ML19322B759
| ML19322B759 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/04/1977 |
| From: | Parker W DUKE POWER CO. |
| To: | Rusche B, Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7912050755 | |
| Download: ML19322B759 (5) | |
Text
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FROM:
DATE oF DOCUMENT Duke Power Company 1/4/77 Charlotte, North Carolina oATE REcEivEo Mr. Benard C. Rusche Mr. William O. Parker, Jr.
1/11/77 MLE TTE R ONOTORf 2ED PROP INPUT FORM N 'MBEh CF COPIES RECEIVED foRiolN AL NNCLASSIFIED CCoPY One signed DESCRIPTaON E N CLOSU R E Ltr. w/ attached....re our 11/29/76 ltr...
concerning fission gas release for Oconee Units 1-2-3.
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Operating Reactors Branch //l Re: Oconee Nuclear Station kgEUlaf00r DOChef fjfg Docket Nos. 50-269, -270, -287
Dear Mr. Rusche:
With reference to your November 29, 1976 letter concerning fission gas release (FGR), it has been determined that the three (3) Oconee Nuclear Station power reactors have already reached a burnup exposure in excess of 20,000 megawatt-days per metric ton of uranium in some fuel rods.
Please note that since the new fission gas release correlation has been submitted strictly for evaluation purposes at this tine, the results should be reviewed accordingly.
In particular, we have strong reservations concerning the applicability of this model to Oconee's low enriched UO2 fuel rod design.
As requested by your letter, we have evaluated the impact of using the new fission gas release correlation in our thermal performance code (TAFY).
Calculations of pin pressure and temperature have been made.
These calculations were based on utilizing the TAFY code with and without the NRC FGR equation.
The TAFY analysis without the NRC FGR equation was taken from the Oconee II Cycle 1 licensing analysis. The input parameters and TAFY NRC restrictions are listed in the attached Table 1.
This input is representative of all three Oconee units. Results of pin pressure and fuel temperature calculations are shown in Table 2.
Since the input to the analysis using the two models are identical, the differences in pressures between the two models can be directly attributed to the dif ference in FGR models.
The fuel temperatures remain the same because of the NRC restriction that BOL temperatures be used for accident analysis.
i 324
Mr. Benard C. Rusche Page 2 January 4,1977 The results indicated that the internal fuel rod pressure remains below the neminal system pressure for fuel burnup up to and including the maximum expected burnup.
The TAFY code without the NRC FGR equation is the code used in the safety analyses of LOCA and other accidents. A survey of the safety analyses of all accidents was performed, and it was concluded that the NRC fission gas release model would not affect the results of the safety analyses of non-LOCA transients. The average fuel temperatures used in the transient cladding temperature calculations for these transients are not changed by the new NRC FGR model. The higher pin pressures at end-of-life would not result in cladding rupture during these transients.
In case of the LOCA analysis, if the NRC FGR Model is used in TAFY, then the LOCA analysis will be impacted in an unfavorable manner. This is the result of higher pin pressures occurring at earlier burnups.
Since initial inside and outside cladding surface oxide layers would be thinner at earlier burnups, the zircaloy-water (metal-water) reaction could possibly be larger than that previously calculated. The increased energy generation in the cladding er 'd possibly result in a peak cladding temperature increase in excess si 20 F.
The evaluation of the exact impact of this presumed higher fission gas release upon LOCA results would require an extensive analysis.
Duke Power Company believes that the staff's suggested fission gas release correlation is not applicable for the type of fuel and extent of burnup utilized in Oconee reactors and that further analysis is not warranted.
B&W has made an independent review of the literature and the available experimental data to determine the appropriateness of the staff's new model.
Although it appears that there is some enhancement of fission gas release at very high bura aps, B&W's evaluation of the available UO2 data indicates that the increase in release rates with high burnups occurs later in life and to a lesser extent than one would predict using the staff's suggested FCR correlation. B&W20 current approved analytical fuel pin model (TAFY),
used in calculating fuel temperatures and pin pressures, contains significant margins to ensure sufficient conservatism. The additional margin provided for in the NRC staff's FCR model becomes overly conservative when compared with current evaluation techniques.
In accordance with your specific request, three signed originals and 40 copies of this letter are submitted.
Very truly yours, d
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William O. Parker, Jr.
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TABLE 1 PIN PRESSURE ANALYSIS INPUT (OCONEE 11 NSSS-4)
FUEL INITIAL MEAN DENSITY - % TD 92.5 INITIAL MEAN DIAMETER - IN 0.370 INITIAL LTL DENSITY - % TD 92.0 FINAL DENSITY - % TD 96.5 DISH RADIUS - IN 0.150 DISH FACTOR 0.0170 INITIAL STACK LENGTH - IN 144 CLAD CLAD ID - IN 0.377 CLAD OD - IN 0.430 CLAD LENGTH - IN 153 INITIAL PLENUM VOLUME - IN 0.75 RESTRICTIONS 25% REDUCTION ON GAP NO RESTRUCTURING SORBED GAS CONTENT - 0.01 CC/GM USE BOL TEMPERATURES FOR ACCIDENT ANALYSIS g
a TABLE 2 PIN PRESSURE AND FUEL TEMPERATURE FOR OCONEE 1, 2, & 3 TAFY B&W FGR Model TAFY NRC FGR Model Pin Press.
Avg. Fuel Pin Press.
Avg. Fuel Peak Rod Burnup (psi)
Temp.* at (psi)
Temp.* at (MWD /MTU) 17 KW/FT (OF) 17 KW/FT (OF) 20,000 1210 2990 1210 2990 22,000 1235 2990 1240 2990 25,000 1295 2990 1320 2990 27,000 1340 2990 1410 2990 30,000 1400 2990 1550 2990 32,000 14aC 2990 1615 2990 35,000.
1470 2990 1745 2990 i
37,000 1510 2990 1865 2990 38,000 1525 2990 1925 2990
- BOL Values