ML20067E777
| ML20067E777 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/12/1991 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20067E778 | List: |
| References | |
| NUDOCS 9102180171 | |
| Download: ML20067E777 (10) | |
Text
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i UNITED STATES NUCLEAR REGULATORY COMMISSION I
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WASmNG TON, D. C. 20555
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l PUBLIC SERVICE ELECTRIC & GAS. COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 SALEM GENERATING STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING. LICENSE Amendment No. 117 License No. DPR-70 1.
The Nuclear Regulatory Comission (the Comission or the NRC) has found that:
A.
The application for amendment filed by the Public Service Electric &
Gas Company, Philadelphia Electric Company,(Delmarva Power and L Company and Atlantic City Electric Company the licensees) dated February 23, 1990, and supplemented by letters dated June 28, 1990 and August 8, 1990, complies with the standards and requirements 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; 1
C.
There is reasonable assurance: (i) that the activities authorized by i
this amendment can be conducted without endangering the health and l
safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon i
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, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-70 is hereby i
l amended to read as follows:
I 9102180171 910212 PDR ADOCK 05000272 P
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(2) Technical Specifications and. Environmental Protection-Plan The Technical Specifications-contained in. Appendices:A and 8, as i
revised through-Amendment.No.117., are hereby -incorporated. in-- the license. The licensee shall: operate the facility 5in accordance with<
'the Technical Specifications.
3.
This license amendment is effective as of'its date of issuance to-be-implemented prior to_startup from the' ninth refueling. outage scheduled to!
begin February,1991.
FOR THE NUCLEAR REGULATORY-COMMISSION--
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Walter R. Butler,' Director j
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Division of Reactor Projects - I/II Project Directorate ~I-2 i
Attachment:
Changes to the Technical Specifications Date of Issuance: February 12, 1991 J
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(2) Technical. Specifications and Environmental. Protection Plan The Technical Specifications contained in Appendices A and'B, as revised through Amendment No.117:, are hereby incorporated-in the license. The licensee shall operate _the facility in accordance with the Technical. Specifications.
3.
This license amendment is effective as of its date of issuance to be implemented ~ prior to startup from the ninth refueling outage scheduled to begin February,1991.
LFOR THE NUCLEAR REGULATORY COMMISSION-b d.- m t
I Wa ter R. Butler, Director Project Directorate I Division of Reactor. Projects --I/II-
Attachment:
Chang >es to the Technical
. Specifications Date of Issuance: February 12, 1991 l
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i ATTACHMENT TO LICENSE AMEN 0 MENT NO.117 FACILITY OPERATING LICENSE NO. OPR-70 DOCKET NO. 50-272 Revise Appendix A as follows:
Remove Pages Insert Pages 3/4 3-53 3/4 3-53 3/4 3-54 3/4 3-54 3/4 3-54a 3/4 3-55 3/4 3-55 3/4 3-56 3/4 3-55a 3/4 3-56 3/4 3-56a 3/4 3-56a 3/4 3-57a 3/4 3-57a l
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TNSTRUMENTATTON ACCIDENT MONITORING INSTRUMENTATION
' LIMITING CONDITION;FOR OPERATION-3.3.3.7 The accident monitoringiinstrumentation channels shown in Table l'
^i 3.1-11 shall.be operable.
APPLICABILITY: MODES'1, 2, and 3.
ACTION:
a.
As shown in. Table 3.3 11.
l-b.
The provisions -of Specification 3.0.4,are not applicable.
i SURVEILLANCE REQUIREMENTS
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4.3.3.7 Each accident monitoring instrumentation channel'shallebe-
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demonstrated OPERABLE by perfc?.mance_of_the CHANNEL CHECK AND CHANNEL CALIBRATION operations at theifrequencies shown in Table 4.3 11.
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I SALEM - UNIT 1 3/4 3 53 Amendment No.117 I
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.us TABLE 3.3-11 M
ACCIDENT MONITORING INSTRUMENTATION c3 REQUIRED MINIMUM NO. OF TNSTRUMENT NO. OF
_ CHANNELS e
CHANNELS ACTION 1.
Reactor Coolant Outlet Temperature -
2 T
(Wide Range) 1 1, 2 2.
Reactor Coolant Inlet Temperature -
2 1
T
-(Wide Range) 1, 2 3.
Reactor Coolant. Pressure (Wide Range) 2 1
1, 2 4.
Pressurizer Water Level 2
1 1,
2-3 5.
Steam Line Pressure 2/ Steam Generator 1/ Steam Generator 1, 2 u
6.
Steam Generator Water Level (Narrow 2/ Steam Generator.
1/ Steam Generator' 1, 2
)
Range)
O.
7'.
Steam Generator Water _ Level (Wide 4-(1/ Steam Generator) 3 (1/ Steam Generator) 1, 2
Range) 8.- Refueling Water Storage Tank. Water
'2
' Level 1
1, 2 9.
Boric Acid Tank Solution Level
'2-(1/ tank) 1 (1/ Tank) 3 10.-Auxiliary Feedwater Flow Rate 4 (1/ Steam Generator).
k 3 (1/ Steam Generator) 4, 6 II. Reactor Coolant System Subcooling w
2 E.
Margin Monitor 1
1, 2 w
E
- 12. PORV' Position Indicator 2/ valve **
1 1, 2 N
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g TABLE 3.3-11 (CONTINUED)
I
- >cy A_CpIDENT MONITORING INSTRUMENTATION f
c REQUIRED MINIMUM 3
NO. OF NO. OF d
INSTRUMENT CHANNELS CHANNELS ACTION 9
- 13. PORV Block Valve Position Indicator 2/ valve **
1 1, 2
- 14. Pressurizer Safety Valve Position 2/ valve **
1 1, 2 Indicator
- 15. Containment Pressure - Narrow Range 2
1 1, 2
- 16. Containment Pressure - Wide Rarige 2
1 7,
2
- 17. Containment Water Level -
2 1
7, 2
(
Wide Range A
Y
- 18. Core Exit Thermocouples 4/ core quadrant 2/ core quadrant 1, 2
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- 19. Reactor Vessel Level Instrumentation 2
1 8
System (RVLIS) 1, 2
(**) Total number of channels is considered to be two (2)'with one (1) of the channels being any one'(1) of the following alternate means of determining PORV, PORV Block, or Safety Valve positions. Tailpipe Temperatures for the valves, Pressurizer Relief Tank Temperature Pressurizer Relief Tank Level OPERABLE.
(***)
Action 8 remains in effect until startup from the 10th refueling outage at which time, PSE&G will g
install the upgraded RVLIS.
Upon expiration, Actions 1 and 2 will apply.
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TABLE 3.3 11 (continuedt l
I TABLE NOTATION
. ACTION 1 With the number of OPERABLE accident monitoring i
channels less than-the Required Number of Channels shcun in Table 3.3 '11. restore-the inoperable-j' channel (s) to OPERABLE status within 7. days, or be in HOT SEJTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 2 Wich the number of OPERABLE accident monitoring-channels less than the MINIMUM Number of Channels I
shown in Table 3.3-11,- restore the inoperable
-l channel (s) to OPERABLE status within 48-hours or be in' 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 />.
ACTION 3 With the number of OPERABLE channels: one less than the Required Number of Channels shown in Table 3.3 11. operation may. proceed provided that'the.
l.
~ Boric Acid Tank associated with the remaining.
1 OPERABLE channel-satisfies all : requirements-of -
Specification 3.-l.2.8.a.
ACTION 4 With the number of OPERABLE channels one less than the Required Number of Channels shown in Table 3.3.11 operation may proceed provided that an
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.i OPERABLE Steam Cenerator Wide Range Level channel is available as an alternate means of indication-for the Steam Cenerator with no OPERATABLE -
Auxiliary Feedwater Flow Rate channel, l
s ACTION 5 With the number of OPERABLE channels less'than the
. Required Number of-Channels show in Table 3.3-11, operation may proceed provided that Steam Tables are available in the~ Control Room'and the following.
i Required Channels shown in Table-3.3 11 are OPERABLE to provide an alternate means of
[
calculating Reactor Coolant System subcooling-margin:
Reactor Coolant Outlet Temperature T
a.
(Wide Range)
HOT b.
Reactor Coolant Pressure (Wide Range)
SALEM UNIT 1 3/4 3 56 Amendment No. 117-1
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-TABLE'3.3-11 (continued)
IABLE NOTATIOND ACTION 6 With the. number of.0PERABLE channels less than the' Minimum Number'of: channels shovn:-in Table 3i3.ll, restore theLinoperable channel (s) to OPERABLE. _.
status-within 7 days, or_be in_ 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 />.
~
ACTION 7 With the number..of OPERABLE ~ channels one11ess than the Required. Number.of Channels shown in Table 3.3-11, operation may proceed _until the next I_
CHANNEL CALIBRATION 1(which shall'be performed upon-the next entry into MODE 5, COLD SHUTDOVN),
ACTION 8 With the number of-OPERABLE channels one less than
. the Required;or Minimum number of channels-shown'in Table 13,3 11,eeither restore ~thel 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:
1.
Operation may proceed providedLthe-Required Channels'shown in Table-3,3-11 for the Reactor Coolant System Subcooling Margin Monitor and-the Core Exit'Thermocouples are' OPERABLE.
With the number of OPERABLE channels for.the; Reactor Coolant.Syreem Subcooling~ Margin Monitor and the-Core Exit Thermocouples. sh r n
.in Table.3.3-11 less.than the Required Numb 6:
.of Channels,. follow the associated action' statement, and-j L
2, Restore the~ system to 0PERABLE status at-the:
l next scheduled CHANNEL CALIBRATION (which l
shall be-performed upon the next entry;into MODE 5, COLD SHUTDOWN).
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i SALEM - UNIT 1 -
3/4 3-56a Amendment No. 117 9
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TABLE 4.3-11 (Continued 1 y
SURVEILLANCE REOUIRF.rir.=45 FOR X
ACCIDENT MONITORINC Iissimmar.wanTION i
a CHANNEL CMNNEL CHANNEL FUNCTIONAL y
INSTRUMENT CHECK CALIBRATION TEST
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12..PORV Position Indicator M
NA g
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- 13. PORV Block Valve Position Indicator M
MA g
14.' Pressurizer Safety Valve Position M
NA R
i Indicator
- 15. Containment Pressure - Narrow Range M
NA NA
- 16. Containment Pressure - Wide Range M
R NA
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- 17. Containment Water Level - Wide' Range M
R NA k
- 18. Core Exit Thermocouples M
R NA i
y-
- 19. Reactor Vessel Level Instrumentation M
R NA System (RVLIS) e e
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