ML20053C520
| ML20053C520 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 05/24/1982 |
| From: | Kane W Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8206020272 | |
| Download: ML20053C520 (29) | |
Text
_. _
MAY Z 4 N Docket No.:
50-395 APPLICANT: South Carolina Electric & Gas Company FACILITY:
Virgil C. Summer Nuclear Station
SUBJECT:
SUM"ARY OF HAY 11, 1982 APPEAL KETING ON TECHNICAL SPECIFICATIONS General Representatives of the HRC staff mt with the applicant to discuss the applicant's bases for appealing certain Technical Specifications. The meeting was held in Bethesda, Maryland. A list of those persons who attended the neeting is included in Enclosure %.
Purcose The purpose of the neeting was to assess the merits of the applicant's argumnts for the appeal.
Discussion The applicant appealed six Technical Specifications as described in their letter dated April 19, 1982. Each of these specifications is included in the Standard Technical Specifications for Westinghouse plants. A discussion of each of the appeal items follows:
1.
Remote Shutdown Instrumentation - T/S 3.3.3.5 The applicant's basis for tne change to this T/S and the proposed revised T/S are included in Enclosure R.
2.
Accident f!onitoring Instrumentation - T/S 3.3.3.6 The applicant's basis for the change to this T/S and the proposed revised T/S are included in Enclosure C.
3.
Steam Generator Insoection - T/S 3.4.5 The applicant's basis for the change to this T/S and the proposed revised T/S are included in Enclosure D.
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Soutt Carolina Electric & Gas MAY 2 4 N Co ncany l
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4.
Operational Leakane - T/S 3.4.6.2
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l The applicant appealed the T/S to limit reactor coolant leakage from I
each reactor coclant system pressure isolation valve to 1 gpm. The j
applicant contenoad that this requirement was not needed because the T/S already requires that the reactor coolant system leakage be l
limited to 1 gpn unidentified leakage and 10 gen of identified leakage l
and that although 1 gpa leakage through the reactor coolant systen l
pressure boundary may be indicative of an incending gross failure, I
leakage of 1 gpm across a valve seat certainly is not indicative of impending gross failure of the */alve. The applicant's proposed revised T/S in included in Enclosure E.
S.
Hot versus Cold Shutdown - T/S 3.7.3, 3.7.4, 3.7.5, 3.8.1.1, 3.8.2.1 and 3.8.3.1 The applicut's basis for the change to this T/S and the proposed revised T/S are included in Enclosure F.
6.
Steam Generator Tube Pluccinq ftarain - T/S 3.4.5 The applicant's basis for the change to this T/S and the proposed revised T/S are included in Enclosure G.
Following each of the applicant's presentations the staf f stated that the appeals involved Standard iechnical Specifications for Westinghouse plants. Because of this the staff was unable to decide the appeals at the meeting as might normally be the case, ye noted that NRR Office letters required that all changes to the Standard Technical Specifications be approved by the recently formed Comnittee to Review Generic Requirements. Ve noted that we would consider their arguments in determining whether changes to the Standard Technical Specifications should s
be made. However, we also noted, that this process would probably not be con-pleted in a time frame that would permit any changes to the applicant's Technical Snecifications orf or to operation.
Original signed t/:
Williec Kant W. F. Kane, Project fianager Licensing Branch No.1 Division of Licensing Encl osures: As stated cc w/encls.:
See next page
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Mr. T. C. Nichols, Jr.
Vice President & Group Executive Nuclear Operations South Carolina Electric & Gas Company Post Office Box 764 Columbia, South Carolina 29218 cc: Mr. Henry.Cyrus Senior Vice President South Carolina Public Service Authority 223 North Live Oak Drive i
Moncks Corner, South Carolina 29461 J. B. Knotts, Jr., Esq.
Debevoise & Liberman 1200 17th Street, N. W.
Washington, D. C.
20036 Mr. Mark B. Whitaker,.Jr.
I Group Manager - Nuclear Engineering & Licensing South Carolina Electric & Gas Company l
Post Office 764 Columbia, South Carolina 29218 p
1 Mr. Brett Allen Bursey Route 1, Box 93C Little Mountain, South Carolina 29076 i
i Resident Inspector / Summer NPS c/o V. S. NRC Route 1, Box 64 Jenkinsville, South Carolina 29065 Mr. James P. O'Reilly U. S. NRC, Region II I
101 Marietta Street 1
Suite 3100 Atlanta, Georgia 30303 8
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ENCLOSURE A LIST OF ATTENDEES NRC Staff South Carolina Elec. & Gas Co.
W. Kane D. Skovholt
-0. Dixon-T. Speis W. Williams T. Dunning 0..Bradham F. Rosa M.. Whi taker J. Wermiel M. Browne
- 0. Parr M. Quinton L. Rubenstein A. Koon J. Knight R. Clary F. Cherny i
W. Johnston I
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ENCLOSURE B REMOTE SHUTDOWN INSTRUMENTATION SPECIFICATION 3.3.3.5 APPLICABILITY - MODES 1, 2, AND 3 i
ACTION:
REQUIRES SHUTDOWN OF PLANT ON LOSS OF ANY ONE CHANNEL FOR 7 DAYS.
BASIS FOR CONCERN:
1.
LOSS OF ONE PARAMETER AT PANEL WOULD NOT IN ITSELF PREVENT SAFE SHUTDOWN OF THE Ut4IT FROM OUTSIDE OF CONTROL ROOM.
2.
OTHER MEANS CAN BE PROVIDED FOR MEASURING MOST PARAMETERS ON A TEMPORARY BASIS:
(1) TEMPERATURE SIGNALS ARE RESISTANCE TYPE, AVAILABLE AT PANEL LOCATION.
(2) ALL OTHER SIGNALS ARE 4-20MA, WITH SIGNAL i
CONDITIONING DONE AT PANEL LOCATION.
ll 3.
AVAILABILITY OF REPLACEMENT COMPONENTS MAY PRESENT l
SERIOUS PROBLEMS DUE TO LONG LEAD TIMES FOR DELIVERY.
4.
SYSTEM WAS NOT DESIGNED NOR INSTALLED FOR SINGLE FAILURE CRITERIA OR TO MEET THIS TYPE RESTRICTIVE TECHNICAL SPECIFICATION.
l S.
REQUEST FOR WAIVER OR TEMPORARY RELIEF FROM TECHNICAL l
SPECIFICATIONS SHOULD BE EXCEPTIONAL CASES NOT NORMAL s-
5.
(CONTINUED)
CONDUCT OF BUSINESS.
6.
LAST, BUT NOT LEAST BY ANY MEANS, IS THE REPLACEMENT ENERGY COST.
ESTIMATED COST OF OUR 900 MW UNIT IS f
$600,000 TO $1,000,000 PER DAY.
PROPOSED ALTERNATIVES:
1.
REVISE WORDING 0F ACTION STATEMENT TO ALLOW PROVISIONS 1
TO EVALUATE IMPACT ON SAFE OPERATIONS AND PROVIDE A MEANS TO READ THE IN0PERABLE PARAMETER TO THE SAME I
DEGREE OF ACCURACY AS PROVIDED BY THE INSTALLED' INSTRUMENTATION ON A TEMPORARY BASIS. PROVIDE SPECIAL REPORT PER SPECIFICATION 6.9.2.
2.
OR, AT A MINIMUM, PROVIDE A MEANS IN THE ACTION STATE-MENT. T0 OBTAIN CONCURRENCE BY NRC 0F THE METHOD TO BE EMPLOYED TO RESTORE THE FAILED CHANNEL OR AN ACCEPTABLE ALTERNATE MEANS TO READ THE PARAMETER. THIS WOULD ESTABLISH A RESOLUTION MECHANISM OTHER THAN A FORMAL REQUEST FOR TECHNICAL SPECIFICATION WAIVER.
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REPOTE SHUTDCWN INSTRUMENTATICN I
LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote snutcown monitoring instrumentation channels snown in Tacle 3.3-9 shall be CPERASLE with reacouts dis;2 layed external to the control rcom.
APPLICA3ILITY: MODES 1, 2 and 3.
ACTICN:
a.
With the numoer of CPERABLE remote shutdown monitoring channels less than required by Tacle 3.3-9,'=ette-the-4neperaal-eeennel-te OP ERAB LE -5tatus-M-thi n 4 ay 5 r -o"- -be- -in-FOT-S WUTCChi-0W,- th e - m ex t 4-hours.
b.
The previsicns of Specification 3.0.4 are not a::plicable.
- C'itltin 7 daljs citite.1 resto.te tJte incpetable eltannells) to ope, table status ct ytcvide atitet means cf. measu. ting tJte pa.tameterts) and subnit a Special Report to t!1e Corni.ssicr. :~'tsuant to Specifica.C.icn 6.9.2 cu,tLining tite plans fct restoring tlte incyc, table inststtrent(s) to cpetable sta,tus.
Ct.11enaise be in HOT SHUTDCC:.4 teLtitin tl1e next 12 (1 curs.
SURVEILLANCE RECUIREMENTS 4.3.3.5 Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK anc CHANNEL CALIBRATICN coerations at the frequencies shcwn in Tacle 4.3-5.
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SUMMER - UNIT 1 3/a 3-53
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ENCLOSURE C POST ACCIDENT MONITORING INSTRUMENTATICN SPECIFICATION 3.3.3.6 APPLICABILITY - MODES 1, 2, AND 3 ACTION:
REQUIRES SHUTDOWN OF PLANT ON LOSS OF ONE CHANNEL FOR 7 DAYS OR LOSS OF BOTH CHANNELS FOR 48 HOURS.
BASIS FOR CONCERN:
1.
LOSS OF ONE DISPLAY IN THE CONTROL ROOM WOULD NOT IMPAIR OUR ABILITY TO SAFELY MONITOR PLANT CONDITIONS DURING AN ACCIDENT DUE TO REDUNDANT INFORMATION BEING AVAILABLE IN THE CONTROL ROOM AREA.
2.
CREDIT SHOULD BE ALLOWED TEMPORARILY FOR THOSE PARAMETERS' DISPLAYED IN THE CONTROL ROOM AND TECHNICAL SUPPORT CENTER (TSC).
3.
QUALIFIED REPLACEMENT COMPONENTS MAY PRESENT SERIOUS PROBLEMS DUE TO LONG LEAD TIMES FOR DELIVERY.
4 4.
SINCE THERE ARE A LARGE NUMBER (50) 0F CHANNELS INVOLVED, THIS POSES GREAT POTENTIAL FOR FORCED OUTAGE DUE TO SINGLE COMPONENT FAILURE.
PROPOSED ALTERNATIVES:
I 1.
REVISE ACTION STATEMENT "a" TO ALLOW CREDIT FOR AT LEAST ONE CHANNEL OR COMPONENT DISPLAYED IN CONTROL
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PROPOSED AL,TERNATIVES 1.
(CONTINUED)
AREA. ATTACHED TABLE SHOWS EXAMPLES OF REDUNDANCY AVAILABLE.
(1) QUALIFIED RECORDERS AND INDICATORS ON SAME CHANNEL. LCSS OF ONE SHOULD NOT REQUIRE PLANT SHUTDOWN.
(2) HARD-WIRED INDICATION AVAILABLE IN TSC FOR BASIC PARAMETERS ON GRAPHIC PANEL. SAME SIGNAL THAT IS PROVIDED TO CONTROL ROOM EVACUATION PANEL.
(3) TSC COMPUTER READ 0UT IN CONTROL ROOM AND TSC.
(4) PLANT PROCESS COMPUTER READ 0UT.
(5) SUBC00 LING MONITORING CABINET IN CONTROL ROOM.
(6) LOCAL INDICATIONS ARE PROVIDED FOR KEY PARAMETERS NEAR EMERGENCY FEEDWATER VALVES.
2.
REVISE ACTION STATEMENT "b" TO ALLOW EVALUATION OF IMPACT ON ACCIDENT MONITORING CAPABILITY, UPON LOSS OF 2 CHANNELS WITH NRC CONCURRENCE OF METHOD TO PROVIDE ALTERNATE MEANS AND SUBMITTAL OF WRITTEN REPORT PER SPEC. 6.9.2.
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REMOTE SHUTOOWN AND ACCIDENT MONITORING INSTRUMENTATION (KEY PARAMETERS)
TSC 2500 SENSOR INDICATION RECORDER COMPUTER PLANT COMPUTER L
RC wl de 0
TE413 Tid 13 (PAMS)
TR413 (PAMS)
YES rES Range (Tgot) 0 Sub Cooling Panel
- 1 P
TE413A Ti4138 (TSC) 1 Tl413A (CREP)
L 0
TE423 Tl423 (PAMS)
TR413 (PAMS)
YES YES 0
Sub Ceoling Panel
- 2 P
TE4 23A Tl4238 (TSC) 2 Ti423A (CREP)
L 0
TE433 Sub Cooling Panel TR413 (PAMS)
YES YES 0
- 3 P
TE433A Tl433B (TSC) 3 T1433A (CREP)
L RC dide Range 0
TE410A Tl410 (PAMS)
TR410 (PAMS)
YES YES (T
0 Sub Cooling Panel
- 1 cold P
T E410 A T l 4108 (TSC) 1 T1410A (CREP)
L 0
TE420 Tl420 (PAMS)
TR410 (PAMS)
YES YES 0
Sub Cooling Panel
- 2 P
TE420A T14208 (TSC) 2 Tl420A (CREP)
L 0
TE430 Sub Cooling Panel TR410 (PAMS)
YES YES O
- 3 P
TE430A T 14 308 (TSC) 3 Tl430A (CREP)
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e REMOTE SHUT 00WN AND ACCIDENT MONITORING INSTRUMENTATION (KEY PARAMETERS)
TSC 2500 SENSOR INotCATION RECORDER COMPUTER PLANT COMPUTER S
SG Wide G
LT477 Ll477 (PAMS)
FR478 YES YES Range Level Ll477A (LOCAL)
- 3 A
LT477A Ll477C (TSC)
Ll477B (CREP)
S G
LT487 Ll487 (PAMS)
FR488 YES YES Ll487A (LOCAL)
B LT487A Ll487C (TSC)
Ll4878 (CREP)
S G
LT497 Ll497 (PAMS)
FR498 YES YES tl497A ( LOC AL)
C LT497A Ll497C (TSC)
Ll4978 (CAEP)
LT459 Ll4 59 A (PAMS)
LR459 YES YES PZR Water Level Ll459C (TSC)
LT459A Ll4598 (CREP)
Ll4590 (TSC)
LT460 Ll460 (PAMS)
LR459 YES YES.
LT461 Ll461 (PAMS)
LR459 YES YES RC Wi de Range PT402 Pl402 (PAMS)
PR402 (P AMS?
YES YES Pressure Sub Cooling Panel
- 1 PIA 028 (MCB)
PT402A Pl402C (TSC)
Pl4020 (CREP)
PT403 Pl403 PR402 (PAMS)
YES YES Sub Cooling Panel
- 2 P T4 0 3 A Pl4038 (TSC)
Pl403A (CREP)
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REMOTE SHUTDOWN AND ACCfDENT MONITORING INSTRUMENTATION (KEY PARAMETERS)
TSC 2500 SENSOR INDICATION RECORDER COMPUTER PLANT COMPUTER EF Flow W
FT3561 F43561 (PAMS)
YES YES I
FT3571 F13571 (PAMS)
YES
'YES D
FT3581 F13581 (PAMS)
YES YES E
Fl3508 (LOCAL)
FT3561A Fl35618 (PAMS)
NO NO N
Fl3561A (TSC)
A R
FT3571A Fl35718 (PAMS)
NO NO R
Fl3571A (TSC) 0 W
FT3581A Fl35819 (PAMS)
NO NO F13581A (TSC)
Steam Li ne S
PT474 Pl474 (PAMS)
YES YES Pressure G
PT475 Pl475 (PAMS)
PR475 (PAMS)
NO YES PT476 Pl476 (PAMS)
NO YES A
PT2000 P12000 (MCB)
NO NO P12000A (CREP)
Pl20000 (TSC)
S PT484 Pl484 (PAMS)
YES YES G
PT485 PI485 (PAMS)
PR485 (PAMS)
NO YES PT486 Pl486 (PAMS)
NO YES B
DT2010 P120!O (MCB)
NO NO Pl2010A (CREP)
Pt20109 (TSC)
YES YES S
PT494 P1494 (PAMS)
G PT495 P1495 (PAMS)
PR495 (PAMS)
YES PT496 Pl496 (PAMS)
YES C
PT2020 Pl2020 (MCB)
P12020A (CREP)
Pl2020B (TSC)
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1 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATICN l
LIMITING CCNDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Tacle 3.3-10 shall be CPERABLE.
APPLICABILITY: MCOES 1, 2 and 3.
ACTION:
a.
With the number of OPERABLE accident monitoring entanels less than tne Required Numcer of Channels snown in Tacle ?.J-10,'restere-the-inoperaole-enannel(+)--to-GatBLE-stuus-WtMn-7-eays3 sc-be 4a a t - l e a s t -H07 -SMTCGhi-tah -0.4 4en, #, curs.
b.
With the number of OPERATsLE accident monitoring enannels less than the Minimum Channels OPERABLE requirements of Tacle 3.3-10;"either restere-tne-inoperat>k-ohannel+-to-GW4L-E-4tatus-Wtin 23-hcurs-cr be-&n-at-least-HGT-SMTCC' N-+4thb-the-4e.c 42 tms, c.
The provisions of Seccification 3.0.4 are not acclicaole.
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- Wititin 7 days eithet.testo.te. tite incpetable channel to opeuble sLttus c.:
p.tovide o.titet means of mea. sating the pttaretet(s) and submit a Special Repo.rt to the Comiss.icn pu,tsuant to Specifica.ticn 6.9.2 catiining the plans fc.t tes.co. ting tite incpeuble instturertt(s) to opetable status. C.thenvise be in HOT SHUTOCWN testitin the next 12 incu.ts.
SURVEILLANCE REOUIREMENTS 4.3.3.6 Eacn accident monitoring instrumentation enannel shall be demonstrated GPERABLE by performanca of tne CHANNEL CHECX and CHANNEL CALI3RATICN ocerations at the frequencies shown in Tamle 4.3-7.
'Wititin 18 heats ei.titet testote the incp st:tble channel to opetabic sLttas c.
p.tovide othet ecans of measating the pa.tanetet(s) and submit a Special Repcrt tu the Ccer.lssion pursuant ta 3peci{icatica 6.9.2 cattbting the plans fe.1
.testcring the inopetable insttunent(s) tc
'etable sbttus. Othencise be in HOT SKUTCCWN tei. thin the acx.t 12 heu.ts.
5tMNR - UNIT 1 3/t 3-56 A
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ENCLOSURE D PROBLEMS WITH CURRENT: SPECIFICATION 1.
L6CAISNOTDEFINED.
2.
THis SPECIFICATION WILL RESULT IN:
A.
ADDITIONAL PLANT COOLDOWNS B.
ADDITIONAL MAINTENANCE
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C.
ADDITIONAL RADIATION EXPOSURE TO WORKERS D.
INORDINANT PLANT DOWNT!!1E E.
FINANCIAL BURDEN i
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' JUSTIFICATION FOR PROPOSED CHAfiGE 1.
WHEN REACTOR COOLANT PRESSURE IS MAINTAINED ABOVE SECONDARY SIDE PRESSURE BY THE SAFETY INJECTION PUMPS, THE STEAM GENERATOR TUBES ARE NOT SUBJECTED TO UNACCEPTABLE LOADING 2.
WITH REGARD TO LOCA, SCE&G STEAM GENERATOR TUBES ARE ANALYZED FOR LOCA WITH A SAFE SHUT-DOWNEARTHbuAKEINADEGRADEDCONDITION f
l (WCAP-9912 " STEAR'. GENERATOR TUBE PLUGGING Jj' MARGIN ANALYSIS FOR THE VIRGIL C. SUMMER NUCLEAR PCWER PLANT llNIT 1.") AND WOULD NOT BE ADVERSELY 1
AFFECTED.
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REACTOR CCOLANT SYSTEM I
C'OF 0. ]E*U'T DU.
SURVEILLANCE REOUIREMENTS (Continued)
' ' ' ' Ub8 I 4
4.4.5.3 Insoection Frecuencies - The above required inservice inspections of steam gener: ct tucc: sncil ce performed at the.folicwing trequencies:
The first inservice inspection shall be performed after 6'E?fective a.
Full Power Months but within 24 calendar months of initial criticality.
Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after the previous inspection.
If two consecutive inspections felicwing service under.
AVT conditions, not including the preservice inspection, result in _-
all inspection results falling into. the C-1 category or if two consecutive inspections demonstrate that previously observed degra-dation has not continued and no additional degradation has occurred.
the inspectica interval may be extended to a maximum cf once per 40 months.
b.
If the results of the inservice inspection of a steam generatcr
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conducted in accordance with Table 4.4-2 at 40 month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months.
The increase in inspection frequency shall apply until'the subsequent inspections satisfy the criteria of Specification 4.4.5.3.a; the interval may then be extended to a naximum of once per 40 months.
Additicnal, unscheculed inservice inspections shall be performed en c.
each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the felicwing concitions:
1.
Primary-to-secondary tubes leaks (not including leaks originating from tuce-to-tube sheet weles) in excess of the limits of Specification 3.4.6.2.
2.
A seismic occurrence greater than the Operating Basis Earthquake.
3.
A icss-of coolant accident e:;.
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u ys gre. Hoi shls +c m>=rais re aero-coolawr e c op ey sr c<8 se< >vre.
sysren pr<crues sL,,,e 4.
.ne er reecwater line break.
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SUMNER - UNIT 1 3/: 4-12
ENCLOSURE E REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System Leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE, b.
1 GPM UNIDENTIFIED LEAKAGE, 1 GPM total primary-to-secondary leakage through all c.
steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System, d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant l
System, and 33 GPM CONTROLLED LEAKAGE AT a Reactor Coolant System e.
pressure of 2235 1 20 psig.
i f.
Reactor Coolant System Pressure Isolation Valves shall
't be OPERABLE per 4.4.6.2.2 below.
APPLICABILITY:
MODES 1, 2, 3 and 4 ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least 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 />, b.
With any Reactor Coolant System leakage greater then.
any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 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 />.
c.
With any Reactor Coolant System Pressure Isolation Valve not OPERABLE per 4.4.6.2.2, provide at least one additional closed manual isolation or deactivated automatic valve between the RCS and low pressure portions of the system for each Reactor Coolant System Pressure Isolation Valve not operable or reduce RCS pressure to 1 600 psig within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
Monitoring the Reactor Building atmosphere (gaseous a.
or particulate) radioactivity monitor 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.
Monitoring the Reactor Building su=p inventory at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
c.
Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 2235 20 psig at least once per 31 days with the modulating valve fully open.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.
d.
Performance of a Reactor Coolant System water inventory balance at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Monitoring the reactor head flange leakcff system e.
at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall be demonstraced OPERABLE pursuant to Specification 4.0.5, except that in lieu of any leakage testing required by 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be less than the maximum allowable value of 10 GPM.*
a.
Every refueling outage during startup.
b.
Prior to returning the valve to service followng maintenance repair or replacement work on the valve.
c.
Prior to entering MODE 2 following valve actuation due to automatic or manual action or flow through the valve for valves denoted on Table 3.4-1 by an asterisk.
d.
The provisions of Specification 4.0.l. are not applicable for entry into MODE 3 or 4.
Leakage rate shall be considered unacceptable if the latest measured rate exceeds the rate determined in the last previous test by an amount that reduces the margin between the current measured leakage rate and the maximum permissible race by 50%
or greater.
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TABLE 3 4-1 Reactor Coolant System Pressure Isolation Valves Valve No.
Description Maximum Allowable Leakage (' ) (2) 8993A SI to Hot Legs Loop A 10 GPM 89933 SI to Hot Leg Loop B 10 GPM 8993C SI to Hot Leg Loop C 10 GPM 8988A SI Low Head to Ho Leg A 10 GPM 8988B SI Low Head to Hot Leg B 10 GPM e998A SI to Cold Leg Loop A 19 GP:t 8998B SI to Cold Leg Loop B 10 3PM e998C SI to Cold Leg Loop C 10 GPM 8973A RER Low Head to Loop A 10 GPM 8973B RHR Low Head to Loop B 10 GPM 8973C RER Low Head to Loop C 10 GPM 8948A+
Accumulator to Loop A 10 GPM 8948B+
Accumulator to Loop 3 10 GPM 8948C+
Accumulator to Loop C 10 GPM 8956A+
Accumulator to Loop A 10 GPM 8956B+
Accumulator to Loop B 10 GPM 8956C+
Accumulator to Loop C 10 GPy (1)
Leakage rate shall be considered unacceptable if the latest measured rate exceeds the rate determined in the last previous test by an amount that reduces the margin between the current mesaured leakage rate and the maximum permissible rate by 50% or greater.
(2)
Leakage measurements shall be conducted persuant to 4.0.5 The minimum differential pressure shall be > 1000 psid and the leakage rate shall be corrected to 2235 psid.
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ENCLOSURE F G
HOT VERSUS COLD SHUTDOWN SPECIFICATION 3.7.3: COMPONE!iT COOLING WATER SYSTEM SPECIFICATION 3.7.4: SERVICE WATER SYSTEM SPECIFICATION 3.8.1.1:
AC POWER SOURCES SPECIFICATION 3.8.1.2:
DC POWER SOURCES SPCCIFICATION 3.8.3.1:
ONSITE POWER DISTRIBUTION APPLICABILITY - MODES 1, 2, 3, A!1D 4 ACTION:
REQUIRES BRINGING PLAtlT TO COLD SHUTDOWN UPON LOSS OF ONE LOOP OR TRAlti 0F ABOVE SYSTEMS.
BASIS FOR CONCERN 1.
THESESYSTEMSAREREQUIREDTOXEEPTHEPLAtlTIN"QLD SHUTDOWN."
2.
PRESENT SPECIFICATIONS FORCE THE PLANT INTO A LESS CONSERVATIVE CONDITION DUE TO LOSS OF REDUNDANCY OF REQUIRED SYSTEMS TO MAINTAIN COLD SHUTDOWN CONDITIONS.
3.
INCREASES POTENTIAL FOR PLANT TRANSIENT UPON LOSS OF REMAIflING TRAIN.
(l') TRANSITION FROM " COLD SHUTDOWN" T0 " HOT SHUTDOWN" WITHOUT DECAY REMOVAL CAPABILITY PRESENTS A SIGNIFICANT CHALLENGE TO THE OPERATOR WITH LITTLE OR fl0 P RGIN FOR ERROR.
4.
HEAT-UP RATE FROM COLD SHUTDOWN TO SATURATION AtlD FROM HOT SHUTDOWN TO 5,TURATION IS ESSENTIALLY THE SAME, ASSUMING TOTAL LOSS OF DECAY HEAT REMOVAL CAPABILITY.
5.
THE CAPABILITY TO REMOVE DECAY HEAT IS ABSOLUTELY ESENTIAL IN MITIGATING CORE DAMAGE.
6.
PANY MORE DIVERSE MEANS ARE AVAILABLE TO REMOVE DECAY HEAT WHILE PLANT IS IN "H0T SHUTDOWN" CONDITION.
(1) 5 POSSIBLE PATHS AVAILABLE FOR " HOT SHUTDOWN."
(2) ONLY 2 POSSIBLE PATHS AVAILABLE FOR " COLD SHUTD0Wil. "
(3)
"H0T SHUTDOWN" PROVIDES ACCESSIBLE ALTERNATE PATHS AVAILABLE OUTSIDE CONTAltlMENT.
7.
THE P01ENTIAL FOR LOSS OF REACTCR COOLANT IflVENTORY IS DECREASED WITH PLANT IN " HOT SHUTDOWN CONDITION."
8.
THE BOTTOM LINE IN CONTROLLING ANY ACCIDENT IS:
(1)
ISOLATE LEAXAGE PATHS TO ENVIRONMENT.
(2) CAPABILITY TO REMOVE DECAY HEAT.
(3) PAINTAIN AS MUCH SAFETY MARGIN AS POSSIBLE TO PROTECT ITEM 1 AND 2.
9.
" COLD SHUTDOWN" SOUNDS SAFER THAN " HOT SHUTDOWN" BUT IN OUR OPINION "H0T SHUTDOWN" PROVIDES CONSIDERABLY MORE j
SAFETY PARGIfl.
PROPOSED ALTERhATE:
1.
REVISE SPECIFICATIONS TO DELETE APPLICABILITY TO MODE 4.
2.
REVISE SPECIFICATIONS TO SUBSTITUTE HOT SHUTDOWN FOR COLD SHUTDOWN.
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[k{ C',p, n PLANT SYSTEMS u..:.,; 4', UUv'y nn.
3/4.7.3 COMPCNENT CCOLING WATER SYSTEM s
LIMITING CCNDITICN FCR CPERATION
- 3. 7,. 3 At.least two. independe.nt c;mponent cooling water, Icops sha.1] be.,0PERABLE.
APPLICABILITY: MODE 3.1,,2, 3, andas.,
.s ACTICN:
With only one component cooling water loco OPERABLE, restore at least two locos 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 STAN08Y 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 C3C3 SHUTDC'nN within the following 30 hcurs.
EOT SURVEILLANCE REOUIREMENTS
'4.7.3 At. least two compenent cool-i'ng water loops ~shall ce demonstrated 0PERh8tE:' :' '
a.
At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related ecuipment that is not l'ocked,' sealed, or otherwise secured in p'osition, is in its correct position.
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SLPMER - UNIT 1 3/4 7-11
ENCLOSURE G
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HISTORY OF EVENTS 1.
NRC ISSUES REGULATORY Gu1DE 1.121 FOR COMMENT AUGUST 1976.
2.
WESTINGHOUSE PR0vIDES COMMENT NOVEMBER 22, 1976.
e 3.
WESTINGHOUSE ISSUES RESAR 414 DECEMBER 31, 1976.
4.
NRC ISSUES NUREG 0491 NOVEMBER 14, 1978 SAFETY EVALUATION REPORT FOR RESAR 414.
5.
SCE8G CONTRACTS WESTINGHOUSE TO PERFORM STEAM GENERATOR. PLUGG ANALYSIS DECEMBER 22, 1980.
6.
SCESG SueMITS WCAP 9912 REVISION 1 TO NRC N09 EMBER 30, 1981.
7.
NRC CALLS SCE&G APRIL 1982 TO DISCUSS WCAP 9912.
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TABLE 1-6 (Continued) l RESAR-414 COMPLIANCE WITH REGULATORY GUIDES, INCLUDlhG DECISIONS Or inc REGULAIGaf REQUIREMENTS
~
RtVitw CGMtiliite Regulaterv Culdes (RG)
Applicable Westinghouse Position in g
Revision RESAR-414 Accendix 3A Staff Discositfori' r
1.106 Outside scope 1.107 Outside scope 1.108 Outside scope 1.109 Outside scope b
1.110 Outside scope 1.111 Outside scope 1.112 Outside scope 1.113 Outside secpe 1.114 Outside scope -.
1.115 Outside scoce 1.116 Outside scope 1.117 1
Outside scope 1.118 1
Met without exception our review will be included with the verification program of the integrated pro-tection system dis-
- cussed in Section 1.4 and h l.2.5 of this report.,
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_ 1.119 withdrawn i
1.120 0
Outside sco:e. Interf aces specified. ' Acceptable inter-faces specified as discussed in Section 9.5 of this report.
I_
f 1.121 0
. (1) westingnouse intends to use a (1) The Westinghouse factor of two for margin to position is A j
steain generator tube burst acceptable as j
pressure for norr.a1 cperation, clarified in Instead of a factor of three.
Section 5.4.2.2 of this report.
(2) Vesting 1ouse will use a leak rate (2) Accepted associa ed with the crack size determined on the basis of acciden: loadings instead of
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normal oceration oressure leads.
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1-34 q*
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TABtf 1-6 (Continued)
RESAR-414 COMPLIANCE WITH REGULATORY CUIDES.
[
INCWDIhG DtC1510tg 0F int _stGulAiGRT Rt0UIREMENTS stVIEW CDP.911ste Regulatory Guides (RG)
Applicable Westinghouse Position in RG Revision RESAR-414 Aeoendix 3A Staff Diseosition (3) Westinghouse will establish a (3) Commitment is minimum acceptable tube wall
.is acceptable.
L thickness (plugging limit) in sccordance with the general intent of the guide.
(4) Two separate tube plugging limits (4) Approach is may be established to address the acceptable.
U-bend area and the straight length of tubes to eliminate unnecessary plugging of tubes.
1.122 1
Outside sch-M 1.123 1
t Westinghouse alternates and Accepted by our clarifications are listed in letter of WCAP-8370, Revision EA.
September 1977.
1.124 1
Westinghouse does not agree with Our positions are paragraphs C.4, C.6(a), C.7(b) nor given in Section C.8 of this guide.
5.2.1 of this report.
1.125 Outside scoce 1.126 0
Vestinghouse's method of fuel Previously densification analysis is based accepted as upon WCAP-8218-PA.
discussed in Section 4.2.1.2 1.127 of this report.
Outside scope
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1.128 Outside scoce
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1.130 0-Westinghouse does not agree with Our positions paragraphs B.), C.3, C.4(a), C.6(a),
are given in nor C.7 of this guide.
Section 5.2.1 of this report.
1.131 Outside scope 1.132 Outside scope 1.133 Outside sco.:-
1.134 Outside scope 1.135 Outside scope 1.136 Not addressad.
Outside scope O
s 1-35
hole design promotes high velocity flow along the tube away from the support plate locations.
, sweeping impurities The quatrefoil tube succort design along with other design changes result i reduced fluid pressure drop across the support plates and therefore a higher na circulation ratio and higher fluid velocities in the tube bundle.
the increased circulation through the tube bundle, an additional set To accommodate tion bars have been Installed in the U-bend region cf the tubes of antivibra-additional strength and resistance to vibration, the quatrefoil. tube supp To provide thickness has been increased compared to previous"Vestinghouse support e
designs.
plate Additional peripheral supports have also been installed to provide stability to the plates so that tube fretting or wear due to flow induced plate vibration at the tube support contact regions will be abated Flow distribution baffles are also provided in this model which produce higher cross flo#v l luediately above the tube sheet and sweep the sludge to the' center of the t b e ocites bundle where the blowdown connection is located.
ue Assurance that damaging flow induced tube vibration will not occur will b through a combination of analyses and testing.
e provided Also, Westinghouse has committed to provide a topical report for our final design review of RESAR-414 justf i that the design bases for the Model H steam generator tubes are !n acc y ng Regulatory Guide 1.121 " Bases for Plugginn Degraced PVR Steam Generator T ordance with with an exception that Westinghouse is proposing for the tube plugging b position stated in Regulatory Guide 1.121 for minimum acceptable ttbe wall asis: Our ness states that tuces with part through wall cracks, wastage, or combinations of these should have a factor of safety against failure by bursting under operating cond' itions of not less than three at any tube location.
normal Westinghcuse proposes to use a factor of two.
information clarifying tnis factor by Amendment 17 to RESAR-414At our Our evaluation of the additional information indicates that the variar.ce in-the positions taken by Westinghouse and that provided in the guide results from Westing' house's cefinitica of tube failure, which is different from that intended in the guide Westinghouse defines tube failure as plastic deformation of a crack to the e t that the sides of the cract open to a non par'allel elliptical configuratio x ent tubing can sustain aeded internal prassure beyond those values before reachi n.
The condition of gross failure.
ng a 1
three against gross failure by bursting. Regulatory Guide 1.121 requires a facto Ve find that the proposed factor of s.afety of two eased uson Vestinghouse_'s definition of failure provides for a,
J factor of safety aednst burst failure which is consistent with that recom nd d by the guide.
e e Therefore, we fina ther Vestinghouse position to te acceptable.{(
Interface information in RESAR-414 for water chemistr cc : trol fo of the ' steam generator requires the use of v' latile ammonium hydroxide or a o
volatile amine and prohibits the use of sodium phosphate.
Westinghouse has used model boiler tests to determine the Joequacy of their all volatile-treatment 5-26 e
=.
I JUSTIFICATIONS 1.
2.
WCAP 9912 REVISION 1 3.
CONSERVATISM IN ANNLYSIS A)
PRIMARY TO SECONDARY DIFFERENTIAL PRESSURE 2650 AcSuMED; 2585 MAX.
.l B)
.76GPMChLCULATED;
.347 GPM ALLOWED.
C)
CHEMICAL AND PHYSICAL PROPERTIES.
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E MEETING
SUMMARY
Document Control (50-395)
NRC PDR MM 2 41932-L PDR TERA NSIC LS#1 Rdg.
MRushbrook Project Manager W. F. Kane Attorney, OELD OIE Regional Administrator, Region II PARTICIPANTS (NRC):
W. Kane D. Skovholt T. Speis T. Dunnirig F. Rosa J. Wermiel
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L. Rubenstein 6
J. Knight p
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F. Cherny J
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W. Johnston 9
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