ML20032B291
| ML20032B291 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 10/30/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0830, RTR-NUREG-830 SLNRC-81-120, NUDOCS 8111050400 | |
| Download: ML20032B291 (6) | |
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4 SNUPPS Standardized Nucteer Unit Power Plant System 5 Choke Cherry Road Nicholas A.Petrick RockvHie, Marytend 20850 Executive Director (301) 8694010 October 30, 1981 SLNRC 81 120 FILE:
0541 SUBJ:
CPB Review x
D][N0y0 Mr. Harold R. Denton, Director /
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6 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission J
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Docket Nos.
STN 50-482 and STN 50-483
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Dear Mr. Denton:
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The NRC's Safety Evaluation Report (SER) for Callaway Unit 1 URG-0830) identified an open issue concerning fuel assembly structural response to seismic and LOCA forces. The attached SNUPPS FSAR change provides the information to resoin the issue and will be incorporated in the next FSAR Revision.
The Callaway SER also identified a confirmatory issue concerning cladding collapse time analysis. Westinghouse has calculated the cladding collapse times for all regions in the SNUPPS initial core using the methods given in the NRC-approved topical report WCAP-8377,
" Revised Clad Flattening liodel." For all fuel regions the cladding collapse time is in excess of 40,000 Effective-Full-Power-Hours. This resicierce time is in excess of the anticipated lifetime of the fuel.
Therefore, this confirmatory issue should be considered closed.
Very tquly yours, D
4%c Nicholas A. Petrick RLS/mtk Attachment cc:
D. F. Schnell UE oI G. L. Koester KGE o
D. T. McPhee KCPL W. Hansen NRC/ CAL T. E. Vandel NRC/WC f (
Q10504co811030 ADOCK 05000482 E
SNUPPS 4
e C.
Recent Technical Issues With regard to the seven current technical issues presented in question 490.1, it is SNUPPS's understanding that many of the generic issues have been resolved in connection with NRC staff reviews of similar plants with fuel assembly designs and fuel fabrication speci-fications that are the same as those for SNUPPS.
The Safety Evaluation Report for the Virgil C. Summer Station (NUREG-0717) is an example of such a plant.
The following paragraphs address these issues.
1.
Supplemental ECCS analysis with NUREG-0630 NUREG-0717 describes the current status of NRC requirements relative to ECCS evaluation models.
SNUPPS plans to comply with current NRC require-ments and provide a supplemental calculation of the plant ECCS analysis per',rmed with the materials models of NUREG-0630 on a mutually agreeable schedule.
We expect this calculation to demon-strate that no total peaking factor reduction will be required for the SNUPPS reactors.
2.
Combined seismic and LOCA loads analysis The combination of seismic effects and loads due to a double ended loss-of-coolant accident are discussed in the SNUPPS FSAR Section 4.2.3. W st nghcu-^ t^pic;l r pcrt WC'" 0230/^2^O ("c' rence
.3 of Se ion 4.
f the S
AR), an on pa 6 of R.
-07 n
e lat repor the d
ce o
e of fuel emblie r se' mi an Insert A l LOCA cads h een al 'ed ith hodol y
occ pta le o th N C, an e resul show h
asse ies wi accomm d e the e 1 d.
If 1
ilar na vsis s
qu' ed i
- PPS, ant' pa at it 1 als how tha the S P
asse.
ies wi ccomm da e the e aqds in an
- cc_ytab.c m_..ncr.
3.
Enhanced fission gas telease analysis at high burnups The subject of fission gas release is discussed in Westinghouse topical report WCAP-8720/8785 (Ref-erence 5 in Section 4.2 of the SNUPPS FSAR).
The NRC Safety Evaluation Report for the Virgil C.
Summer Station (NUREG-0717) indicates that the analysis presently docketed for that plant is acceptable for first cycle operation at full power.
Once the NRC review of WCAP-8720/8785 has been completed and the remaining issues have been resolved, SNUPPS anticipates that operation of the fuel for subsequent cycles will be shown to be acceptabic 490.1-6 Rev. 9
INSERT A 4
The fuel assembly response resulting from the most limiting main coolant pipe break was analyzed using time history numerical techniques.
The vessel motion for this type of accident causes primarily lateral loads on the reactor core. Consequsntly, a finite element model similar to the seismic model described in References 1 and 2 was used to assess the fuel assembly deflections and impact forces.
The time history motions of the upper and lower core plates and the barrel at the upper core plate elevation which are simultaneously applied to the simulated reactor core model as input motion were obtained from the time history analysis of the reactor vessel and internals. The fuel assembly response, namely the displacements and grid impact forces, were obtained with the reactor core model by using the motions resulting from a reactor pressure vessel inlet nozzle break which produced the limiting structural loads for the fuel assembly. The maximum g. id impact forces for both the LOCA and seismic accidents occurred at the peripheral fuel assembly locations adjacent to the baffle wall.
The maximum grid impact forces obtained from the nozzle inlet break and seismic analyses were approximately 39 and 60 percent of the allawable grid strength, respectively.
It should be noted that the maximum grid impact forces obtained from the two accidents did not occur at the same grid elevations.
INSERT A (CONT.)
s With respect to the guidelines of Appendix A of SRP Section 4.2, Westinghouse has demonstrated that a simultaneous SSE and LOCA event is highly unlikely. The fatigue cycles, crack initiation and crack growth due to normal op( t ating and seismic events will not realistically lead to a pipe rupture (Reference 3).
The factor applied to the LOCA grid impact load due to flashing is considered unrealistic since the thermal / hydraulic conditions for flashing are not present at the time of peak grid impact load.
However, a calculation of the grid maximum combined impact forces for the SNUPPS units was performed consistent with the guidelines of SRP Section 4.2, Appendix A.
The resulting value was approximately 73 percent of the allowable grid strength.
SNUPPS 4.
Fuel rod bowing analysis The subject of fuel rod bowing is discussed in Section 4.2.3 of the SNUPPS FSAR, as well as Westinghouse topical report WCAP-8691/8692 (Ref-ere.lce 11 of Section 4.2 of the SNUPPS FSAR).
Although review of this topical report by the NRC has not been completed, SNUPPS anticipates that the current methods used by Westinghouse to eval-uate fuel rod bowing will be found to be accept-able.
This was the case with the Virgil C.
Summer Station evaluation.
5.
Fuel assembly control rod guide tube wear analysis Westinghouse topical report WCAP-8278/8279 (Ref-erence 10 of Section 4.2 of the SNUPPS FSAR) presents flow test results for fretting wear at contact points between the control rods and con-trol rod guide thimbles.
Additional experimental data has been submitted to the NRC by Westinghouse (see W letters NS-TMA-1936, 1992, and 2102), and a post-Irradiation examination program has been established to address this specific subject (see NUREG-0717).
We anticipate that the information j/ -
derived from this program will confirm the Westing-house predictions, and that this issue will be resolved for SNUPPS as it was for Virgil C.
Summer Station.
6.
Fuel assembly design shoulder gap analysis Appropriate rod-to-nozzle gaps will be provided in the SNUPPS fuel to accommodate thermal expansion and irradiation-induced growth of the fuel rods relative to the overall fuel assembly structure.
Westinghouse's ability to model fuel rod growth has been confirmed by comparison with measurements from 15 x 15 and 17 x 17 in-reactor data, and also is in good agreement with established experimental results as discussed in Reference 4.
i Reference Insert B F~~
4, Balfour, J.
B.,
- Destefan, J.,
Melehan, M.
G.,
and
- Cerni, S.
" Evaluation and Performance of Westing-house 17 x 17 Fuel," presented at the ANSI Topical Meeting on LWR Fuel Performance held April 30 through May 2, 1979.
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490.1-7 Rev. 6
INSERT B 1.
Gesinski, L. and Chaing, D., " Safety Analysis of the 17 x 17 Fuel' Assembly for Combined Seismic and Loss-of-Coolant Accident,"
WCAP-8236 (Proprietary) and WCAP-8288 (Non-Proprietary),
December 1973.
2.
Beaumont, M. D. et. al., " Verification Testing and Analyses of the 17 x 17 Optimized Fuel Assembly," WCAP-9401-P-A (Proprietary) and UCAP-9402-A (Non-Proprietary), August 1981.
3.
Witt, F.
J., Bamford, W. H., and Esselman, T. C., " Integrity of the Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," WCAP-9283, March 1978.
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