this Table will be delcted, TABl.E 4.1.1 (C32t i2ued )
N Channels Action
- Minimus Frequency 17 13.
Boric Acid Tank Calibration At each refueling 12/20/74 Level shutdown Test Once per month during operation 17 14.
Residual Heat
' Calibration At each refueling l12/20/74 Pump Flow shutdown 17 15.
Refueling Tank Calibration At each refueling 12/20/74 14 vel shutdown Test once per month during operation 17 16.
Volume Control Calibration At each refueling 12/20/74 Tank Level shutdown Test bnce per month during operation 17.
EIE'IED 17 18.
Area Radiation Calibration
,0nce per month 12/20/74 Monitora Test once per day 19.
Hydrazine Tank Calibration At each refueling shut-level down 34 Test One per month during
'/1/77 operation NOIE FDR EFEENG: This table was rent =bered Table 4.1.1.1 in Proposed Change No. 83.
4-4 Revised: 5/14/77 7
EDDILDRIGILil
4.1.4 CONTAINMENT ISOLATION INSTRUMENTATION Applicability: Applies to instrumentation which actuates the containment sphere isolation valves, containment sphere purge and exhaust valves, and containment sphere instrumentation vent header val ves.
Objective:
To ensure reliability of the containment sphere isolation provisions.
Specification:
A.
Each in=trumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations for the MODES and at the frequencies shown in Table 4.1.4-1.
Basis:
The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are suf ficient to demonstrate this capability.
References:
(1) NRC letter dated July 2,1980, from D. G. Eisenhut to all pressurized water reactor licensees.
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TABLE 4.1.4-1 CONTAlhMENT ISOLATION INSTRtlMENTATION SURVEILLANCE REQUIREMENTS MODES IN WHICH CHANNEL CHANNEL CHANNEL SURVEllLANCE FUNCTIONAL UNIT CHECK CAllBRATION TEST REQUIRED Jontainment Isolation (Valves listed in Table 3.6.2-1) a) Manual N.A.
N.A.
M(1) 1, 2, 3, 4 b) Containment N.A.
R M(2) 1, 2, 3 Pressure-High c) Sequencee Subchannels N.A.
N.A.
M 1,2,3,4 d) Safety injection
- 1) Containment Pressure-High N.A.
R M(2) 1, 2. 3
- 2) Pressurizer Pressure-Low N.A.
R M
1, 2, 3, 4 Purge and Exhaust Isolation TPOV-9, POV-10, CV-10, CV 40, CV-116) a) Manual N.A.
N.A.
M(1) 1, 2, 3, 4 b) Containment Radioactivity-High S
R M
1,2.3,4
TABLE 4.1.4-1 (Crntinuid)
TABLE NOTATION (1) Manual actuation switQes shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual safegu 'ds actuation shall receive a CHANNEL TEST at least once per 31 days.
(2) The CHANNEL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.
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3.5.6 Accident M:nitoring Instrumentatien Applicability: Applies to the accident monitoring instruments shown in Table 3.5.6-1 for MODES 1, 2 and 3.
Objective:
To ensure reliability of the accident monitoring instrumentati on.
Specification:
A.
The accident monitoring instrumentation channels shown in Table 3.5.6-1 shall be OPERABLE.
B.
With the number of OPERABLE accident monitoring instrumentation channels less than the Total Number of Channels shown in Table 3.5.6-1, either restore the inoperable channel (s) to OPERA 8LE 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 />.
C.
With the number of DPERABLE accident monitoring instrumentation channels less than the MINIMUM CHANNELS OPERABLE requirements of Tabel 3.5.6-1, 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 />.
Basis:
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 Power Plants to Assess Plant l
Conditions During and Following an Accident," December 1975 and h" DREG-0578, "TMI-2 Lessions Learned Task Force Status l
Report and Short-Term Recommendations.'
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References:
(1) NRC letter dated July 2,1980, from D. G. Eisenhut to all pressurized water reactor licensees.
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TABLE 3.5.6-1 ACCIDENT MONITORING INSTRUMENTATION i
MINIMUM i
TOTAL NO.
CHANNELS INSTRUMENT OF CHANNELS OPERABLE l
Pressurizer Water Level 3
2 i
Auxiliary Feedwater Flow Indication
- 2 per steam generator 1 per steam generator Reactor Coolant System Subcooling Margin Monitor 2
1 PORV Position Indicator 1/ valve 1/ valve PORV Block Valve Position Indicator 1/ valve 1/ valve Safety Valve Position Indicator 1/ val ve 1/ valve
- Auxiliary feedwater flow indication for each steam generator is provided by one channel of steam generator level and one channel of auxiliary feedwater flow rate.
These comprise the two channels of auxiliary feedwater flow indication for each steam generator.
4.1.5 Accident Monitoring Instrumentation Applicability: Applies to the accident monitoring instruments shown in Table 4.1.5-1 for MODES 1, 2 and 3.
Objectives:
To ensure the reliability of the accident monitoring instrumentation.
Specification:
A.
Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHAMMEL CHECK and CHANNEL CALIBRATION operations at the frequencie-shown in Table 4.1.5-1.
Basis:
The surveillance requirements specified for these systems ensure that the overall functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
References:
(1) NRC letter dated July 2,1980, from D. G. Eisenhut to all pressurized water reactor licensees.
TABLE 4.1.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANilEL CHANNEL INSTRUMENT CHECK CALIBRATION i
Pressurizer Water Level M
R Auxiliary Feedwater Flow Indication
- M R
l Reactor Coolant System Subcooling Margin Monitor M
R PORV Position Indicator M
R PURV Block Valve Position Indicator M
R Safety Valve Position Indicator M
R a
1
- See footnote of Table 3.5.6-1.
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3.1.5 Prassurizer Relief Valves Applicability: Applies to the power operateo relief valves (PORVs) and their associated block valves for MODES 1, 2 and 3.
Objective:
To ensure reliability of the PORVs and block valves.
Specification:
A.
Two PORVs and their associated block valves shall be OPERABLE.
l B.
With one or more PORV(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 PORV(s) to OPERABLE status or close the l
associated block valve (s) and remove power from the block l
valve (s); otherwise, be in at least HOT STANDBY withir. t59 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 />.
C.
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 0FERABLE status or l
close the block valve (s) and remove power from the block valve (s); otherwise, be in at least H0T 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 />.
l Basis:
The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a l
relief valve become inoperable.
The air supply for both the l
relief valves and the block valves is capable of being supplied from a backup passive nitrogen source to ensure the ability to seal this possible RCS leakage path.
References:
(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.
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4.1.6 Pressurizer Raliof Valv2s Applicability: Applies to the power operated relief valves (PORVs) and their associated block valves for MODES 1, 2 and 3.
Objective:
To ensure the reliability of the PORVs and block valves.
Specification:
A.
Each PORV shall be demonstrated OPC.iABLE:
1.
At least once per 31 days by performance of a CHANNEL TEST, which may include valve operation, and 2.
At least once per 18 months by performance of a CHANNEL CALIBRATION.
B.
Each block valve shall be demonstrateo OPERA 8LE at least once per 92 days by operating the valve through one complete cycle of full travel.
C.
The backup nitrogen supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by transferring motive power from the normal air supply to the nitrogen supply and operating the valves through a complete cycle of full travel.
Basis:
The prier operated relief /alves (PORVs) operate to relieve RCS pressure below the setting of the pressurizar code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The air supply for both the relief valves and the block valves is capable of being supplied from a backup passive nitrogen source to ensure the ability to seal this possible RCS leakage path.
References:
(1) NRC letter dated July 2,1980, from D. G. Eisenhut to all pressurized water reactor licensees.
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3.1.6 Pressurizer i
Applicability: Applies to the pressurizer heaters and pressurizer water level for MODES 1, 2 and 3.
Objective:
To ensure that pressurizer heaters are available during a loss of offsite power condition.
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Speci fication:
A.
The pressurizer shall be OPERABLE with at least 125 l
kilowatts of pressurizer heaters r.nd a water level between 5 percent and 70 percent.
B.
With the pressurizer inoperable due to the loss of capability to energize the pressurizer heaters from an emergency diesel generator, either restore the capability to energize the pressurizer heaters from an emergency diesel generator within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or 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 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 least 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 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 />.
Basis:
The requirement that 125 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency diesel generator provides assurance l
that these heaters can be energized during a loss of offsite-power condition to raintain natural circulation at HOT STANDBY.
References:
(1) NRC letter dated July 2,1980, from D. G. Eisenhut to all pressurized water reactor licensees.
(2) SCE letter dated October 17, 1979, from J. H. Drake to D. G. Eisenhut, " Responses to NRC Requirements Related to the Three Mile Island Accident," Item 2.1.1 of the enclosure.
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4.1.7 Pressurizer 9
Applicability: Applies to pressurizer heaters and pressurizer water level for MODES 1, 2 and 3.
Objective:
To ensure proper pressurizer water volume and to ensure the capability to energize the pressurizer heaters from the emergency diesel genarator.
Specification:
A.
The pressurizer water level shall be determined to be between 5% and 70% at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
B.
The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal supply to the emergency diesel generator and energizing the' heaters.
Basis:
The requirement that the pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency diesel generator provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.
References:
(1) NRC letter dated July 2,1980, from D. G. Eisenhut to all pressurized water reactor licensees.
Appendix A, Specification 3.6, CONTAINMENT, will be retitled and subdivided as follows:
3.6 Containment Systems 3.6.1 Containment Sphere Appendix A, Specification 4.3, CONTAINMENT TESTING, wil be retitled and subdivided as follows:
4.3 Containment Systems 4.3.1 Containment Testing Specification 4.3.II.C.1 of the existing Technical Specifications reads:
C.
Isolation Valve Testing 1.
Tests All isolation valves shall be tested for operability and leak rate characteristics.
Isolation valves normally operating with pressure less than 50 psig shall be tested at an initial pressu.e (beginning of test) of 49.4 psig.
Specification 4.3.II.C.1 will be changed to:
1.
Tests All isolation valves shall be tested for leak rate
-- characteristics.
Isoittion valves normally operating with pressure less than 50 psig shall be tested at an initial pressure (beginning of test) of 49.4 psig.
3.6.2 Containment Isolation "va',e_s Applicability: Applies to :he containment isolation valves listed in Table 3.6.2-1 for MODES 1, 2, 3 and 4 Objective:
To provide assurance that containment isolation will function when initiated by appropriate sensors.
Speci fication:
A.
The containment isolation ve.!ves specified in Table 3.6.2-1 shall be OPERABLE.
B.
With one or more of the '. solation valve (s) specified in Table 3.6.2-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
1.
Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or 2.
Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one deactivated automatic valve seccred in the isolation position, or 3.
Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange, or 4
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 />.
Basis:
The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
References:
(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.
1
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TABLE 3.6.2-1 POWER OPERATED OR AUTUMATIC CONTAlletENT ISOLATION VALVE SUf94ARY DESCRIPi10N INSIDE SPHERE AL IGNMENT*
DUTSIDE SPHERE ALIGNMENT
- 1.
Sphere Sump Discharge CV-102(SV-108)
B CV-103 (SV-109 A
2.
RL5 Dr ik utscharge CV-104 (SV-1100 B
CV-10$ (SV-111 A
3.
RCS Dr ik Vent CV-106 (SV-112) 8 CV-107 (SV-113 A
l 4
N2 to RCS Drain Tank and PRT CV-b36 A
CV-b35 B
5.
DRMS 1211/1212 Sphere CV-147(SV-1212-1) 8 SV-1212-9 A
Sample Supply 6.
DRMS 1211/1212 Sphere CV-146 (SV-1212-6) 8 SV-1212-8 A
Sample Return 7.
A Stm. Gen. Stm. Sample None SV-119 A
8.
B Stm. cen. Stm. Sample None SV-120 A
9.
C Stm. Gen. Sta. Sample Mone SV-121 A
i
- 10. A Stm. Gen. Blowdown Sample None SV-123 A
i
- 11. B Stm. Gen. Blowdown Sample None SV-122 A
12 C Stm. Gen. Clowdown Sample None SV-124 A
- 13. Service Water to Sphere CV-537 A
CV-115(SV-126)
B
- 14. Service Air to Sphere Check Valve SV-125 A
- 15. SI Loop C Vent SV-702B A
SV-702A B
- 16. SI Loop B Vent SV-702D A
SV-702C B
i
- 17. PRT Gas Sample CV-948**
A CV-949 i SV-9491 8
SV-992J)
- 18. RC Loop Sample (CV-955, CV-956, CV-962)**
A CV-957 L SV *57 8
- 19. Pressurizer Sample (CV-9bl. CV-953)**
A CV-992 L B
- 20. Sphere Purge Air Supply POV-9 LSV-29)
A
- 21. Sphere Purge Air Outlet POV-10 LSV-30)
A
- 22. Sphere Equalizing / Sphere Vent CV-116 (SV-27)
B CV-10 (SV-28)
A Inst. Air Vent CV-40 (SV-19)
B 1
- 23. Primary Makeup to Press CV-b33 A
CV-634 5
Rif. Ik
- 24. Cont. Cooling Out CV-515**
A
- 25. Cont. Cooling In CV-bl6**
5
- 26. N2 Supply to PORV CV-b32**
B Check Valv.
- 27. Letdown CV-525**
A CV-526**
8 1
- 28. Seal Water Return CV-b27**
A CV-528**
B
- 29. Hydrogen Monitoring System SV-3004 B
SV-2004 A
Logic Nest C. Train A is aligned to power train F; Logic Mest D. Train B is aligned to power train G.
These valves do not receive an automatic containment isolation signal. They are operated by remote manual switch (RMS).
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4.3.2 CONTAINMENT ISOLATION VALVES Application:
Applies to the containment isolation valves listed in Table 3.6.2-1 for MODES 1, 2, 3 and 4.
Objective:
To ensure reliability of containment isolation valves.
Specification:
a.
The isolation valves specified in Table 3.6.2-1 shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by performance of a cycling test.
B.
Each isolation valve specified in Table 3.6.2-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
1.
Verifying that on containment isolation test signal, each automatic isolation valve actuates to its isolation position.
2.
Verifying that on a containment radiation-high test 1
signal, each purge supply and purge outlet automatic valve actuates to its isolation position.
C.
Each power operated or automatic valve of Table 3.6.2-1 shall be determined to be OPERABLE when tested la accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
Basis:
The OPERABILITY of the containment isolation valves ensures l
that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the centainment. Containment isolation ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
References:
(1) NRC letter dated July 2,1980, from D. G. Eisenhut to all pressurized water reactor licensees.
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3.5.7 Am idwater Instrumentation Applicaoility: Apr.its to automatic initiation of the auxiliary feedwater pumps Objective:
To ensure reliability of automatic initiation of the auxiliary feedwater pumps.
Specification:
A.
The instrumentation channels shown in Table 3.5.7-1 shall be OPERABLE with their trip setpoints set consistent with the Trip Setpoint column of Table 3.5.7-2.
B.
With an instrumentation channel trip setpoint less I
cunservative than the value shown in the Allowable Values column of Table 3.5.7-2, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.5.7-1 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint Value.
C.
With one instrumentation channel inoperable, take the action shown in Table 3.5.7-1.
D.
With more than one channel inoperable, take ACTION G of Table 3.S.7-1.
Basis:
The OPERABILITY of the auxiliary feedwater instrumentation ensures that 1) the associated action will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and
- 4) sufficient s>.; tem functional capability is available from diverse parameters.
The OPERABILITY of this instrumentation is required to provide the overall reliability, redundancy, and diversity assumed available for the protection and mitigation of accident and l
transient conditions. The operation of this instrunentation is consistent with the assumptions used in the accident analyses.
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References:
(1) NRC letter cated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.
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TABLE 3.S.7-1 AUXILIARY FEEDWATER INSTRUMENTATION MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION a) Stm. Gen. Water Level-Low i.
Start Motor Driven Pumps 3
2 2
1,2,3,4 F
- 11. Start Turbine-Driven Pump 3
2 2
1,2,3,4 F
ACTION F - With the number of OPERABLE channels one less than the Total i
Number of Channels, operation may proceed until performance of the next required CHANNEL TEST provided the inoperable channel is placed in the tripped condition within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i.
ACTION G - With neore than one channel inoperable, an operator shall assume continuous surveillance and actuate manual initiation of auxiliary feedwater, 'f necessary. Restore the system to no more than one channel inopers51e within 7 days, or be in HOT STANDBY within the following 6 b~
- 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 />.
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TABLE 3.5.7-2 AUXILIARY FEEDWATER INSTR,'IMENTATION TRIP SETPOINTS 4
I FUNCTIONAL UNIT TRIP SETPolNT ALLOWABLE VALUES I
a) Steam Generator Water Level-Low 2:5% of narrow range 2 0% of narrow range 4
instrument span each instrument span each steam generator steam generator 4
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'4.1.8 Auxiliary Feedwater Instrumentation Applicability: Applies to the instruments shown in Table 4.1.8-1.
Objective:
To ensure reliability of automatic initiation of the auxiliary feedwater pumps.
Specification:
A.
Each instrument **. ion channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations for the MODES and at the frequencies shown in Table 4.1.8-1.
Basis:
The surveillance requirements specified for this instrumentation ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the _
minimum frequencies are sufficient to demonstrate this capability.
References:
(1) NRC letter dated July 2, 1980, from D. G. Eisenhut to all pressurized water reactor licensees.
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I TABLE 4.1.8-1 AUXILIARY FEEDWATER INSTRUMENTATION SURVEILLANCE REQUIREMENTS i
l MODES IN WHICH CHANNEL CHANNEL CHANNEL SURVEILLANCE FUNCTIONAL UNIT CHECK CAllBRATION TEST REQUIRED i
l a) Steam Generator Water Level-Low S
R M
1,2,3,4 4
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On the following pagas, ccpies of Prepostd Change No. 87 and Propos d Change No. 90 are provided. Both of these proposed changes were submitted as part of SCE Amendment 88 which was transmitted on February 8,1980 by letter from R. Dietch to H. R. Denton.
The existing technical specification 4.4.E should be deleted and Proposed Change No. 87 should be cancelled, as both are superseded by new technical specification 4.1.9 which follows.
Item 20, page 1, and paragraph 2, page 3, of Proposed Change No. 90 should be deleted, as both are superseded by new technical specification 4.1.8 which follows.
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, II)R REFERENG ONLY Description of Proposed Change and Safety Analysis Proposed Change No. 87 to the Technical Specifications Provisional Operating License DPR-13 This request would. revise Section 4.4.E. of the Appendir A Technical Specifications for San Onofre Unit 1.
l Reason for Proposed Change
{
These changes are submitted in response to NRC Staff requests contained in to a November 15, 1979 letter from D. G. Eisenhut to J. H. Drake.
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These requests were based on the NRR Bulletins and Orders Task Force review of operating reactors in light of tht accident at Three Mile Island, Unit 2.
f Existing Specifications l
Ibe existing Specifications are constit'uted in Section 4.4 of Appendix A Technical Specifications for Provisional Operating License DPR-13 I
Proposed Specifications 1
Technical Specification 4.4.E. would be revised to read as follows:
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"E.
Auxiliary Feedwater System At least every fourteen (14) days when the reactor coolant system 1.
pressure is greater than 500 psig, the auxiliary feedwater pumps shall be started to demonstrate satisfactory operation.
At least once every thirty-one (31) days when the reactor. coolant 2.
system pressure is greater than 500 psig, the electric driven auxiliary feedwater pump shall be started and the normal flow path motor operated discharge valve shall be opened using the automatic and the remote manual actuation circuitry. This testing may be done in conjunction with 4.4.E.1 above.
When the reactor coolant system pressure remains less than 500 psig 3
for a period longer than fourteen (14) days, a flow test shall be performed to verify the normal flow path from the condensate storage tank to each steam generator using the motor driven auxiliary feedwater pump prior to increasing reactor coolant system pressure
. The flow test shall be conducted with the auxiliary above 500 psig.
As soon as steam feedwater system valves in their normal alignment.
becomes available, the steam driven auxiliary feedwater pump shall be I
started to demonstrate satisfactory operation.
i At least once every thirty-one (31) days when the auxiliary feedwater
[
system is required to be operable, an inspection shall be made to 4
verify that manuai valves in the auxiliary feedwater system suction piping and tne normal path from the auxiliary feedwater pumps to the main feedwater header that could interrupt all AFW flow are locked i
open.
IUR REFERENCE ONIX
- 1 S.
At least once every eighteen months, all normally closed manual valves in the alternate auxiliary feedwater system suction line and in the emergency flow path from the auxiliary feedwater pumps to the steam generator feedwater lines shall be demonstrated operable."
Safety Analysis The proposed revisions to the Technical Specifications require additional surveillance verifications of AFW system valve positions and component This surveillance activity is intended to decrease the operability.
probability that equipment failure or system misalignment will go undetected while the AFW system is required to be in standby status.
Accordingly, it is concluded that (1) the proposed change does not involve an unreviewed safety question as defined in 10CFR5v.59, nor does it present significant hazards considerations not described or implicit in the Final Safety Analysis, and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change.
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R)R REFERENCE ONLY DESCRIPTION OF-PROPOSED CHANGE AND SAFETY ANALYSIS PROPOSED CHANGE NO. 90 TO THE TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE DPR-13 This is a request to (1) revise the definition of Containment Integrity as contained in Section 1.0; (2) add requirements for calibration and testing of Auxiliary Feedwater Flow and Condensate Tank level instrumentation in Table 4.1.1 of Section 4.1, and; (3) revise the requirements for the contain-ment isolation system design in Section 5.2 of the Appendix A. Technical Specifications for San Onofre Unit 1.
Reason for Proposed Change These changes are submitted to provide revisions and additions to reflect station modifications being performed during the January,1980 outage to complete the Category A NRC TMI Short Term Lessons Learned Requirements as described in our letters dated January 17 and 23.1980.
Existing Specifications The existing specifications are as constituted in Sections 1, 4, and 5 of the Appendix A, Technical Specifications for Provisional Operating License DPR-13 Proposed Specifications l
The definition of " Containment Integrity" in Technical Specifiation 1.0 would be revised to read:
i
" Containment Integrity":
Containment Integrity means that all of the conditions below are satisfied:
(1) All manual containment isolation valves (or blind flanges) are closed.
(2) The equipment door is properly closed.
(3) At least one door in each personnel air lock is properly closed.
(4) All automatic and remote manual containment isolation valves are operable."
Technical Specification 4.1 would be revised by adding items 20 and 21 to Table 4.1.1 to read :
l Channels Action Minimum Frequency "20.
Auxiliary Feedwater Test Once per month flow during operation
FOR REFERENCE ONLY 21.
Condensate Tank Calibration At each refueling level shutdown Test Once per month during operation" The last paragraph of Technical Specification 5.2 would be revised to read:
"The automatically actuated containment isolation valves shall be designed to close upon high-pressure in the containment (set point no higher than 5 psig) or upon safety injection initiation.
In addition, design provisions shall prevent automatic reopening of any isolation valves upon reset of the containment isolation signal. The actuation system shall be designed such that no single component failure will prevent containment isolation if required."
Safety Analysis Each of the proposed Technical Specification revisions discussed above is required as part of the implementation of the NRC TMI Short Term Lessons Learned Requirements. The basis for each revision is discussed below:
1.
By letters dated December 17, 1979 and January 17, 1980, the results of the Essential /Non-essential study of the containment isolation systems were provided to the NRC as required by NUREG-0578, Section 2.1.4 Based on the results, two automatically isolated systems which were previously identified as non-essential systems have been identified as essential systems (i.e., those required to mitigate an accident or which, if unavailable, could increase the magnitude of the event).
These systems are:
Tne turbine plant cooling water supply and return lines may be a.
required to support extended operation of the reactor coolant pumps since they supply cooling water to the reactor coolant pump enclosure air conditioning units, and The nitrogen supply line to the Pressurizer Relief Tank currently b.
provides a redundant source of pneumatic motive power to the power-operated relief valves and will provide a similar function to their associated block valves.
As discussed in our January 17, 1980 letter, these two systems will be modified to provide remote-manual containment isolation capability consistent with other essential systems.
The revision of the
" Containment Integrity" definition requires that the remote-manual isolatico valves be operable, and allows them to be open containment during operating conditions which require " Containment Integrity."
Based on our review of the Technical Specifications which might be affected by the implementation of the NHC TMI Short Term Lessons Learned Requirements, we have determined that the definition of " Containment Integrity" should have been revised as part of our thorough review of the San Onofre Unit 1 containment isolation design completed in 1
IOR REFERENCE ONLY,
-3 April, 1976 for the Sphere Enclosure Project and assessment of compliance with 10CFH50, Appendix J.
As a result of this review, the containment isolation design was modified to improve the leak tightness capability of the containment in the event of an accident requiring containment isolation by installation of new automatic and remote-manual valves. These new valves, as well as existing remote-manual valves, were designated as containment isolation valves. The remote-manual valves provide the ability to isolate essential systems, if necessary, following an accident to improve the leak tightness of containment.
However, the definition of " Containment Integrity" was not revised to reflect the improved containment isolation design utilizing both automatic and remote-manual valves as containment isolation valves.
2.
The auxiliary feedwater flow test requirement and condensate tank level calibration and test requirements added to Table 4.1.1 specify the minimum frequency and type of surveillance to be applied to the instrumentation. These calibration and test requirements are consistent with existing surveillance requirements for instrumentation installed in other systems.
3.
The revision to Containment Design features (Technical Specification 5.2) accurately describes the containment isolation design.
Based on the above, it is concluded that (1) the proposed change does not involve an unreviewed safety question as defined in 10CFR50.59, nor does it present significant hazard considerations not described or implicit in the Final Safety Analysis, and (2) there is reasonable assurance that the health and safety of the public will not be endangered by the proposed change.
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3.4.3 Auxiliary Feedwater System Applicability: Applies to the motor driven auxiliary feedwater pump and the turbine driven auxiliary feedwater pump for MODES 1, 2 and 3.
Objective:
To ensure the availability of auxiliary feedwater to remove decay heat.
Specification:
A.
Both steam generator auxiliary feedwater pumps and i
associated flow paths shall be OPERABLE as follows:
1.
One auxiliary feedwater pump capable of being powered from an emergency electrical power source, and 2.
One auxiliary feedwater pump capable of being powered from an CTERABLE steam supply system.
B.
With one auxiliary feedwater pump inoperable, restore both auxiliary feedwater pumps (one capable of being powered from an emergency electrical power source and one capable of being powered by an OPERABLE steam supply system) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT i
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 SHUTDONN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Basis:
The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than
~
3500F from nornal operating conditions in the event of a total loss of off-site power.
References:
(1)
NRC letter dated July 2,1980 from D. G. Eisenhut to all pressurized water reactor licensees.
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4.1.9 Auxiliari'Feedwater Pumps and Valves Applicability: Applies to the motor driven auxiliary feedwater pump, the turbine driven auxiliary feedwater pump, and auxiliary feedwater valves for MODES 1, 2 and 3.
Objective:
To ensure the reliability of the auxiliary feedwater system.
Specification:
A.
Each auxiliary feedwater pump shall be demonstrated OPERABLE by testing each pump in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55alg), except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.554(g)(6)(1).
B.
At least once per 31 days, verify that each non-automatic valve in the flow path that is not locked, sealed, or othentise secured in position, is in its torrect position.
C.
Each auxiliary feedwater pump shall be demonstrated OPERABLE at least once per 18 months during shutdown by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of each auxiliary feedwater actuation test signal.
2.
Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of each auxiliary feedwater actuation test signal.
Basis:
The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 3500F from normal operating conditions in the event of 4 total loss of off-site power.
+
The electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 208 gpm at a pressure of 1100 psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 260 gpm at a pressure of 1100 psi; to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 3500F when the Residual Heat Removal System may be placed into operation.
References:
(1) NRC letter dated July 2, 1980 from D. G. Eisenhut to all pressurized water reactor licensees.
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- 3.6.3 Hydrogen Monitors and Hydrogen Recombiners Applicability: Applies to containment sphere hydrogen monitors and hydrogen recombiners for MODES 1 and 2.
Objective:
To ensure the capability to maintain the hydrogen concentration within the containment sphere below its flammable limit during post-LOCA co.1ditions.
Specification:
A.
Two independent containment hydrogen monitors shall be OPERABLE.
B.
With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLE status within 30 days or 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 />.
C.
Two independent containment hydrogen recombiner systems snall be OPERABLE.
D.
With one hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Basis:
The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain tre hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit (or the purge system) is capable of controlling the expected hydrogen generation associated with radiolytic decomposition of water and corrosion of metals within containment.
(Cumulative operation of the purge system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31-day period is sufficient to reduce ti.e buildup of moisture on the adsorbers and HEPA filters.) These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA," March 1971.
The hydrogen mixing systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent 1ccalized accumulations of hydrogen from exceeding the flammable limit.
References:
(1) Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA,"
March, 1971.
,,_.._,.,m
4.3.3 Hydrogen Monitors and Hydrogen Recembiners Application:
Applies to containment sphere hydrogen monitors and hydrogen recombiners for MODES 1 and 2.
Objective:
To ensure reliability of the equipment and systems required for the detection and control of hydrogen gas.
Specification:
A.
Each hydrogen monitor shall be demonstrated OPERABLE at least once per 92 days.on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gases containing:
1.
One volume percent hydrogen, balance nitrogen.
2.
Four volume percent hydrogen, balance nitrogen.
B.
Each hydrogen recombiner system shall be demonstrated OPERABLE at least once per 6 months by verifying that the minimum heater sheath temperature increases to greater than or equal to 7000F within 90 minutes. Upon reaching 7000F, increase the power setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 60 Kw.
C.
Each hydrogen recombiner system shall be demonstrated OPERABLE at least once per 18 months by:
1.
Performing a CHANNEL CALIBRATION of all recombiner instrumentat. ion and control circuits.
i 2.
Verifying through a visual examination that there is no evidence of abnormal conditions within the l
recombiner enclosure (i.e., loose wiring or structural connections, deposits or foreign materials, etc.), and 3.
Verifying the integrity of all heater electrical circuits by performing a resistance to ground test following the test in Specification B above. The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.
Basis:
The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this I
equipment will be available to maintain the hydrogen concentration witnin containment below its flammable limit during post-LOCA conditions. Either recombiner unit (or the purge system, is capable of controlling the expected hydrogen generation associated with radiolytic decomposition of water and corrosion of metals within containment.
(Cumulative operation of the purge system with the heaters on for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31-day period is sufficient to reduce the buildup of
~.. -
.~
moisture on the adsorbers and HEPA filters). These hydrogen control systems are consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA," March,1971.
References:
(1) Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA,"
March, 1971.
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Table 6.2.2.2 will be revised as follows:
TABLE 6.2.2.2 MINIMUM SHIFT CREW COMPOSITION #
l LICENSE CATEGORY APPLICABLE MODES QUALIFICATIONS 1, 2, 3 & 4 5&6 SR0 1
1*
R0 2
1 Non-Licensed Auxiliary 1
1 Operator Shift Technical Advisor 1
None Required 1
- Does not include the licensed Senior Reactor Operator or Senior Reactor Operator limited to Fuel Handling, supervising CORE OPERAIIONS.
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- Shift crew composition may be one less than the minimum requirements for 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 accomodate unexpected absence of on-duty shift crew members provided inanediate action is taken to restore the shift crew ccmposition to within the mini;num requirements of Table 6.2.2.2.
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.
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- 6.3 Unit Staff Qualifications 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the Health Physics Manager who shall meet or exceed the qualifications of Reguietory Guide 1.8, September 1975 and (2) the Shift Technical Advisor who shall hnve a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant desiga, and response ano analysis of the plant for transients and accidents.
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- The following license conditions will be added to Provisional Operating License No. UPR-13 Section 3:
1.
Systems Integrity The licensee shall implement a program to reduce leakage from systems outside containment that would or could cortain highly radioactive fluids during a serious transient or accident to as low as practical levels.
This program shall include the following:
1.
Pr.ovisions establishing preventative maintenance and periodic visual inspection requirements, and i
2.
Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
J.
Iodine Monitoring The licensee 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.
Trair.ing of personnel, 2.
Procedures for monitoring, and 3.
Provisions for neintenance of sampling and analysis equipment.
K.
Backup hethod for Determining Subcooling Margin The licensee 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 l
2.
Procedures for monitoring.
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