ML20062D652
| ML20062D652 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 07/30/1982 |
| From: | Dixon O SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20062D656 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8208060183 | |
| Download: ML20062D652 (16) | |
Text
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l SOUTH CAROLINA ELECTRIC & GAS COMPANY l
POST OFFICE 764 COLUMetA. SOUTH CAROUNA 29218
- o. w. oixos. JR.
July 30,1982 VICE PRESIDENT NUCLEAR OpEmatioNS Mr. Harold R. Dentcn, Director Office of n2 clear Reactor Regulation U. S. Nuclear Regulatory Ccmmission Washington, DC 20555
Subject:
Virgil C. Stamer Nuclear Station Docket No. 50/395 Safety and Relief Valve Report; NURED 0737 Item II.D.1
Dear Mr. Dentcxt:
In response to NUREG 0737, Item II.D.1, South Carolina Electric & Gas 02npany (SCE&G) has participated in the Electric Power Research Institute (EPRI)
Safety and Relief Valve test program to demcastrate the operability of the pressurizer Safety Valves, Power Operated Relief Valves (PORV's) ard the PORV Block Valves and the adequacy of the piping and supports associated with these conponents.
The included attachments will address each area as follows:
ATTACINENT 1 - Safety Valve Perfonnance Evaluation ATTAONENr 2 - Power Operated Relief Valve Perfonnance Evaluation ATTACHMENT 3 - PORV Block Valve Performance Evaluation ATTACINENT 4 - Pressurizer Relief System Pipirg and Support Evaluation The basic conclusion is that the valves have been destonstrated to perform their intended function as described in the subject NUREE and the piping and supports are adequate for design loads during valve operation. Any variations in valve performance or test conditions are discussed in the identified attachnents.
The final design verification of the analysis contained in Attachment 4 is in progress and is expected to be ccmplete by mid-August. We do not anticipate any change in the results of the analysis or the conclusions based on this analysis.
This sutmittal in conjunction with the April 1,1982 subnittal constitute a final report on NURED 0737, Iten II.D.l.
82080601EG 820730 J
PDR ADOCK 05000395 1
A PDR
Mr. Harold R. Denton July 30, 1982 Page 2 If you have any questions or require additional information, please advise.
Very trul yours, w
- 0. W.
ixon, J.
MDO:OWD:glb Attachnents V. C. Stmner w/o atts.
cc:
G. H. Fischer w/o atts.
H. N. Cyrus T. C. Nichols, Jr. w/o atts.
O. W. Dixon, Jr.
M. B. Whitaker, Jr.
J. P. O'Reilly H. T. Babb D. A. Natanan C. L. Ligon (NSRC)
W. A. Williams, Jr.
R. B. Clary O. S. Bradham A. R. Koon M. N. Browne G. J. Braddick J. L. Skolds J. B. Knotts, Jr.
B. A. Bursey NPCF File
Ietter Fran O. W. Dixcm, Jr., SC%G, To H. R. Denton, Dated July 30, 1982 ATTACINENP 1 l
Safety Valve Performance Evaluation July, 1982 Page 1 of 3
References
References:
a) Letter fran T. C. Nichols, Jr., South Carolina Electric and Gas Conpany, to H. R. Denton, dated April 1,1982, with attachments.
b) WCAP 10105, " Review of Pressurizer Safety Valve Performance As Observed In %e EPRI Safety and Relief Valve Test Program," June,1982.
Reference (a) and the attendent EPRI Reports docunented the Safety Valves installed at the V. C. Stamer Nuclear Station (VCS) and that the Safety Valves tested during the EPRI PWR Safety and Relief Valve %st Program represent the safety valves installed at VCS. We conditions for which the valves were tested envelope the range of expected operating ard accident conditions for VCS. Furthermore, the testing demonstrated the functionability of the safety valves while identifying scme ancmalies in valve performance.
Westinghouse, through the Westinghouse Owners Group, has evaluated the observed safety valve performance during full scale testing and subnitted reference (b). %e discussion regarding upstream safety valve piping pressure oscillations is not applicable to VCS.
VCS utilizes a hot loop seal whose tenperature exceeds that addressed in i
reference (b) with regard to pressure oscillations in the upstream piping.
Wese pressure oscillations are addressed in Attachment 4 of this subnittal l
and have been found to be acceptable.
1 In addition to reduced loads on the downstream piping, the hot loop seal also improves valve performance as shown in the 350*F loop seal tests.
Crosby Valve and Gage O3., the safety valve vendor, has evaluated the safety valve performance against VCS plant specific piping and valve information and determined that valve performance at VCS should be as good or better than observed valve performance during the EPRI test program.
Page 2 of 3
Valve accelerations and nozzle loads are discussed in Attachment 4 of this subnittal.
Based on the test p ogram output, the evaluations of reference (b), the valve vendor evaluation and the plant specific piping and support evaluation, the operability of the VCS Safety Valves has been demonstrated in accordance with the requirements of NURE 0737.
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Letter Fran O. W. Dixon, Jr., SCE1G, 'Ib H. R. Denton, Dated July 30, 1982 i
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Power Operated Relief Valve Performance Evaluation l
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Peferences a) Ietter from T. C. Nichols, Jr., South Carolina Electric and Gas Chupany, to H. R. Denton, dated April 1,1982, with attachnents.
Reference (a) and the attendent EPRI Reports doctanented the Power Operated Relief Valves (PORV's) installed at the V. C. Stmner Nuclear Station (VCS) and that the PORV's tested during the EPRI PWR Safety and Relief Valve Test Program represent the PORV's installed at VCS. The conditions for which the valves were tested envelope the range of expected operating and accident conditions for VCS. Furthennore, the testing demonstrated.the operability of the PORV's.
Since that time, Westinghouse has cotpleted the VCS plant specific cold overpressure protection analysis. Test conditions envelope these results. In order to Irovide conplete information, Appendix A of this attachment contains the revised pages of the " Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse Designed Plants."
Valve nozzle loads and accelerations are discussed in Attachment 4 of this subnittal.
Based on the test program output uid the results of the plant specific piping and support evaluation, the operability of the VCS PORV's has been demonstrated in accordance with the requirenents of NUREG 0737.
Page 2 of 2
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IAtter Fran O. W. Dixon, Jr., SCE&G, To H. R. Denton, Dated July 30, 1982 i
APPENDIX A f
" Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse Designed Plants" j
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Plants Name Owne'r NSP Prairie Island fl Northern States Power O ~ NRe erairie Isia#d #2 nort8er# stete ee er WPS Kewaunee Wisconsin Public Service Three-Loop s
Plants Name Owner s
SCE San Onofre #1 Southern California Edison i
CPL H. B. Robinson #2^
Carolina Power & Light Co.
FPL Turkey Point #3 Florida Power & Light Co.
FLA Turkey Point #4 Florida Power & Light Co.
VPA Surry'fl Virgihia Electric & Power.Co.
VIR Siirry #2 Virginia Electric & Power Co.
DLW Beaver Valley #1 Duquesn'e Light Co.
VRA North Anna #1 Virginia Electric & Power Co.
ALA Joseph M. Farley #1 Alabama Power Co.
VGB North Anna #2 Virginia Electric & Power Co.
APR Joseph M. Farley #2 Alabama Power Co.
CGE
- Virgil C. Sunmer #1 South Carolina Electric & Gas DW Beaver Valley #2 Duquesne Light Company CQL Shearon Harris #1 Carolina Power & Light Co.
CRL Shearon Harris #2 Carolina Power & Light Co.
CSL Shearon Harris #3 Carolina Power & Light Co.
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CTL Shearon Harris #4 Carolina Power & Light Co.
Four-Loop
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.s Plants Name Owner s.
i IPP Indian Point f2 Consolidated Edison Co. of New York INT Indian Point #3 Power Authority, State of New York x
' CWE Zion #1-N Comonwealth Edison
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CCN Zion #2 Comonwealth Edison AEP Donald C. Cook #1 American Electric Power Co.
Anert' an Electric Power Co.
AMP Donald C. Cook #2 c
PGE Diablo Canyon #1 Pacific Gas & Electric Power Page 3 of 7 N
'2.3 COLD OVERPRESSURE TRANSIENTS 2.3.1 Mass Input Events' Based on probability of occurrence and in-plant operating experience, the most credible mass input events producing a net infection of mass into the reactor coolant system (RCS) involve failure in the air supply system, which causes the charging flow control valve to open, and/6r isolation of letdown.
Mass injection based on single charging purgp operation is the most likely mass input mechanism, producing typical charging rates up-to 120 gpm following isolation of letdown, and higher r&tes for air supply syst'em failure.
Although precluded at low temperature by administrative procedure, two-chargic, pump operation was considered in all plants except Virgil C. Summer to develop; maximum input capability an'd thus provide additional flexibility in the opera-tion of the cold overpressure mitigation system.
For V.C. Summer, consistent with administrative procedure, single charging. pump operation was considered, together with the higher mass input rates associated with air supply system failure.
Maximum input capability associated with these mechanisms as applied to all plants analyzed to date is shown in Figure 2-1.
The PORV inlet condi-tions presented in Figure 5-1 also include these mechanisms.
Operation of t[he PORY at a predetermined setpoint pressure is employed by Westinghouse in the Cold Overpressure Mitigation System (OMS) to arrest the pressure transient caused by the~ above mechanismsP-
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Mitigation of the transient on valve opening results in ti$e 'RCS
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This produces a transieE peak"cverp'ressure.
The PORY continues to open until valye capacity matches the net mass -
injection rate, after which the reset pressure is reached and the valve begins to close.
PORY closure arrests the decreasing RCS pressure and reinitiates the pressure increase to complet!e the pres-d sure transient cycle.
This mimimum pressure is tenned the transient pressure undershoot and is detennined by the blowdown setting of the PORVs (nominally 20 psi).
Pressure cycling continues until action is taken to remove the mass input mechanism.
Selection of.PORY setpoints for pressure control of mass input-induced i
transients are based on a water-solid reactor coolant system, which produces pressure excursions significantly higher than for a RCS with 1
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Except for the V.C. Summer Plant, valve opening and closure times of 2 seconds are assumed, and valve setpoints are staggered such that opera-tion of the first valve will mitigate'the event so that the other valve will not be challenged.
2.3.2 Heat Input Events The heat input case which has the potential for the most severe pres-sure transient is that in which the steam ganerators exhibit a higher O
temperature than the remainder of the reactor coolant system.
The magnitude of the difference in temperature is dependent on the meahs by which the temperature asymetry was achieved, but a typical-0 difference is considered to be about 50 F.
For the heat input transient w'ith the initial reactor coolant tempera-0 ture 50 F less than the temperature in the steam generators and with all reactor coolant pumps off, one of the two reactor coolant pumps is started to circulate the reactor coolant through the warmer steam generators. As the coolant flow begins, the wam water in the tubes of the steam generator in the active loop is forced out and into the reactor coolant pump where it is pumped into and mixed with the colder reactor coolant.
In the inactive loops, the wamer water from the tubes of the steam generator is forced out in a reverse direction due to the backflow in the inactive loops, and also mixed with the cooler reactor coolant.
This initial mixing of the wam water with the larger, volume of cooler water causes an initial shrinkage effect which tends to decrease the initial coolant pressure.
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Simultaneously, the cooler reactor coolant that enters the steam generator begins to be heated as it moves through the tube bundle. As heat is added to the coolant due to heat transfer from the secondary water in the steam generator, the coolant attempts to expand and cause a resultant pressure increase. The net effect of the expansion due to the heat transferred to the coolant and the shrinkage effect due to the mixing of the wam water with the cooler coolant is a relatively 3
Measured valve opening / closure times for V.C. Uumer were 1.5 sec/1.0 sec Page 6 of 7
2507 psia with a maximum discharge rate of 628.3 gpm.
No liquid discharge from the safety valves of the 2-and.3-loop reference plants was observed during the U
analysis.
The fluid conditions at the inlet to PORVs range from 498 F to 569 F at 2353 psia with a maximum discharge rate of 1104.1 gpm.
In this cas'e no liquid diccharge from the PORVs of the 2-loop reference plant is observed during the interval that the transient was analyzed.
In general valves open on steam and no liquid. discharge is observed until the ressurizer becomes water solid..This is plant dependent and can vary anywhere from 20 minutes to more than six hours.
.4 PLANT-SPECIFIC VALVE INLET CONDITIONS RESULTING FROM COLD OVERPRESSURIZATION EVENTS
- Setpoints for the cold overpressurization mitigation system are conservatively determined to accommodate the rapid pressurization rates (up to 100 psi /sec) pro-duced by cold overpressure transients (Section 2.3) during water-solid, low temperature operation of the reactor coolant system.
In practice, however, fluid conditions at the relief valve inlet are not restricted to low temperature, sub-cooled water.
A variable fluid condition (steam or water) and temperature (satur-ated to subcooled) at the valve inlet is possible due to administrative require-O Vments for maintaining a pressurizer steam bubble during low temperature operations when pressure excursions due to cold overpressurization events are a possibility (Section 4.3).
The maximum range of potential cold overpressure fluid conditions at the relief valve inlet, covering all Westinghouse plants analyzed to date, may be inferred from Figure 5-1.
These plants include:
Comanche Peak Units 1 and 2, SNUPPS, Sequoyah Units 1 and 2, Watts Bar Units 1 and 2, South Texas Units 1 and 2, Byron /Braidwood Units 'l and 2, and Virgil C. Summer. A description of the' indexed curves used to define the range of potential fluid conditions is presented below.
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Legend Applicable To Figure 5-1 Index Description 1
Locus of maximum primary system pressures developed following PORY f
f2 operation.(limiting condition / water-solid RCS) 5-8 Page 7 of 7
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IAtter Fran O. W. Dixon, Jr., SCI %G, To H. R. Denton, Dated July 30, 1982.
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References:
a) letter fran T. C. Nichols, Jr., South Carolina Electric and Gas 03npany, to H. R. Denton, dated July 29, 1981.
b) letter fran R. C. Youngdahl, Chairman, EPRI Researdi Advisory Chmtittee, (bnsumers Power Ocznpany, to H. R. Denton, dated June 1,1982, with attachments.
As stated in reference (a) and docmented in reference (b), the PORV Block Valves installed at the V. C. Stumer Nuclear Station (VCS) are the same as the Westinghouse 3GM88 valves with the Limitorque type SB-00-15 operator that were tested at the Marshall Steam Station in conjunction with the EPRI Safety and Relief Valve test program. '1hese tests denonstrated the operability of the PORV Block Valves during full flow steam conditions.
Additionally, during the start-up test program, a less rigorous but equally damnstrative test was performed on each PORV Block Valve at VCS by successfully stroking then closed and then open with the PORV open at normal systen operating tenperature and pressure.
Included in reference (b) is a report prepared by Westinghouse that details tests and analyses perfoned cn Westinghouse Gate Valves to evaluate valve performance. Of significance in this report is that the Limitorque SB-00-15 develops adequate sten thrust to close the valvo and that the saturated steam fluid condition poses the greatest challenge to gate valve closure.
The successful testing of the VCS plant specific valves at Marshall, the in-plant tests during Hot Functional Test 2, and the tests, analyses and conclusions by Westinghouse conclusively demonstrates the PORV Block Valve operability for fluid cxmditions described in NURm 0737, Item II.D.l.
Furthermore, the operability demnstrated by the PORV's for all expected inlet fluid cxmditions greatly enhances the expected reliability of the PORV Block Valves.
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