ML19309A321
| ML19309A321 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Rancho Seco  |
| Issue date: | 04/03/1973 |
| From: | Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8003270749 | |
| Download: ML19309A321 (3) | |
Text
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AEC DP7tIBUTION FOR PART 50 DOCXET MF" RIAL (TEMPORARY FORX)
CONTROL NO 2270-
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DATE OF DOC:
DATE REC'D LTR MEMO RPr OIHER Sacramento Municipal Utility.Dist] ict Sacramento, Calif 4 95813 E.K.Dwis 4-3-73 4-6-73 X
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1 50-312 DESCRIPIION:
EECLOSURES:
Ltr trans the following:
REPORT: Interim Report on Fuel Densification
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Mr. Angelo Giambusso i
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Deputy Director for Reactor Projects rp Directorate of Licensing
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U. S. Atomic Energy Commission
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20545
..f. a m Rancho Seco Nuclear Generating Station No. 1 AEC Docket No. 50-312 Inte' rim Report on Fuel Densification
Dear Mr. Giambusso:
The Sacramento Municipal Utility District he.reby submits an Interim Report on Fuel Densification. This submittal discusses the applicability of B&W Topical Reports BAW-10055, " Fuel Densifi-cation Report," and BAW-1388, "Oconee I Fuel Densification Report,"
and summarizes the difference between Oconee I and Rancho Seco 1 fuel design.
l Subject to the AEC completing their review of the B&W analytical models by May 1,1973, the District intends to submit a more thorough analysis of fuel densification in June 1973.
Sieglyyours, A., ~
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. K. Davis General Manager
' Enclosure
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AN ELECTRIC SYSTEM $@RVING MORM THAO @0@.000 !@ THE HEARi 0F C a t ' F 0 R B 11
1.
INTRODUCTION
'Ihe Regulatory Staff is currently evaluating B&W Topical Reports BAW-10055,
, " Fuel Densification Report," and BAW-1388, "Oconee I Fuel Dentification.
Report." It is anticipated that by May }, 1973 a,n evaluation will bc
~ completed on the application of B&W analytical models in accordance with the guidelines set forth in the USAEC report, " Technical Report on Densifi-cation of Light Water Reactor Fuels," dated November 14, 1972.
This interim report has two objectives:
(1) to provide the Staff with suf-ficient information to permit a comparison of Rancho Seco I with Oconee I and to dcscribe any changes in methods or epplication thereof in the final report for Rancho Seco 1, and (2) to present a preliminary evaluation of the tated power capability of Rancho Seco 1 even considering fuci densifi-cation effects.
2.
SUMMARY
Based on thIe analysis performed for Rancho Seco 1 using the methods outlined in BAW-10055, including a change in the power spike model and a change in g
input to the model (see Section 3), the following. conclusions are made assuming that the fuel pellets densify up to 96.57. theoretical density.
1.
Cladding will not collapse before 21,000 effective full power hours of operation (greater than one fuel cycle).
2.
The mechanical performance of B&W fuel rods will not be impaired.
3.
The interim acceptance criteria for the ECCS are met.
4.
The reactor can be safely operated at a power level of 2772 MWt with minor modifications to the reactor protection system setpoints resulting in a reduction of allowable imbalance limits.
3.
RESULTS 3.1 Power Spike Model The power spike model presented in BAW-10055 will be used as described, except for one minor modification,to the model and one change in input.
These are described in the following subsections, which are revisions of the original subsections in BAW-10055:
s2.3.4 Calculation of. Maximum Void Size in Axial Interval The equation for determining the maximum gap size as a function of axial position has been modified from (0.965 - pi + 2er(z)) + 0.004 L(z) 6L -
2 to the following k0.965-pi)+0.004 L(z)
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p The term a(s) has been eliminated as a result of discussion with the Staff 'and their evaluation of the Point Beach 2 Nuclear Plant analysis.
The calculational methods described in BAW-10055 have been changed to reflect the use of the DOT, Sn transport theory code in both X-Y and R-Z geometries. This changa will alter the two following subsections in the manner described.
B.W-10055 2.4.1 Power Spikes Due to a Single Gap page 2-6 The power spike in the adjacent rods will be obtained using DOT in geometry and the variation with gap length will be
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obtained using R-Z geometry. This change re-defines the power spike as the power increase in the adjacent pin rather than "the percentage increase at the edge of the surrounding fuel region" as given on page 2-6, Line 1, of BAW-10055.
B&W has known for some time that this defidition of the power spike due to a single gap is conservative and unrealistic and related work to determine the true' power increase in the adjacent pin has been completed.
The results of this work have been compared with those reported by Brookhaven National Laboratory in the BNL report, " Peaking Factors in Pressurized Water Reactors with Fuel Densification" (December 1972).- The last figure in this report (unnumbered),
" Correction Factor F on Finite Gap ~ Lengths," shows the' BNL values that are less than B&W values calculated for the same fuel l
enrichment. Therefore, B&W's method is more conservative than BNL's method.
l 2.4.2 Power Spikes Due to Co-Planer Gaps The DOT transport theory code in X-Y geometry will also be used instead of the PDQ07 diffusion theory code to determine the 4
power spike in a center rod due to more than one gap at the same axial height in a surrounding array of fuel rods. The transport calculation provides a better description of the physical system than does the diffusion theory calculation.
For the Rancho Seco 1 core, these correction factors will be calculated for the second-zone fuel enrich.nent (about 2.67 wt%)
since this zone experiences the highest power peaking of the three fuel enrichments in core 12 The preliminary data indicate that the power spike factor shown
._ in figure 2-12 of BAW-lu055 will have a maximum valus range of l
1.06 to 1.08 at 140 inches up the fuel rod channel. This large reduction in the spike factor compared to Oconee I is due to the increase in initial fuel density for Rancho Seco 1 core 1 to about 95% T.D.
The following discussions on several additional sections of BAW-10055 indicate the manner in which the final analysis for Rancho' Seco 1 will be performed.
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a-3.2 Thermal Analysis In the fuel temperature and DNBR analysis, the power spike factor will be applied in the manner approved by the Staff for BAW-10055. The as-built data for. Rancho Seco 1, of course, will be used in the final report on fuel densification.
The. preliminary evaluation has shown that the DNBR loss due to densi-fication from an initial value of about 95 to a final value cf 96.5%
T, D. reduces the power peaking margin by about 3 to 4%.
The fuel me: ting criterion will be about 21 kW/f t based on a maximum fifst cycle turnup of 20,800 mwd /mtU for the hot pin inithe hot assembly, incl ding a 10% uncertainty margin.
These effects will be compensated for by a.modificatira of the imbhlance limits such that the rated power level of 2772 MWt can be achieved with--
out violation of the thermal design criteria.
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3.3 Nuclear Analysis The Rancho Seco I reactor was designed to maneuver with a boric acid feed and bleed system. This results in lower steady-state power peaks compared to a " rodded" maneuvering plant like Oconee I and much lower operational peaks considering the design power transient.
The fuel for-the Rancho Seco 1 reactor is high density, about 95% T.D.,
resulting in a much smaller fuer densification penalty than the plants-with lower initial fuel density.
The net result is that the allowable imbalance limits, which are trans-lated into protection system setpoints, are broad compared to Oconee I including densification effects. After applying the samp philosophy to
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Rancho Seco 1 core as approved for Oconee I regarding LOCA, it is expected that operational flexibility will be greater than that shown for the Oconee I reactor.
3.4 Safety Analysis The ground rules and acceptance criteria adopted for the Oconee I report will be adhered to in the Rancho Seco 1 analysis.
3.5 Mechanical Analysis The application of the B&W clad collapse model will be the same for this application as that described in the Topical Report BAW-10055. The only change will be in the as-built data; and also cladding collapse is not predicted before 21,000 effective full power hours of operaticn.
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