ML20126L233
| ML20126L233 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Millstone, Calvert Cliffs, Davis Besse, Browns Ferry, Salem, Oconee, Palisades, Indian Point, Kewaunee, Saint Lucie, Point Beach, Sequoyah, Arkansas Nuclear, Prairie Island, Surry, North Anna, Turkey Point, Crystal River, Haddam Neck, Ginna, Farley, Robinson, San Onofre, Cook, Yankee Rowe, Maine Yankee, Rancho Seco, Zion, Fort Calhoun, McGuire, Trojan, Crane  |
| Issue date: | 05/01/1981 |
| From: | Wigginton D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20126L232 | List: |
| References | |
| TASK-2.K.2.13, TASK-TM NUDOCS 8106020078 | |
| Download: ML20126L233 (11) | |
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STRIBUTION DWigginton WBrock PDR DELD EIgne MY 01 1981 L PDR IE-3 WDorie TERA CParrish JAustin f
NSIC ACRS-10 Docket flos.':
See Attached List of PUR's ORB #1 Rd9 NRC Participants J0bshinski j
JHeltemes, AEOD SVa rga SUMMAPY OF MEETING HELD ON APRIL 29, 1981, WITH THE PWR OWNERS GROUPS TO DISCUSS THERMAL SHOCK TO REACTOR PRESSURE VESSELS On April 29, 1981 the NRC staff met with the PWR Owners Groups and their vendor representatives (a) to discuss the status of the reports due to the Conmission on May 15, 1981, addressino this concern, (b) to pemit the owners groups the' opportunity to provide a generic basis for continued operation of PWP.s for the short tem and (c) to address the concerns ad-dressed in the D. Basdekas' letter to Concressman Udall dated April 10, 1981.
The list of Attendees is attached as Enclosure 1.
Each of the Owners Grouos was asked to respond to the three items above. The following sumnarizes that discussion.
BfM Owners Group The Owners Group noted that the analyses perfomed on B&W reactors in response to requirenent II.K.2.13 of 11UREG-0737 identified the worst them31 tansient occuring during the small break loss of coolant accident (LOCA).
This analysis is contained in the B&W generic themal shock report BAW-16a8.
BAV-1648 addresses small break LOCA and overcooling assuming 40 F ECCS water temperature without overpressurization. This analysis showed unaccept-able results for this case but acceptable results for the case of 90 F ECCS water. In the discussions, it was recalled that the overcooling was calculated to occur within 60 seconds. The Owners Group stated that this overcooling rate and other assumptions used in the 40 F case is too conserva-tive. The staff requested a reanalysis of the overcooling transient with repressurization, with suitable conservative assumptions to get an aopropriate bounding analysis. The Owners Group report to be submitted on May 15th will include discussions on the design basis LOCA, the small break LOCA, other overcooling transients, and a discussion of possible relaxation of certain conservative assumptions, for example, themal mixing of ECCS wcter with primary coolant.
l The fluence for B&W vessels is calculated usina a method described in a B&W reoort (BAW-1151-P) submitted to the staff in March 1981. Generic curves in BAW-1151-P and the effective full power years (EFPY) operation j
of the specific plant can be used to detemine the fluence on the vessel.
The EFPY for each reactor vessel is to be provided for the B&W reactors in the May 15, 1981 submittal.
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On the question of continued operation, the B&W Owners Group stated that operating BAH reactors have been shown to be adequately safe for the next year or two. The May 15th report will show that the fracture toughness of each operating reactor vessel is no less than the most limiting case used in the bounding thermal transient analysis. The report will not rely on a reduced probability of occurrance of an overcooling event but rather, will rely on naterial toughness as predicted by calculation of vessel fluence as the basis for continued operation.
The May 15th report will cover the near tern situation and not the requirement for the 40 year life of the vessel.
With regard to the uncertainty of calculating vessel fluence, the B&W Owners Group pointed out that they have not seen nore than a 15% difference between the actual and predicted fluence from surveillance capsule measurements.
They do not consider the difference at Maine Yankee, as was reported by the licensee, to apply to B&W operating reactors. With regard to overcooling transients resulting from control system f ailures a representative of Duke Power Conpany reported on a planned meeting between the staff and Duke Power to review control systen failure and their effects analyses of the Integrated Control System (ICS). This review should provide additional insight into system f ailures and'a better understanding of possible tran-sients initiated as a consequence of control systems failures and their overcooling effects compared to the worst case small break LOCA. The B&W plant Owners Group. indicated that operator actions to prevent pressurized thermal shock problems are being addressed in the on-going emergency procedure guideline development in response to the TMI action plan.
In conclusion, B8W owners group stated that based on BAW-1648 that operating reactors were adequately safe for the next year or two.
Westinnhouse Owners Group The Westinghouse (W) Owners Groups began with a statement that the analysis of the thernal shoch resulting from steam line break and LOCAs was acceptable for the W reactors for the next few years. The Owners Group will provide a scheduTe and program for resolution of this issue in the May 15th report.
The W Owners Group was directed to consider the comments made to the BM!
Owne7s Group and address them in the liay 15th report.
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'1 W stated that values of fluence calculated today differ from values pro-vided in Final Safety Analysis Reports. The comparison of measured fluence and calculated fluences today_ agrees within about 15% to 20% for both the inside wall fluence and the thru wall fluence. W presented a viewgraph,
!, which compared a calculated fluence (liiie) with measured fluence (points) as a function of time. This graph represents data and analysis for 2-loop W plants. U also presented a viewgraph showing the agreenent between the azmuthal fTuence obtained from calculations and from measurements (Enclosure 3). W was requested to provide a report discussing this compari-son. For any W reactor, the highest known vessel copper content was stated
.to be 0.36%. ~
Although the Rancho Seco overcooling event which occurred in 1978 would not be expected to occur in a W reactor, the themal and hydraulic conditions resulting from it were use7 to evaluate the effects on a reactor vessel in a W plant. This analysis indicated that the themal shock would be similar to that occurring from a small steam line break in a W reactor. Westinghouse pointed out that the themal shock effects would be bo_,unded by the worst large steam line break analysis. They further noted that the large steam l
generator water volume and associated themal inertia in a W plant nakes
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it less responsive to excess feedwater transients and subseliuent themal j
shock considerations.
In sunmary, W indicated that their analysis of vessel themal shnck included consideratioli of spec.ific material properties, evaluation of measured versus calculated fluences and review of operational events (including Rancho Seco event) to confim the bounding design-basis event. W concluded that all plants are acceptable though at least the end of 1982 and niost operating plants, if not all operating plants, could be denonstrated to have acceptable lifetimes significantly in excess of the end of 1982 (Enclosure 4).
CE Owners Group l
l The CE Owners Group prefaced their remarks by a statement that not all CE plant owners were represented by the Owners Group. The Owners Groups were advised that the staff Generic Letter 81-19 dated April 20, 1981 requests a commitment from each utility to participate in owner group discussions and provide a docketed response identifying the specific actions i
they would take for their facility.
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The CE plant Owners Group further stated that.no technical basis exists to discontinue reactor operation at this time based on themal shock with repressurization. The CE representative noted that as of now, the longest operation of arty CE plant is 5 EFPY.
Assuning a factor of 2 error in cal-culating vessel fluence, CE stated that these plants could operate for an ad-ditional 5 years based on a conservative analysis of the most severe overcooling transient (large steam line break).
CE stated that their stean line break was very conservative since it took no credit for operator action and assumed worst case initial conditions (i.e., zero reactor power and low levels in the steam generator).
In conclusion, CE stated that lower EFPY on CE reactor vessels plus large thermal inertia to reduce the severity of overcooling transients permitted them to conclude that a basis exists for continued operation for the next several years.
The CE Owners Group was also directed to consider the comments made in the pre-vious discussion with B&W and W and address them in the fiay 15th report.
Staff Sumary The staff nade the following sumary comments:
The Owners Group reports should present a generic basis for continued safe operation of the PWRs in the short term.
Owners Groups.should meet the May 15th date.
Owner Groups shoulb list the areas of conservatisms in their analyses in the flay 15th report.
Solutions of the problem should be in plac2 in the next couple of years; the schedule should be in the liay 15th reports.
All parties to the meeting agree that thermal shock with subsequent repressurization is a safety. concern that needs pronpt evaluation.
I ppp SIGED David Wigginton, Project Manager Ooerating Reactors Branch #1 Division of Licensing
Enclosures:
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UNITED STATES j
NUCLEAR REGULATORY COMMISSION y
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MAY 0 i 1981 i
MOTE TO: Tad Marsh B. D. Liaw John Austin Bill Dorie FROM :
Thomas M. Novak, Assistant Director for Operating Reactors, DL Attached for your information is the meeting minutes from the staff's April 24, 1981 meeting with the PWR Owners Group on thermal shock to reactor vessels. Because 'of your interest in this subject as indicated by your attendance, I am providing you with copies of the minutes.
We will keep you informed as this issue progresses.
2W N inomas M. Novak, Assistant Director for Operating Reactors, DL 1
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See Attached List of PWR's
SUMMARY
OF MEETING HELD ON APRIL 29, 1981, WITH THE PWR OWNERS GROUPS TO DISCUSS THERMAL SHOCX TO REACTOR PRESSURE VESSELS On April 29, 1981 the NRC staff met with the PWR Owners Groups and their vendor representatives (a) to discuss the status of the reports due to the Commission on May 15, 1981, addressing this concern, (b) to permit the owners groups the opportunity to provide a generic basis for continued operation of PWRs for the short tem and (c) to address the concerns ad-dressed in the D. Basdekas' letter to Congressman Udall dated April 10, 1981.
The list of Attendees is attached as Enclosure 1.
Each of the Owners Groups was asked to respond to the three items above. The following summarizes that discussion.
B&W Owners Group The Owners Group noted that the analyses perfomed on B&W reactors in response to requirement II.K.2.13 of HUREG-0737 identified the worst thermal tansient occuring during the small break loss of coolant accident (LOCA).
This analysis is contained in the B&W generic themal shock report BAW-1648.
BAW-1648 addresses small break LOCA and overcooling assuming 40*F ECCS water temperature without overpressurization. This analysis showed unaccept-able results for this case but acceptable results for the case of 90*F ECCS water. In the discussions, it was recalled that the overcooling was calculated to occur within 60 seconds. The Owners Group stated that this overcooling rate and other assumptions used in the 40*F case is too conserva-tive. The staff requested a reanalysis of the overcooling transient with repressurization, with suitable conservative assumptions to get an appropriate bounding analysis. The Owners Group report to be submitted on May 15th will include discussions on the design basis LOCA, the small break LOCA, other overcooling transients, and a discussion of possible relaxation of certain conservative assumptions, for example, thermal mixing of ECCS water with primary coolant.
The fluence for B&W vessels is calculated using a method described in a BaW report (BAW-1151-P) submitted to the staff in March 1981. Generic curves in BAW-1151-P and the effective full power years (EFPY) operation of the specific plant can be used to detemine the fluence on the vessel.
The EFPY for each reactor vessel is to be provided for the B&W reactors in the May 15, 1981 submittal.
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On the question of continued operation, the B&W Owners Group stated that operating B&W reactors have been shown to be adequately safe for the next The May 15th report will show that the fracture toughness year or two.
of each operating reactor vessel is no less than the most limiting case used in the bounding themal transient analysis. The report will not rely on a reduced probability of occurrance of an overcooling event but I
rather, wil'1 rely on material toughness as predicted by calculation of vessel fluence as the basis for continued operation.
The May 15th report will cover the near tem situation and not the requirement for the 40 year life of the vessel.
With regard to the uncertainty of calculating vessel fluence, the B&W Owners Group pointed out that they have not seen more than a 15% difference between the actual and predicted fluence from surveillance capsule measurements.
They do not consider the difference at Maine Yankee, as was reported byWith regard the licensee, to apply to B&W operating reactors.
transients resulting from control system failures a representative of Duke power Company reported on a planned meeting between the staff and Duke power to review control system failure and their effects analyses This review should provide additional of the Integrated Control System (ICS).
insight into system failures and a better understanding of possible tran-sients initiated as a consequence of control systems failures and their overcooling effects compared to the worst case small break LOCA. The B&W plant Owners Group 1,ndicated that operator actions to prevent pressurized thermal shock problems are being addressed in' the on-going emergency procedure guideline development in response to the TNI action plan.
In conclusion, B&W owners group stated that based on BAW-1648 that operating reactors were adequately safe for the next year or two.
Westinghouse Owners Group The Westinghouse (W) Owners Groups began with a statement that the analysis of the thermal sholik resulting from steam line break and LOCAs was acceptable The Owners Group will provide for the W reactors for the next few years.
a scheduTe and program for resolution of this issue in the May 15th report.
The W Owners Group was directed to consider the comments made to the B&W Owne7s Group and address them in the May 15th report.
~
. t W stated that values of fluence calculated today differ from values pro-vided in Final Safety Analysis Reports. The comparison of measured fluence and calculated fluences today agrees within about 15% to 20% for both the inside wall fluence and the thru wall fluence. W presented a viewgraph,, which compared a calculated fluence (line) with measured fluence (points) as a function of time. This graph represents data and analysis for 2-loop W plants. W also presented a viewgraph showing the agreement between the azmuthal fTuence obtained from calculations and from measurements (Enclosure 3).. W was' requested to provide a report discussing this compari-son. For any W reactor, the highest known vessel copper content was stated to be 0.36%.
Although the Rancho Seco overcoolin'g event which occurred in 1978 would not be expected to occur in a W reactor, the thennal and hydraulic conditions resulting from it were use7 to evaluate the effects on a reactor vessel in a W plant. This analysis indicated that the thermal shock would be similar to that occurring from a small steam line break in a W reactor. Westinghouse pointed out that the thennal shock effects would be bounded by the worst They further noted that the large steam large steam line break analysis.
generator water volume and associated thermal inertia in a W plant makes it less responsive to excess feedwater transients and subseliuent thermal shock considerations.
In summary, W indicated that their analysis of vessel thermal shock included consideration of spe.cific material properties, evaluation of measured versus calculated flueness and review of operational events (including Rancho Seco event) to confinn the beunding design-basis event. W concluded that all plants are acceptable though at least the end of 1982 and most operating plants, if not all operating plants, could be demonstrated to have acceptable lifetimes significantly in excess of the end of 1982 (Enclosure 4).
CE Owners Group The CE Owners Group prefaced their remarks by a statement that not all The Owners Groups CE plant owners were represented by the Owners Group.
were advised that the staff Generic Letter 81-19 dated April 20, 1981 requests a commitment from each utility to participate in owner group discussions and provide a docketed response identifying the specific actions they would take for their facility.
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b The CE plant Owners Group further. stated that no technical basis exists to discontinue reactor operation at this time based on thermal shock with repressurization. The CE representative noted that as of now, the longest operation of any CE plant is 5 EFPY.
Assuming a factor of 2 error in cal-culating vessel fluence, CE stated that these plants could operate for an ad-ditional 5 years based on a conservative analysis of the most severe overcooling transient (large steam line break). CE stated that their steam line break was very conservative since it took no credit for operator action and assumed worst case initial conditions (i.e., zero reactor power and low levels in the steam i
generator).
In conclusion, CE stated that lower EFPY on CE reactor vessels plus large themal inertia to reduce the severity of overcooling transients pemitted them to. conclude that a basis exists for continued operation for the next several years.
The CE Owners Group was also directed to consider the comments made in the pre-vious discussion with B&W and W and address them in the May 15th report.
Staff Sumary The staff made the following summary comments:
The Owners Group reports should present a generic basis for continued safe operation of the PWRs in the short tem.
Owners Groups should meet the May 15th date.
Owner Groups shoul'd list the areas of conservatisms in their analyses in the May 15th report.
Solutions of the problem should be in place in the next couple of years; the schedule should be in the May 15th reports.
All parties to the meeting agree that themal shock with subsequent repressurization is a safety concern that needs prompt evaluation.
David Wigginto roject Manager Operating Reactors Branch #1 Division of Licensing
Enclosures:
As stated l
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PRESSURIZED WATER REACTOR LICENSEES Docket No. 50-348 Docket No. 50-334 Beaver Valley Unit 1 Farley Unit i Docket No. 50-313 Docket No. 50-302 Arkansas Unit-1 Crystal River 3 Docket No. 50-368 Docket No. 50-335 Arkansas Unit 2 St. Lucie i Docket No. 50-317 Docket No. 50-250 Calvert Cliffs Unit 1 Turkey Point Unit 3 Docket No. 50-318 Docket No. 50-251 Calvert.. Cliffs Unit 2 Turkey Poir.t Unit 4 Docket No. 50-261-Docket No. 50 '15 H. B. Robinson Unit 2 D. C. Cook Unit 1 Docket N6. 50-295 Decket No. 50-316 Zion Unit i D. C. Cook Unit 2 Docket No. 50-304 DockeENo.50-309 Zion Unit 2 Maine Yankee Docket No. 50-213 Docket No. 50-289 Connecticut Yankee (Haddam Neck)
Three Mile Island Unit 1 Docket No. 50-3 Docket No. 50-320
' Indian Point Unit 1 Three Mile Island Unit 2 Docket No. 50-247 Docket No. 50-336 Indian Point Unit 2 Millstone Unit 2 Docket No. 50-296 Docket No. 50-282 Indian Point Unit 3 Prairie Island Unit 1 Docket No. 50-255 Docket No. 50-306 Palisades Prairie Island Unit 2 Docket No. 50-269 Docket No. 50-285 Oconee Unit 1 Ft. Calhoun
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Docket No. 50-270 Docket No. 50-344 Oconee Unit 2 Trojan Docket No. 50-287 Docket No. 50-272 Oconee Unit 3 Salem Unit 1
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t Docket-No. 50-244 R. E. Ginna 1 Docket No. 50-312 Rancho Seco Docket No. 50-206 San Onofre Unit 1 Docket No. 50-346 Davis-Besse 1 Docket No. 50-338 North Anna 1 Docket No. 50-280 Surry Unit 1.
Docket No. 50-281 Surry Unit 2 Docket No. 50-266
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Point Beach Unit 1 Docket No. 50-301 Point Beach Unit 2 Docket No. 50-305 Kewaunee Docket No. 50-29 Yankee-Rowe Docket No. 50-339 North Anna 2
! Docket No. 50-311 j
l Salem.2
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' Docket No. 50-327 Sequoyah 1
! Docket No. 50-359
' McGuire 1 Docket No. 50-364 Farley 2
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ENCLOSURE 1 THERMAL SHOCK MEETING
' MEETING PARTICIPANTS r
April 29, 1981 Orcanization Name Dave Wigginton NRR-ORB-1 William Bock ACRS Vince Panciera NRC/0CM R. Lobel NRC/ DST Earl J. Brown NRC/AE00 J. F. Walters Babcock & Wilcox R. J. Baker B&W R. E. Wascher B&W Robert Dieterich SMUD R. W.'K1ecker NRR/MTEB S. S. Pawlicki NRR/MTEB W. S. Hazelton NRR/MTEB John H. Austin NRC/0CM B. D. Liaw NRC/0CM Robert Gill Duke Power B. J. Short B&W T. A. Meyer Westinghouse W. J. Johnson Westinghouse B. 5. Monty Westinghouse i
J. N. Chirigos Westinghouse B. K. Singh NRR/RSB Bruce King Westinghouse
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Consumers Power R. A. Vincent TSE&E R. M. Douglas Baltimore Gas & Electric Co.
R. C. L. Olson Arkansas Power & Light Larry D. Young Florida P!wer Corporation William R. Klien Arkansas Power & Light Dan Howard i
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Meeting Participants r Name
~0rcanization Ted J. Meyers Toledo Edison P. K. Niyogi NRC/RES/DRA i
Howard Levin NRC/DE Richard P. Snaider NRC/DL/ ORB-5 Richard E. Johnson NRC/ DST /GIB E. Murphy DOE Guy A. Arlotto NRC l
C. Z. Serpan, Jr.
NRC/RES
. illiam J. Collins NRC/IE/ REB W
Paulette Tremblay NUS Divid G. Maire Westinghouse T. B. Natan CE W. E. Burchill CE D. J. Ayres CE J. M. Westhoven
' CE C. B. Brinkman CE(Bethesda)
P. S. Chech NRC/DSI A. Thadani NRC/ DST J. W. Roe NRC/ DOL R. B. Borsum B&W(Bethesda)
- 5. Varga NRC/DL Ed Wenzinger NRC/I&C'/DF0 l
Joyce M. Nelson QUADREX l
David W. Lippard VEPC0 Dennis Ziemann NRC l
Jim Clifford NRC OPPD-CEOG i
J. k. Gasper l
Kr.n Morris CEOG/0 PPD W Owners Group / Northeast Utilities I
rdk W. Wells John J. Mattimoe B&W SMUD I
John Olshinski NRC NRC D. Eisenhut NRC/ DST Tom Merley D. L. Easdekas I;RC/RES 1
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Meeting Participants -
I Name-
' Organization Larry Shao NRC E. G. Igne ACRS-
' Gary Holahan NRC/0L Don Croneberger GPU Daniel M.'Speyer con Edison R. M. Bernero NRC i
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