ML20215C512
| ML20215C512 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 10/03/1986 |
| From: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| To: | Sells D Office of Nuclear Reactor Regulation |
| References | |
| LIC-86-495, TAC-60896, NUDOCS 8610100315 | |
| Download: ML20215C512 (5) | |
Text
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l Omaha Public Power District 1623 Harney Omaha. Nebraska 68102 2247
[C-g 402/536 4000 October 3, 1986 LIC-86-495 Donald E. Sells, Project Manager PWR Project Directorate #8 Division of PWR Licensing - B Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.
20555
References:
(1) Docket No. 50-285 (2)
Letter, LIC-86-044, R. L. Andrews (OPPD) to A. Thadani (NRC), dated February 28, 1986 (3)
Letter, D. E. Sells (NRC) to R. L. Andrews (0 PPD),
dated July 31, 1986
Dear Mr. Sells:
CEA Inspection Results During the end-of-Cycle 9 refueling outage, Omaha Public Power District I
conducted a Control Element Assembly (CEA) inspection program.
The re-sults of this program were reported in Reference (2).
Reference (3) contains three questions from the NRC's review of the CEA inspection report.
The attachment to this letter contains the responses to those i
questions.
If you have any further questions, please contact us.
Sincerely, l
i R. L. Andrews Division Manager Nuclear Production RLA/KCH/bjb Attachments l
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ATTACHMENT Ouestion 1:
Section 2.2 of the subject report has calculated an incremental strain for Cycle 10 that appears to be an estimated average incremental strain.
If so, what is the 95/95 upper confidence bound strain expected for the CEA's at the end of Cycle 107
Response
The upper 95 percent strain expected for the Fort Calhoun CEA's at the end of Cycle 10 is 1.13 percent. This value was calculated by statistically combining B0r-10 strain data and strain rate data.
The B0C-10 strain distribution was obtained from the measured E0C-9 data on the outer fingers of the CEA's present in the core during Cycle 10. The strains associated with the center fingers were conservatively ignored because the strains of the center fingers are, on the average, about 25 percent less than the strain of the outer fingers. This result is expected because the thermal flux of the center finger is partially shielded by the outor fingers.
The incremental strain predicted during Cycle 10 was based on strain rates obtained from data on CEA fingers that had been measured at both E0C-7 and E0C-9. A Gaussian distribution, based on the mean and standard deviation of the strain rate data, was assumed for the incremental strain accumulated during Cycle 10.
The standard deviation of the distrib:: tion was adjusted to account for the limited number of data points.
The E0C-10 strain predictions were calculated by convoluting the 80C-10 strain histogram with the Gaussian distribution of incremental strain. The reported value of 1.13 percent corresponds to the upper 95 percent value of the resulting histogram.
Question 2:
Section 3.4 of the subject report has estimated the effects of reduced scram worth on Hot Zero Power and Hot Full Power Main Steam Line Break accidents. Are these the bounding accident scenarios for evaluating control rod worth? Also, why is Cycle 10 considered to have a less severe cooldown curve than some previous cycles for this plant?
Response
The Hot Zero Power and Hot Full Power Main Steam Line Break acci-dents are the bounding scenarios. These events, as analyzed, re-sult from the break or rupture of a large pipe in the main steam system causing an uncontrolled heat extraction from the primary system.
In the presence of a negative moderator temperature co-efficient of reactivity, the steam line break causes a positive reactivity insertion which must be offset by CEA insertion (through reactor trip) to limit a return-to-power which occurs as a result of increased subcritical multiplication or a return-to-critical.
Attachment Page 2 The Hot Zero Power Main Steamline Break is the limiting event for determination of the Technical Specification on a shutdown margin, i.e. sufficient scram worth must be available to prevent a return-to-power, beyond that previously analyzed as documented in the Updated Safety Analysis Report.
Similarly, the Hot Full Power Main Steamline event also requires sufficient scram worth to prevent a return-to-power in excess of that previously analyzed.
This requirement prevents an excessive heat flux from occurring which could result in local boiling and departure from nucleate boiling.
The reason that Cycle 10 has a less severe cooldown curve than those of Cycles 1 through 8, is that in Cycle 9 the District began using DIT cross-sections in the physics reload safety analysis.
These cross-sections are valid for a range of moderator temperatures from room temperature to 800*K while previous analyses (i.e. Cycles 1 through 8) were performed with cooldown curves derived by conservatively extrapolating CEPAK cross-sections to low temperatures. Thus, due to enhanced modeling capabilities, the Cycle 10 cooldown curve does not contain these excess conservatisms of earlier analyses.
Question 3:
Do the ComSusion Engineering (CE) evaluation models for B(4)C swelling and CEA cladding strain underpredict the strain data from Cycle 9 of Fort Calhoun Station Unit I? If so, what plans are to be implemented to eliminate further underpredictions of cladding strains for CEA's Responsu Combustion Engineering designs CEA's for a ten year lifetime based on conservative estimates of B(4)C burnup, allowable strain of the Inconel cladding and dimensional tolerances between the elements and guide tubes.
Typically, the life limiting parameter for CEA's has been cladding strain.
This strain is caused by radial swelling of the B(4)C pellets as they are exposed to thermal flux. The rate at which the pellets swell and stress the cladding is dependent upon local exposure to thermal flux.
CE has attempted to use theoretical and empirical approaches in an effort to evaluate CEA cladding strain. Theoretical evaluation is complicated by several factors listed below:
1.
Local accumulated exposure levels to CEA's are difficult to obtain because they are dependent on CEA insertion histories, core power histories and CEA shuffling.
Attachment Page 3 2.
Part of the total B(4)C pellet growth closes the pellet-to-clad gap before it strains the cladding.
3.
Some B(4)C powder due to B(10) depletion is displaced in the gap as a result of normal flow induced vibrations.
As a result of difficulties in performing a theoretical evaluation, CE has adopted an empirical approach. Once the pellet-cladding gap closes and hard contact occurs, the cladding strain is approximated by any additional pellet swelling.
By measuring the cladding strain and plotting the data as a function of effective full power days (EFPD), the data can be used to conservatively project cladding strain as a function of additional EFPD's.
Strain measurements on six CEA's (17 fingers) were taken at Fort Calhoun at EOC-7.
Referring to Column 2 of Table 1, it can be seen that the maximum measured strain was 0.55 percent.
Using expected strain rates an a function of EFPD, it was not anticipated that the strain values at E0C-9 would be as high as were measured. At FCC-7 only a limited number of CEA's had been measured at Fort Calhoun and other utilities with a CE NSSS.
It was not recognized at tt,a time that the variation in strain could be so large. When measurements from all CEA's were taken at E0C-9, it became apparent that there were CEA's in the core at E0C-7 which had considerably greater strains than were measured.
It is now recognized that strain variations in CEA's in a core can be relatively large, and this variation is being accounted for by CE to avoid underprediction of projected cladding strains for CEA's.
4 s
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Table 1 Fort Calhoun CEA Mechanical Strain Measurements (%)
Measured Delta Strain CEA/Finaer E00-7 E00-9 Delta Strain 2 Cycles 1 Cycle 2/1 0.33 0.82 0.49 0.25 2/2 0.46 0.59 0.13 0.06 2/5 0.38 0.89 0.51 0.26 8/1 0.24 0.57 0.33 0.17 8/3 0.39 0.63 0.24 0.12 8/5 0.46 0.59 0.13 0.06 13/2 0.32 0.68 0.36 0.18 13/3 0.08 0.29 0.21 0.11 13/5 0.17 0.44 0.27 0.14 35/1 0.55 1.00 0.45 0.23 35/3 0.24 0.83 0.59 0.30 35/5 0.36 0.74 0.38 0.19 37/1 0.40 0.85 0.45 0.23 37/2 0.37 0.87 0.50 0.25 37/3 0.23 0.65 0.42 0.21 41/1 0.21 41/5 0.34 AVG.
0.33 0.70 0.36 0.18 St. DEV.
0.07
- Mechanical strain measurements not performed.
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