ML20207T178
| ML20207T178 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/03/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20207T177 | List: |
| References | |
| TAC-60896, NUDOCS 8703230284 | |
| Download: ML20207T178 (4) | |
Text
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?,['..N SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OMAHA PUBLIC POWER DISTRICT FORTCALHOUN STATION, UNIT NO.1 DOCKET NO. 50-285 END OF CYCLE 9 CONTROL ELEMENT ASSEMBLY INSPECTIONS
1.0 INTRODUCTION
Ouring the refueling outage at the end of Cycle 9 (EOC-9), all full length control element assemblies (CEA) in Fort Calhoun Station, Unit No. 1, were inspected by Combustion Engineering (C-E) for Omaha Public Power District (OPPD) (Ref. 1).
The inspection program (Ref. 2) included eddy current test (ECT) profilometry measurements of circumferential cladding strain on all fingers of the 45 CEA's; also, mechanical profilometry was performed on selected TEA fingers in order to validate the ECT profilometry tech-nique.
Similar measurements were previously made on selected CEA fingers during the refueling outage at the end of Cycle 7 (EOC-7).
The fingers are composed of Inconel tubes containing pellets of B C Duringoperation,neutronreactionswiththeBCcausethepelleti.
3 to swell. When the initial gap between the peTlets and the Inconel cladding closes, additional swelling of the pellets produces strain in the cladding.
As the swelling increases, there is an increased chance of rupturing the cladding and exposing the pellets to the water in the reactor core.
The B C could then be leached away by the water, therebyreducingtheeffectivenessoftheCEA.
The supplier (C-E) of the CEA's to OPPD uses a conservative limit of "1% mean unrecoverable circumferential cladding strain" as a design criterion.
Because of the difference between the thermal expansion enefficients of Inconel and B,C, the limit for total measured strain is a combination of elastic add inelastic strain that is a proprietary amount.
Combustion Engineering used the results of the E0C-7 strain measurements to predict the maximum strains at EOC-9.
However, it appears that there were CEA fingers "in the core at EOC-7 which had considerably greater strains than were measured." Consequently, the maximum cladding strains measured at EOC-9 were higher than predicted; it should be noted, however, that all of the fingers had maintained their integrity. -
This review was performed in conjunction with the staff's technical assistance consultant at Battelle Pacific Northwest Laboratories (Ref. 5).
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2.0 EVALUATION l
a.
Actions Taken and Proposed Omaha Public Power Distr'ct replaced the 10 CEA's which contained fingers with the highest measured cladding strains, and plans to replace the remaining 33 CEA's at the end of Cycle 10.
C-E estimated the average incremental strain per operating cycle, using the difference in strain measured mechanically on 15 CEA fingers both at E0C-7 and at EOC-9 and OPPD has submitted these results in the subject report.
This average difference plus two standard deviations, calculated from the 15 measurements and incre6 sed i
to account for the number of data points, was added to the measured l
strains to predict the maximum strains at EOC-10.
The results of this projection indicate that there is less than a 5% probability that any i
of the 35 original CEA's would contain a finger with more than 1.13% total strain at EOC-10 (see response (Ref. 3) to Question 1 of staff request (Ref. 4) for additional information).
This level of total strain fs stated to be less than the design limit for total strain of the Inc3nel cladding.
"Although not necessary to justify continued use of the 35 CEA's of the original design," OPPD performed reactor physics calculations to analyze the effects that multiple finger failures would have on the reactivity following a main steamline break.
For the purpose of these calculations, the strain ilmit was assumed to be lower than their design criterior, and it was assumed that all 8,C was lost from all fingere w5tch tad a projected 5% probabilit9 of attaining this lower strain limit.
Note that eight fingers from the 10 CEA's, which were replaced at E0C-9, had attained or exceeded this assumed lower strain ilmit and maintained their integrity..The results indicate that, even under these conditions, the safe shutdown margin exceeds the requirements specified in the Fort Calhoun Station Updated Safety Analysis Report.
In addition, some of the original 35 CEA's lef t in for Cycle 10 operation 3
were moved to locations where a finger failure would have less effect on reactivity, and other original CEA's were rotated to reduce the neutrun fluence (and resultant incremental swelling) which the fingers having the highest strains would sustain during Cycle 10.
b.
Evaluation of Actions Taken.and Proposed 8y replacing 10 of the original control element assemblies (CEA) in Fort Cathcun Station Unit No. 1, and moving or rotating the remaining 35 original full-length CEA's, Omaha Public Power District has virtually ensured that none of the CEA fingers will exceed the design cladding strain Ifmit at the end of Cycle 10.
In addition, the design cladjing l
~s
. strain limit, set by OPPD and C-E, may be quite conservative; thrs, it appears highly unlikely that a failure of any of the CEA flagers will occur during Cycle 10.
The results of the reactor physics calculations indicate that multiple finger failures will not imperil the ability of Fort Calhoun Unit No. I to attain and maintain a safe shutdown condition in the avent of an accident involving a main steamline break.
By moving some of the original CEA's to positions where they have less effect on reactivity, OPPD has further increased the safa shutdown margin in the unlikely event of a catastrophic failure of a CEA finger and, perhaps, decreased the likelihood that such a failure could occur.
The planned replacement of the remaining 35 original full length CEA's for Cycle 11 on will further ensure that the design CEA cladding strain criterion will be met.
- 3. 0 CONCLUSIONS Based on the review described above, the staff concurs with the assessment made by OPPD regarding the CEA insrection results obtained at EOC 9 on cladding strain, the actions taken and the actions proposed.
The staff concurs that safe operation during Cycle 10 and future cycles is assured with regard to this CEA cladding strain issue.
Dated: March 3, 1987 Principal Contributor: D. Fieno I
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4.0 REFERENCES
1.
Letter (LIC-86-044) from R. L. Andrews (OPPD) to A. Thadant (NRC),
, dated February 28, 1986.
2.
"End of Cycle 9 CEA Inspection Results and Impact On Cycle 10 Operation," Special Report, Fort Calhoun Station Unit No.1, (transmitted by Reference 1).
3.
Letter (LIC-86-495) from R. L. Andrews (0 PPD) to D. E. Sells (NRC),
dated October 3, 1986.
4.
Letter from D. E. Sells (NRC) to R. L. Andrews (0 PPD), dated July 31, 1986.
5.
Letter from W. C. Morgan (PNL) to D. Fieno, (NRC), dated January 29,
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