ML19309B028
| ML19309B028 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/05/1977 |
| From: | Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| 256, NUDOCS 8004020505 | |
| Download: ML19309B028 (5) | |
Text
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NRC P Ru 196
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U.s. NUCLEAR REIULATCRY C'MISSloN DOCKET NUMCER (2 7e>
50 312 NRC DISTRrBUTION FoR PART 50 DOCKET MATERI AL TO:
FROM:
DATE oF DOCUMENT Sacramento Municipal Utility Dist.
1/5/77 Mr. Robert W. Reid Sacramento, California DATE RECElVED J. J. Mattimoe 1/10/77 k ETTER O NoToRIZE D PROP INPUT FORM NUMBER oF COPIES RECEIVED I
. AgRsGINAL EUNCLASSIFIE D OCoPY One signed DESCRIPTION ENCLOSURE Ltr. w/ attached....re our 11/29/76 itr...
concerning evaluation to determine the impact on the safety analysis and LOCA analysis of tha new NRC ' fission gas release model.
ACKNOWLEDGED
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SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 s Street, Box 15830 Sacramento, Califomia 95813; (916) 452-3211 January 5,1977
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9 ro Director of Nuclear Reactor Regulation 7
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Attention:
Mr. Robert W. Reid, Chief 9
s Operating Reactors, Branch 4 U. S. Nuclear Regulatory Comission b
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Washington, D. C.
20555 M U.s.rt.
cc s-u:/.'CY Docket No. 50-312 "E
Rancho Seco Nuclear Generating x\\
Station, Unit No. 1
Dear Mr. Reid:
Your letter of November 29, 1976 requested an evaluation to determine the impact on the safety analysis and LOCA analysis of the new NRC fission gas release (FGR) model.
The District understands this modal has been submitted at this time strictly for evaluation purposes and *.he results provided will be reviewed with this in mind. The District feels that this is appropriate as we have strong reservations concerning the applicability of this model to B&W's low enriched U02 fuel rod design.
It has been estimated that the Rancho Seco reactor will not exceed a burnup exposure of 20,000 megawatt-days per metric ton of uranium in any fuel rod until after June 1977.
Calculations of pin pressure and fuel temperature have been made for Rancho Seco as requested in your letter. These calculations were based on utilizing the TAFY code with and without the NRC FGR equation and are provided in Table 2.
The TAFY analysis without the NRC FGR equation was taken from the Rancho Seco Cycle 1 licensing analysis. The input parameters and TAFY NRC restrictions are listed in Table 1.
This input is representative of current B&W operating plants. Since the input to the analyses using the two models are identical, the differences in pressures can be directly attributed to the differences in the two FGR models.
Th'e' fuel temperatures remain the same because of the NRC restriction that BOL temperatures be used for accident analysis. The results also indicate the internal fuel rod pressure remains below the nominal system pressure for a fuel rod burnup equivalent to the maximum expected burnup.,
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Robert' W. Reid January 5, 1977 The TAFY Code without the NRC FGR equation is the code used in current 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 FGR model would not affect the results of the safety analyses of the non-LOCA transients. The average fuel temperatures used in the transient cladding temperature calculations are not changed by the new NRC FGR model and the higher pin pressures at E0L would not result in cladding ruptura 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 would be larger than that previously calculated.
The increased energy generation in the cladding would probably result in a peak cladding temperature increase in excess of 20 F.
The evaluation of the exact impact of this presumed higher fission gas release upon LOCA results would require an extensive analysis.
The District believes that the staff's suggested fission gas.
release correlation is not applicable for the type of fuel and extent of burnup utilized in Rancho Seco 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 appropriat6 ness of the staff's new model.
Although it appears that there is some enhancement of fission gas release at very high burnups, B&W's evaluation of the available U02 data indicate that the increase in release rates with high burnups occurs Tater in life and to a lesser extent than one would predict using the staff's suggested FGR correlation.
B&W's 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 FGR model becomes overly conservative when compared with current evaluation techniques.
Sincerely yours,
. AEh J. J. Mattimoe Assistant General Manager and Chief Engineer Attachments
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TABLE 1 PIN PRESSURE ANALYSIS INPUT
_ FUEL:
Initial Mean Density
- % TD 95.
Initial Mean Diameter
- IN
.368 Initial LTL Density.
- % TD 94.0 i
Final Density
- % TD 96.5 i
Dish Radius
- IN 0.150 Dish Factor 0.0170 Initial Stack Length
- IN 142.
CLAD:
i Clad ID
- IN
.377-Clad OD
- IN 0.430 j
Clad Length
- IN 153.
Initial Plenum Volume
- IN3 0.96
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RESTRICTIONS:
25% Reduction on Hgp No Restructuring Sorbed Gas Content
- 0.01 CC/GM Use BOL Temperatures for Accident Analysis i
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r TABLE 2
' DATA TAFY 88M FGR Model TAFY NRC FGR Model Avg.-Fuel Avg. Fuel Peak-Rod Burnup Pin Pressure Temp.* at Pin Pressure Temp.* at (PWD/MTU)
(psi) 17 KW/FT ( F)
(psi) 17 KW/FT (*F) 20,000.
1495 2865 1495 2865 22,000 1510 2865 1510 -
2865 25,000 1550 2863' 1550 2865 27,000 1570 2865 1570 2865 30,000 1625 2865 1665 2865 32,000 1665 2865 1725 2865 4
35,000 1725 2865 1855 2865 37,000 1765 2865 1950 2865 38,000 1785 2865 2000 2865 4
- BOL Values i
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