ML19331D756
| ML19331D756 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 08/29/1980 |
| From: | Justin Fuller PUBLIC SERVICE CO. OF COLORADO |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| FOIA-81-127 P-80281, NUDOCS 8009030540 | |
| Download: ML19331D756 (4) | |
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public service company o? esmeo 12015 East 46th Avenue, Suite 440; Denver, CO 80239 August 29, 1980 Fort St. Vrain Unit No. 1 P-80281 Mr. Robert L. Tedesco Assistant Director of Licensing Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Docket No. 50-267
Subject:
Thermal Stresses in Core Support Blocks
References:
(1) NRC letter from S. A.
Varga to D.W. Warembourg, G-79144, dated August 16, 1979 (2) IIRC letter from S. A.
Varga to D.W. Warembourg, G-79162, dated September 19, 1979 (3) LASL letter Q-13:80:218 from Thomas Butler, LASL, to Dr. iiichel Tokar, NRC, datec July 16, 1980 (4) NRC letter from R.L. Tedesco to J.K. Fuller, G-80130, dated August 2, 1980
Dear Mr. Tedesco:
References (1) and (2) notified the Public Service Company of Colorado (PSC) in 1979 that the Oak Ridge National Laboratory (0Rfil) had used the 0RECA code to predict that large temperature differences would exist between adjacent Fort St. Vrain (FSV) refueling regions during a firewater cooldown following a 90 minute loss of forced circulation (LOFC) accident, and that the flRC was having Los Alamos Scientific Laboratory (LASL) detennine if these large differential temperatures resulted in excessive thernal stresses in core support blocks.
At a meeting with NRC and PSC on November 1,1979, LASL reported that various preliminary thermal stress calculations had all resulted in acceptable stress levels. At the flovember 1,1979 meeting, the ilRC stated that no action by PSC was required or requested at that tine.
N000030 W O
l P-80281 l
August 29, 1980 Page 2 l
Reference (3), which was infomally submitted to PSC in July 1980, was the first indication that the large differential temperatures in the core support block area might result in high calculated themal stresses under postulated accident conditions.
Upon receipt of reference (3), PSC and the General Atonic Company (GAC) initiated an intensive review of the LASL thermal stress work completed since last flovember 1,1979.
Preliminary technical discussions with LASL and ORNL personnel were held, ar.d arrangecents were made for a technical neeting between, PSC, GAC, LASL and OPJiL on August 25, 1980, the earliest date on which key LASL and OR?il personnel were available.
Subsequent to these activities, PSC received reference (4) from the NRC officially forwarding reference (3) to PSL.
Reference (4) asked for PSC to reply to the LASL thermal stress analysis and discuss the potential crack stability and propagation effects and the effects the LASL analysis might have on the operation of the Fort St. Vrain reactor.
Re ference (4) requested PSC to reply by August 29, 1980.
A technical meeting was held on August 25, 1980 between personnel from LASL, OR!1L, GAC, PSC and NRC to review the recent LASL results and develop recommended approaches to further define and resolve the potential themal stress concerns. At the neeting LASL presented the results of their themal stress analysis for the highest differential temperature region of a worst case FSV equilibrium core (full fission product inventory) operating at the technical specification limits for region tenperature dispersion and region power peaking factors. This worst case equilibrium core was assumed to be operating at a 105 percent power level at the onset of an LOFC accident followed by a firewater cooldown 90 minutes into the LOFC.
At present, it is not clear what the net effect of the various analytical models and assunptions used by LASL might be on the calculated stresses, as compared to stresses which might be calculated using more rigorous and time consuning forms of analyses. At the meeting GAC identified several factors including, but not limited to, the effects of radiation heat transfer and the widely varying temperatures of the six blocks surrounding the hot core support block, that were not explicitly accounted for in the LASL analyses and which could significantly affect the calculated themal stresses.
PSC and GAC are continuing to review the LASL themal analyses that were perfomed using the assuned worst case equilibrium core and the postulated accident conditions to ascertain if a real problen in fact exists.
Based on several ORECA sensitivity studies perforced by ORfil in preparation for the August 25, 1980 meeting, PSC is confident that a themal stress problem does not exist with the current FSV Cycle 2 core and the present 70 percent themal power limitation.
Further, it is likely that even under full power conditions the Cycle 2 core, due to the relatively
P-80281 August 29, 1980 Page 3 unifom Cycle 2 region power peaking factors, will not experience excessive themal stresses under postulated accident conditions.
Because of their recent activities, ORfil and LASL at this time are in the best position to immediately analyze core support block temperature distributions and themal stresses.
Therefore, to confim these Cycle 2 conclusions on an expeditious schedule, ORfil and LASL participation in the analysis of the Cycle 2 FSV core will be necessary.
To better define the thermal stresses which may occur under postulated accident conditions, and the consequences of these themal stresses, it was the consensus of the participants in the August 25, 1980 meeting that the fallowing actions should be undertaken in the short tem:
1.
GAC will identify 2 to 3 bounding sets of Cycle 2 core perfomance parameters within or up to technical specification limits ud will provide these Cycle 2 parameter sets to ORf1L and LASL foc themal analysis.
2.
ORill will enhance the axial node detail of the ORECA code in the area of the bottom reflector elements 'and core support blocks to provide more temperature nodes and heat transfer rates between nodes for use by LASL. ORfil will take the GAC Cycle 2 core performance parameters and use the enhanced ORECA code to calculate bulk graphite temperatures and heat transfer rates between nodes for the Cycle 2 core.
3.
LASL will perfom the necessary 2-0 and 3-0 analyses to match the ORECA temperature patterns and calculate the associated themal stresses for the 2 or 3 bounding sets of Cycle 2 core perfomance parameters to confim there would not be an inmediate themal stress problen under Cycle 2 postulated accident conditions.
4.
GAC will continue in their efforts to obtain pemission from DOE to release the following reports concerning core support graphite fracture properties. Upon receipt of DOE pemission, GAC will submit the core support graphite reports to PSC, f1RC and LASL for infomation.
1 l
Test Evaluation Report of the Themal Stress Test for Core a.
Support Graphite l
b.
Test Evaluation Report for PGX Fracture l'echanics c.
Graphite Design P.aterial Properties S.
GAC will perfom scoping studies and analyses to detemine if a core support block with themal stress cracking would be able to continue to perfom its core region support function.
P-80281 August 29, 1980 Page 4 6.
LASL, ORNL, GAC, PSC and NRC representatives will reconvene in approximately one month to review the results of these short term efforts and determine the scope and direction of any long term efforts which may be required to resolve the potential core support graphite high thermal stress condition during a firewater cooldown following an LOFC.
By copy of this letter, the cooperation and participation of ORNL and LASL in performing the above short term analyses and efforts is hereby requested. PSC intends to pursue the resolution of the high calculated thermal stress concern in as short a time frame as possible, and use of the immediately available ort'L and LASL capabilities would greatly aid this effort.
If there are any questions or comments concerning the above approach to the potential high thermal stress condition, please contact PSC immediately.
Very truly yours,
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% 3-J.K. Fuller, Vice President Engineering and Planning JKF/f1HH:pa cc: 11r. Ron Foulds Division of Reactor Safety Research U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. Syd Ball Oak Ridge National Laboratory P.O. Box Y Oak Ridge, Tennessee 37830 Dr. Charles A. Anderson it.S. 576, Group Q-13 Los Alamos National Laboratory los Alamos, New flexico 87544 r
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