ML19318B976
| ML19318B976 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/26/1980 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8006300372 | |
| Download: ML19318B976 (8) | |
Text
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Omaha Public Power District
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1623 HARNEY 3 OMAHA, NEGRASMA 68102 s TELEPHON E S36-4000 AREA CODE 402 June 26, 1980 Director of Nuclear Reactor Regulation ATTN: Mr. Robert A. Clark, Chief Operating Reactors Branch No. 3 U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Reference:
Docket No. 50-285 Gentlemen:
The Omaha Public Power District received a telecopy from the Commission containing questions in regard to the report entitled " Fort Calhoun Cycle 6 Small Break ECCS Performance Evaluation at 1500 MWt",
which was submitted to the Commission on May 21, 1980, in support of an Application for Amendment of Operating License (" Stretch Application")
filed on July 17, 1979.
Attached hereto are responses to the telecopy questions.
Sincerely, j./
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' WsC. Jones Division Manager Production Operations WCJ/KJM/BJH:jmm Attachment cc: Mr. Philip C. Wagner Project Manager U. S. Nuclear Regulatory Comission Mail Stop 228 Washington, D. C.
20555 LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N. W.
Washington, D. C.
20036
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Tills DOCUMENT CONTAINS P0OR QUAllTY PAGES 8006800 3Q
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ATTACHMENT Responses to Questions on Small Break LOCA Question 1 Provide Reference 8.
Response
Reference 8 of the submittal is " Letter from T. E. Short to George E.
Lear dated December 12, 1976".
Question 2 Provide statement and appropriate documentation that this evalu-ation is in conformance with Appendix K of 10CFR50.
Response
Reference 1 provides the results of the Fort Calhoun small break LOCA ECCS performance analysis at 1530 MWt.
The computer programs and calculative methods employed in the analysis were approved by the NRC and are identified in Reference 1.
The references for these computer progrc..;s and calculative methods listed in Reference 1 are presented again in this response as References 2 through 5 and 10.
The NRC's safety evaluation report, documenting approval of these codes and calcu-lative methods, is presented in References 6 through 9.
The NRC's safety evaluation reports provide the appropriate documentation that this evaluation is in conformance with Appendix X of 10CFR50.
Question 3 Provide justification that the break spectrum between 0.5 and 0.1-ft is not limiting. That is, further discussion that verifies the parameters of importance to the breaks at either end of this spectrum are at the extremes of their influence and that no new phenomena become significant for the intervening breaks, if required.
Response
The major reasons for the non-limiting nature or low peak clad temperatures for breaks between 0.5 and 0.1 ft2 are as follows. As the 2
break size increases above 0.1 ft, the depressurization rates in:rease.
The higher depressurization rates result in increased steam production during the blowdown due to increased flashing of the coolant which produces greater two phase level swell.
Although the fluid inventory loss increases as the break size increases, the dominant phenomena is the increased level swell as a consequence of the higher depressuri-zation rates.
Thus, these higher depressurization rates minimize core uncovery for breaks greater than 0.1 ft2 and smaller than 0.5 ft,
2 Also, for break; in this range, the depressurization rates are also sufficient to actuate the safety injection tanks which quickly terminate the clad temperature transient and minimize the duration of the core uncovery period.
The high pressure safety injection pumps therefore do not play a significant role in affecting core uncovery for these breaks.
- Response (Continued) 2 As the break size approaches 0.5 ft, the inventory loss becomes the dominating phenomenon.
As a consequence, as the break size in-creases, the core two phase level transients are characterized by increasing core uncovery.
This results in an increasing peak clad 2
temperature characteristic as the break size approaches 0.5 ft,
Break sizes 0.1 ft2 and smaller are characterized by slower de-pressurization rates which delay or prevent safety injection tanks from actuating.
The core two phase level transient and hence peak clad temperature for these break sizes are therefore controlled by the high pressure safety. injection pumps.
Reference 10 supports this discussion that breaks between 0.5 and 0.1 ft2 are non-limiting based on a spectral analysis of a range of breaks from 1.0 to 0.03 ft2 in area which demonstrated the limiting peak clad temperature to occur at the 0.1 ft2 break.
In sunmary, as a consequence of the higher depressurization rates, shorter blowdown times and increased core two phase level swell, breaks less than 0.5 ft2 and greater than 0.1 ft2 are not limiting for the Fort Calhoun plant.
Question 4 Provide justification, preferably experimental, for the " corrected low pressure SI pump flow" which has been assumed in the 0.5 ft break evaluation. Assure that the justification applies to the range of operational conditions under the hypothesized LOCA.
Provide evaluation of the uncertainties in pump performance characteristics.
Response
The high and low pressure safety injection performance curves, used in the small break LOCA analysis at 1530 ffdt presented in Reference 1, were based on experimentally measured head and flow parameters.
The safety injection flow credited in the Reference 1 analysis includes a 1%
uncertainty in head and a 3% uncertainty in flow.
The safety injection performance curves used in this analysis are shown in Figures 1 and 2.
Question 5 Provide documentation th the requirements of t4UREG-0635 have been satisfied.
Response
The District is implementing the recommendations of t1UREG-0635.
Detailed below are the major recommendation areas of fiUREG-0635 and the District's action.
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' Response (Continued)
Auxiliary Feedwater System A letter from Darrell G. Eisenhut to W. C. Jones, dated October 22, 1979, specified the auxiliary feedwater system requirements for Fort Calhoun Station Unit No.1.
The District has addressed the short term requirements and is in the process of addressing the long term require-ments.
Small Break LOCA Methods The submitted small break LOCA analysis uses the most recently approved small break LOCA model for the core when the reactor coolant pumps are assumed to be tripped.
This assumption is consistent with current operating procedures. We believe this model to be the best available licensing model.
The District is participating in the CE Owners Group for confirm-ation of the small break LOCA analysis methods.
The District will submit small break LOCA analysis applicable to Fort Calhoun Station Unit No. I when a revised licensing model is approved consistent with the schedule prescribed in the letter from Darrell G. Eisenhut, dated May 7, 1980.
Automatic Trip of RCP's The District is participating with the CE Owners Group for re-solution of the automatic trip of RCP's issue.
Natural Circulation The CE Owners Group has submitted predictions for the S-07-10B and L3-1.
We believe these analyses show the adequacy of the small break LOCA model's handling of natural circulation phenomenon.
PORV Failures in CE Plants The District will participate in the CE Owners Group's production of a report on P0FV failure reduction and will design and install an automatic PORV isolatior, system if required by the report.
Other Items The District is addressing other " Final Recommendations of the B&O Task Force" in accordance with our response to the letter from Darrell G. Eisenhut to All Operating ReactLe Licensees, dated May 7,1980.
Question 6 Provide references for NRC's safety evaluation for the quoted code verions which were used in this analysis.
Response
The computer code versions used in the small break ECCS analysis, listed in Section 6 of Reference 1, are presented again below along with the appropriate reference to the NRC's safety evaluation report.
Code Version No.
NRC Approval Letter CEFLASH-4AS 77019 Reference 6 STRIKIN-II 77036 Reference 7 COMPERC-II 74223 Reference 8 PARCH 77004 Reference 9 Question 7 Provide verification that the current core is identical to that assumed in the Cycle 6 reload analyses, e.g., no substitutions or burnup differences.
Response
The operating Cycle 6 core is within the bounds of the assumptions used in the Cycle 6 reload analysis.
Question 8 Provide verification that (1) there are no unusual anomalies associated with the CE, high burnup, demonstration assembly (e.g., power spike), (2) the demo assembly burnup will remain less than 40,000 EFPH, and (3) the reactor coolant system chemistry has been and will continue to be " normal".
Response
The Cycle 6 performance evaluation for the high burnup demonstra-tion assembly was provided in a letter from W. C. Jones to Robert W.
Reid, dated February 12, 1980.
The demonstration assembly has no unusual anomalies.
The assembly bi:rnup will remain within the bounds assumed in the analysis.
The reactor coolant system chemistry will be maintained within the Technical Snecification limits.
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-5 REFERENCES 1.
Fort Calhoun Small Bred ECCS Performance Results at 1530 FNt, Letter, W. C. Jones to Robert A. Clark, May 21, 1980.
2.
CENPD-133, Supplement 3, "CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident", January 1977.
3.
CENPD-135, Supplement 5-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", April 1977.
4.
CENPD-134, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core", April 1974.
5.
CENPD-138, Supplement 2, " PARCH, A Fortran IV Digital Program to Evaluate Pool Boiling, Axial Rod and ' olant Heatup", January 1977.
6.
Letter, Karl Kniel to A. E. Scherer _atitled, " Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P", September 27, 1977 (attached).
7.
Letter, Robert L. Baer to A. E. Scherer entitled, " Evaluation of Topical Report CENPD-135, Supplement 5", September 6, 1978 (at-tached).
8.
Letter, Olan D. Parr to F. M. Stern entitled, "NRC Staff Review of the Combustion Engineering ECCS Evaluation Model", June 13, 1975 (attached).
9.
Letter, Karl Kniel to A. E. Scherer entitled, " Evaluation of Topical Report CENPD-138, Supplement 2-P", April 10, 1978 (at-tached).
10.
CENPD-137, Supplement 1, " Calculative Methods for the C-E Small Break LOCA Evaluation Model", January 1977.
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