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f' S NUCLEAR REGULATORY COMMisslON
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WASHINGTON, D. C. 20555
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JUL 2 21977 Docket No. STN 50-561 VENDOR: Babcock & Wilcox Company (B&W)
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS B&W RESPONSES TO STAFF POSITIONS ON TWO OUTSTANDING ISSUES IN BSAR-205 REVIEW On July 12, 1977 representatives of B&W and our staff met in Bethesda to discuss B&W's responses to our positions on decay heat removal system isolation and overpressure protection of the reactor coolant system.
Our positions are described in Sections 5.2.2 and 5.4.3 of the Report to the ACRS issued on July 8,1977 and are the subject of outstanding issues which must be resolved prior to our decision on issuance of a Preliminary Design Approval for BSAR-205. is an attendance list.
B&W first presented their approach to DHRS isolation in the event of a pipe failure in the DHRS outside of containment. Their position is included as Enclosure 2 of this summary. B&W would establish an interface requirement in BSAR-205 that would require a referencing applicant to provide power to the required valves in such a way that no single active component failare would result in loss of power to either of the two series valves in either of the two DHR trains. B&W states that the failure of an electrical bus by a 3-phase.. fault should not be assumed (as a single active failure) in conjunction with the DHRS pipe failure outside containment. They contend that the concurrent failure of both the piping and the electrical bus are events for which the combined probability is too low to warrant design changes to accommodate such a postulated event. Staff members indicated their intent to require the assumption of bus failure, and also indicated that the probability values which B&W selected from WASH-1400 may not be conservatively applicable to the types of failures involved in the currer,t DHRS evaluation. We indicated that a documented B&W. position would be carefully considered.
The second issue discussed was that of overpressure protection for the reactor coolant system. B&W's R. Brockman described a graph showing two curves of reactor coolant pressure as a function of temperature. One curve iij h i g 75 810403 i
MADDEN 80-515 PDR
JUL %
1977 4
represented the Appendix G (to 10 CFR 50) limiting curve for heatup and 1
i cooldown operations, calculated assuming that a BSAR-205 plant had been operating for 32 effective full power years. The other curve showed the allowable reactor coolant system pressure limits which B&W proposed should be acceptable for the BSAR-205 system during an overpressure event.
The reactor coolant temperature range of interest, as presented by B&W, was between 305 and 345 degrees Fahrenheit. At lower temperatures the DHRS relief valves are designed to limit pressure to less than the Appendix G heatup and cooldown limits. Above 345 degrees Fahrenheit, the pressure allowed by the Appendix G limit is higher than the pressurizer safety valve setpoint.
In the temperature range between 305 and 345 degrees Fahrenheit',
the DHRS relief valves will be isolated from the reactor coolant system and the heatup and cooldown pressure limit will be less than the pressurizer safety valve setpoint. B&W proposed that the overpressurization event in this region has such a low probability that the allowable vessel stress during the incident should be based on emergency condition allowables (modified by fracture toughness requirements) of the ASME Boiler and Pressure Vessel Code.
The allowable pressure curve proposed by B&W is higher than the pressurizer safety valve setpoint in the 305 to 345 degrees Fahrenheit temperature range, thus their claim that equipment modifications are not necessary to mitigate the consequences of the potential overpressure event.
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Mazetis of the staff stated that if B&W feels that there is a strong argument that the overpressure event Trid" failure to miiiigate that
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combined occurrence probability equal to or less than 10~{ vent have a
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per year, then B&W should document that argument. Staff members present at the meeting agreed that sufficient justification for a revised allowable pressure curve, based on emergency conditions and including other changes to remove conservatism, would be difficult to achieve.
B&W committed to early documentation, perhaps within a week, of their position on the two issues discussed in this meeting.
Thomas H. Cox Light Water Reactors Branch No. 3 Division of Project Management Enclosures :
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Attendance List 2.
DHRS Isolation Capability cc:
See next page j
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Babcock & Wilcox Company ATIN: Mr. James H. Taylor JUL 2 21977 Manager of Licensing Nuclear Power Generation Division P. O. Box 1260 i
Lynchburg, Virginia 24505 cc: Washington Public Power Supply System ATIN: Mr. N. O. Strand Managing Director (Acting)
P. O. Box 968 3000 George Washington Way Richland, Washington 99352 Mr. Robert Borsum Bethesda Representative Babcock & Wilcox Nuclear Power Generation Division Suite 5515, 7735 Old Georgetown Road Bethesda, Maryland 20014 B. G. Shultz Project Engineer Stone & Webster Engineering Corporation P. O. Box 2325 7
I Boston, Massachusetts 02107 Mr. W. E. Kessler Ccanonwealth Associates, Inc.
209 East Washington Jackson, Michigan 49201 Robert J. Kafin, Esq.
115 Maple Street i
Glen Falls, New York 12801 Mr. B. M. Miller Chio Edison Company 76 South Main Street 4
Akron, Ohio 44308 i
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N ENCLOSURE 1 ATTENDANCE LIST MEETING OF B&W AND NRC STAFF July 12,1977 NAME ORGANIZATION i
0.1Parr DPM L. Riani DSS /ASB R. Fitzpatrick DSS /PSB J. Fair SD/EMSB J. Watt DSS /RSB J. Hamilton B&W J. Happell B&W R. Brockman B&W L. Cartin B&W D. Newton B&W G. Mazetis DSS /RSB S. Burwell DPM/ LWR #2 S. Newberry DSS /RSB T. Novak DSS /RSB T. Cox DPM
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D. Fischer
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J. Burdoin DOR /PSB
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SUMMARY
DISTRIBUTION
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Participants:
NRC PDR Local PDR L. Rfani TIC R. Fitzpatrick NRR Reading J. Fair r
LWR #3 File J. 'latt E. G. Case G. Mazetis R. S. Boyd S. Surwell R. D. DeYoung S. Newberry D. B. Vassallo T..Novak-J. Stolz D. Fischer K. Kniel J. Burdoin O. Parr S. Varga L. Crocker D. Crutchfield F. Williams R. J. Mattson i
H. Denton D. Muller Project Manager Tom Cox m
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DHRS Isolation Capability B&W Position The effects of pipe breaks outside containment are to be evaluated per the guidance specified in SRP 3.6.1 and BTP's APCSB 3-1 and MEB 3-1.
B&W's application of these criteria results in Classification of the DERS has a moderate energy system with a postulated 1.
pipe failure of a through wall leakage in the piping outside containment.
2.
Two power supplies, offsite and the emergency diesels are available.
3.
A single active component failure, as defined in Appendix A to APCSB 3-1, is assumed in the systems used to mitigate the pipe failure. The single active failure is pertinuat to the effected DHR train only (Section B.3.6 (3) of BIP APCSB 3-1).
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No assumption of a passive failure other than the piping failure itself.
Mitigation of the postulated piping failure is accomplished from the control room by
- 1) initiating HPI to make up for RCS leakage and 2) closure of one of the two redundant isolation valves in the DHR piping inside containment. Since both isolation valves are powered from the same electrical bus, the following interface criteria are proposei to accommodate a single active component failure:
1.
The applicant shall assure that the failure of isolation valves or their O &ndsa.it electrical power suppliesg re independent of effects of a pipe f ailure in a
the same fluid system outside containment.
2.
The applicant shall demonstrate that no one single active component failure, assuming both offsite and emergency diesel power are available, shall result in loss of all electrical power to each set of redundant series DERS isolation
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valves while the RCS is operating on the DHRS.
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o B&W does not consider failure of the electrical bus to be an active failure. This component has no moving parts, and the loss of structural integrity '.s precluded by definition of a single active component failure given in Appendix A to APCSB 3-1.
The postulation of a passive f ailure, such as the existence of a foreign object creating a 3-phase short, would be an event independent of the postulated piping failure and not a required assumption per the SRP and BIP noted above. Furthermore, i
the probability of such an event is quite remote. Based on the failure data in Appendix III to WASH-1400, the probability of the pipe failure alone is estimated to be N1.75x10" /yr (Table III-2 of WASH-1400). This event combined with a passive bus
-0 failure further decreases the probability to %3x10
/yr. It is our position that no design changes are warranted tb accommodate such an unlikely event.
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Fo 3 lI6 pfpe" Docket No.
STN 50-561 i
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VENDOR: Babcock & Wilcox Comapny (B&W)
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SUBJECT:
SUTtARY OF MEETING TO DISCUSS B&N RE.SPONSES TO STAFF POSITI0flS
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ON TWO OUTSTANDING ISSUES Ifl BSAR-205 REVIEW On July 12, 1977 representatives of B&W and our staff met in Bethesda to discuss B&W's responses to our positions on decay heat removal system isolation and overpressure protection of the reactor coolant system.
Our positions are describpd in Sections 5.2.2 and 5.4.3 of the Report to the ACRS issued on July,8,1977 and are the[ubject of outstanding issues which must be resolved prior to our decision on issuance of a Preliminary Design Approval for BSAR-205. An attendance list is enclosed.
B&W first presented their approach to DHRS isolation in the event of a pipe failure in the DHRS outside of containment. Their position is included as Enclosure 2 of this summary. B&W would establish an interface requirement in BSAR-205 that would require a feferencing applicant to provide power to the required valves in such a way that no single active component failure would result in loss of power to either of the two series valves in either of the two DHR trains.
B&W states iht the failure.of an electrical bus by a 3-fault should not be assumed (as a single active failure) in conjunction with the OHRS pipe failure outside cor.cainment. They contend that the concurrent failure of both the piping and the electrical bus 1%
are events for which the copbined probability is too low to warrant design changes to accommodate such a postulated event. Staff members indicated their intent to require the assumption of bus failure, and also indicated f,
W that the probability valyes which B&W selected from WASH-1400 may not be v
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l2-dIc> F 3 conservatively applicable to the types of failures involved in the current We indicated that a documented B&W position would be DH'lS evaluation.
s carefully considered.
The second issue discussed was that of overpr' essure protection for the reactor coolant system. B&W's R. Brockman,0escribed a graph showing two a,
t One curve curves of reacfor coolant pressure as a function of temperture.
4 represented the Appendix G (to 10 CFR 50) limiting curve for heatup and cooldown operations, calculated assuming that a BSAR-205 plant had been The other curve showed the operating for 32 effective full power years.
allowable reactor coolant system pressure limits whcih B&N proposed should
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be acceptable for the BSAR-205 system during an overpressure event.
.The reactor coolant temperature range of interest, as presented by B&W) was between 305 and 345 degrees Fahrenheit. At lower temperatures the DHRS relief valves are designed to limit pressure to less than the Appendix N** k G heatup and cooldown limits. Above 345{Fahrengeit, the pressure allowed 6&S valve setpoint.
by the Appendix G limit is higher than the pressurizer grew In the temperature range between 305 and 345jFahrenheit, the DHRS relief valves will. be isolated from the reactor coolant system and the heatup Cd and cooldown pressure limit will be less than the pressurizer valve i r 7 Y rs t1r y s w rr B&W proposed that the overpressurization evenghas such a 1%
setpoint.
probability that the allowable vessel stress during the incident should be based on emergency condition allowables (modified by fracture toughness The allowable requirements) of the ASME Boiler and Pressure Vessel Code.
pressure curve proposed by B&W is gher than the pressurizer valve setpoint in the 305 to 345* Fahrengeit temperature range,thus their claim S
that equipment modifications are not necessary to mitigate the congequences j
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3 c P.3 of the potential overpressure eveng. Mazetis of the staff stated that if B&W feels that there is a strong argument that the overpressure event and failure to mitigate that event have a combined occurrence probability deeditrirtf b f0 W
.1 g less than 104 per year, then B&N should W that argument. Staff SWA7eNyrl
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members present at the meeting agreed thap*JMt justification for a revised
%*y allowable pressure curve, based on emergency conditions and,gother changes to remove conf,ervatism, would be difficult t[
B&W committed to early documentation, perhaps within a wee of their position on the two issues discussed in this meeting.
Thomas H. Cox Light Water Reactors Branch No. 3 Division of Project Management A
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sin 50-561 m,m VENDOR: Babcock & Wilcox Comapny (B&'4)
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS B&W RESPONSES TO STAFF POSITIONS ON TWO OUTSTANDING ISSUES IN BSAR-205 REVIEW On July 12, 1977 representatives of B&W and our staff met in Bethesda to discuss B&W's responses to our positions on decay heat removal system isolation and overpressure protection of the reactor coolant system.
Our positions are described in Sections 5.2.2 and 5.4.3 of the Report to the ACRS issued on July,8,1977 and are the[ubject of outstanding issues which must be resolved prior to our decision on issuance of a cM l 4 gg/n attendance list.h crM"d Preliminary Design Approval for BSAR-205.
S&W first presented their approach to DHRS isolation in the event of a pipe failure in the DHRS outside of containment. Their position is included as Enclosure 2 of this summary. B&W would establish an interface requireme, in BSAR-205 that would require a,feferencing applicant to provide power to the required valves in such a way that no single active component failure would result in loss of power to either of the two series valves in either of the two DHR trains. B&W states that the failure of an electrical bus by a 3 fault should not be assumed (as a single active failure) in conjunction with the OHRS pipe failure outside containment. They contend that the concurrent failure af both the piping and the electrical bus e
are events for which the copbined probability is too low to warrant design changes to accommodate such a postulated event. Staff members indicated their intent to require the assumption of bus failure, and also indicated that the probability valyes which B&W selected from WASH-1400 may not be
4 do F 3 conservatively applicable to the types of failures involved in the current DH'lS evaluation. We indicated that a documented 3&W position would be carefully considered.
The second issue discussed was that of overpressure protection for the reactor coolant system. B&W'sR.Brockmanfescribedagraphshowingtwo t
a, One curve curves of reacfor coolant pressure as a function of temperture.
4 represented the Appendix G (to 10 CFR 50) limiting curve for heatup and cooldown operations, calculated assuming that a BSAR-205 plant had been operating for 32 effective full power years. The other curve showed the allowable reactor coolant system pressure limits whcih B&N proposed should be acceptable for the BSAR-205 system during an overpressure event.
The reactor coolant temperature range of interest, as presented by B&W) was between 305 and 345 degrees Fahrenheit. At lower ' emperatures the DHRS reTief valves are designed to limit pressure to less than the Appendix
- k ure allowed G heatup and cooldown limits. Above 345{ ahrengeit, the by the Appendix G limit is higher than the pressurizer
.c.... valve setpoint.
In the temperature range between 305 and 345jFahrenheit, the DHRS relief valves will be isolated from the reactor coolant system and the heatup
$af and cooldown pressure limit will be less than the pressu r q...
valve setpoint. B&W proposed that the overpressurization evInt has a1 A
probability that the allowable vessel stress during the incident should be based on emergency condition allowables (modified by fracture toughness requirements) of the ASME Boiler and Pressure Vessel Code. The allowable pressure curve proposed by B&W is gher than the pressurizer valve setpoint in the 305 to 345* Fahrengeit temperature range thus their claim 3
s that equipment modifications are not necessary to mitigate the congequences
3 P.3 of the potential overpressure eveng. Mazetis of the staff stated that i
if B&W feels that there is a strong argument that the overpressure event and failure to mitigate that event have a co7, bined occurrence probability doc M
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.1 sat 1ess than 104 per year, then B&W should eesssst that' argument. Staff 4
members present at the eeting agreed tha justification for a revised allowable pressure curve, based on emergency conditions and,M other changes g
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to remove congervatism, would be df fficult to achieve.
i B&W committed to early documentation, perhaps within a wee f their position on the two issues discussed in this meeting.
Thomas H. Cox Light Water Reactors Branch No. 3 Division of Project Management s
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