ML20003B032
| ML20003B032 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 02/05/1981 |
| From: | Justin Fuller PUBLIC SERVICE CO. OF COLORADO |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20003B033 | List: |
| References | |
| FOIA-81-127 P-81044, NUDOCS 8102100244 | |
| Download: ML20003B032 (3) | |
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public service company oe essendo Februcry 5,1981 Fort St. Vrain Unit No. 1 P-81044 Mr. Robert L. Tedesco Assistant Director of Lice 1 sing Division of Licensing U. S. Nuclear Regulator; Commission Washington, DC 20555 Docket No. 50-267
Subject:
Themal Stresses in Core Support Blocks
References:
(1) NRC letter from R. Tedesco to J.K. Fuller, G-80130, dated August 2,1980 in.
(2) PSC letter from J.K. Fuller U
to R. Tedesco, P-80281,
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'5 il dated August 29, 1980 5
,R (3) ORNL letter from S. Ball
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to M. Tokar, G-80225, a
dated December 17, 1980 l
(4) LASL letter from p,
T. Butler to G. Kuzmycz, m
Q-13-81:46, dated
{
February 4, 1981
Dear Mr. Tedesco:
Public Service Company of Colorado (PSC) was asked, by reference (1), to evaluate the thermal stresses in core support blot.b of the Fort St.
Vrain (FSV) reactor during a postulated loss of forced circulation (LOFC) accident followed by a fire water cooldown.
PSC responded, in reference (2), with the conclusion that a themal stress problem did not exist with the current FSV Cycle 2 core during operation up to 70 percent themal power. PSC also proposed several short term actions that would be undertaken to address the' thermal stress question. The results of those short term actions were presented to the NRC by representatives of PSC, General Atomic Company (GAC), Los Alamos Scientific Laboratory (LASL) and Oak Ridge National Laboratory (ORNL) at a meeting in NRC offices on November 7,1980. At this meeting the participants concurred that there was not a thermal stress concern for the FSV Cycle 2 core at up to 70 percent thermal power, and future actions were agreed upon to analyze the full power, equilibrium worst case FSV core.
I 8102100#W
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P-81044 Page 2 February 5,1981 Subsequent to the November meeting, ORNL completed their thermal analysis of the urst case FSV reactor core, referred to as the Equilibrium Shim Bank-3 (EQSB3) core, at 1057, thermal power using an enhanced ORECA Code. ORNL transmitted their themal results to LASL in reference (3) for stress analysis.
LASL, in turn, has completed their worst case stress analysis of the core support block. The latest results, reported in reference (4),
indicate that the maximum tensile stress in a core support block was calculated to be 1016 psi. This compares favorably with the minimum ultimate strength of PGX graphite and supports the conclusion that cracking of the core support block will not occur under the postulated accident conditions.
Nevertheless, recognizing the uncertainties in the analyses and that the predicted stress is rela:ively close to the minimum ultimate strength, PSC has had GAC perform a further safety assessment conservatively assuming that cracks occur from the corners of the keyways to the adjacent coolant channels and, further, between coolant channels. GAC's structural analysis of this configuration, GA-C16190; " Structural Safety Evaluation of a Cracked Fort St. Vrain Core Support Block During LOFC with Firewater Cooldown", is submitted as an enclosure to this letter. The GAC analysis leads to the conclusion that, even if a core support block were conservatively assumed to crack in all of tne most probable cross sections, minimal disarray of the fuel columns would occur ;nd it would not interfere with safe cooldown of the core.
PSC believes that this matter has been thoroughly evaluated and that the results provide sufficient assurance that the structural integrity 7
of the FSV reactor core will be maintained under the postulated circumstances, and that safe cooldown of the core would not be adversely affected in any event.
PSC concludes that operation of the Fort St. Vrain reactor is not affected by core support block thermal stress considerations.
1 If there are any questions or comments concerning this submittal, please contact Mr. Michael H. Holmes (303) 571-6711.
l Very truly yours, e
J. K. Fuller, Vice President Engineering and Planning l
JKF/JL/pa Attachment i
l P-81044 4
Page 3 3
February 5, 1981 cc: Mr.-Ron Foulds Division of Reactor Safety Research U.S. Nuclear Regulatory Commission i-Washington, D C.
20555
-Or. Syd Bal',
Dak Ridge f.ational Laboratory P.O. Box Y Oak Ridge, Tennessee 37830 Dr. Charles A. Anderson M.S. 576, Group Q-13 Los Alamos Scientific Laboratory Los Alamos, New Mexico _ 87544 4
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