ML20039C130
| ML20039C130 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 12/23/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SLNRC-81-131, NUDOCS 8112280456 | |
| Download: ML20039C130 (7) | |
Text
_ _ _ _ - _ _ _ _
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f SNUPPS i Stander' dised Nuclear Unit Power Mont System.
5 Choke Cherry Road Nicholas A.Petrick RoskvHle, Maryland 20060 Executive Director 1301) 800 8010 r
December 23, 1981 SLNRC 81-131 FILE: 0541 SUBJ: CPB Review e
Mr. Harold R. Denton, Director 8
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" D ;) g Office of Nuclear Reactor Regulation
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3 U.S. Nuclear Regulatory Commission
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Docket Nos: STN 50-482 and STN 50-483
Reference:
SLNRC-81-120, dated October 30, 1981: Same subjec 'd.iTTsil({p rr s
Dear Mr.' Denton:
The referenced letter provided information on the outstanding issue in the Callaway Safety Evaluation Report concerning fuel assembly structural response to seismic and LOCA forces. The Core Performance Branch sub-a sequently indicated that additional'information. was required to completely close this. issue. Attached to this letter is a change to the enclosure of the referenced letter. This attachment will be incorporated in the next SNUPPS FSAR Revision.
Very truikyours, s
i
@VV licholas A. Petrick RLS/jdk Attachment cc:
D. F. Sch.wll UE l
G. L. Koeste.r KGE D. T. McPhee KCPL T. E. Vandel NRC/WC Resident Inspector NRC/ CAL p/
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4 8112280456 811223 PDR ADOCK 050008 E
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SNUPPS I
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C.
Recent Technical Issues 1
i 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 j
with fuel assembly designs and fuel fabrication speci-fications that are the same as those for SNUPPS.
The i
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 l
l requirements relative to ECCS evaluation models.
l SNUPPS plans to comply with current NRC require-ments and provide a supplemental calculation of the plant ECCS analysis performed 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.dr;n;ce U
t
'..;h u-- t pical r port h'C'" C22 0'^2 ^2 e
.3 of Se ion 4.
f the S
AR), an on pa 6 of
-07 n
e lat repor the d
re o
e of fuel semblie r se' mi an Insert A l LOCA oads h een al ed ith hodol ac pta le o th N
an e resul show asse.
ies wi accomm d e the e 1 d.
If 1
ilar na sis s
qu' ed f
- PPS, ant'
.pa at it 1 als how tha the S P
asse
'es wi ccomm da the e 1 ds in an
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3.,
Enhanced fission gas release 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 1
been completed and the remaining issues have been resolved, SNUPPS anticipates that operation of the fuel for subsequent cycles will be shown to be i
acceptable.
I Rev. I i
490.1-6
INSERT A The fuel assembly respense resulting _from.the most limiting main coolant pipe break was analyzed using time history; numerical techniques.
The vessel motion for this type of' accident causes7primar.ily lateral
~
i loads on the reactor core.
Consequently,- a finite element model'siinilar
~
to the seismic model described in References 1 and '2 was used' te as ss 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 simultan.eously applied to the simulated reactor core model as input motion were'
^
obtained from the time history analysis of the reactor vesse11and-internals.
The fuel assembly response, namely the displacemsnts and
[
~
@ impact forces, were obtained with the reactor core model by-using the motions resulting from a reactor pressure vessellinlet nozzle break which produced the. limiting structural loads for the fuel assembly.
The d
Gnd Amtvses maximum grid ~ impact forces for both the LOCA and seismic accidehts s
I-occurtEl at the peripheral fuel assembly locations adjacent ta the s
~
baffle wall.
The maximum grid impact forces obtained from the nozzle inlet break and seismic analyses were approximately 39 and 60. percent-of he r
allowable grid strength, respectively.. It should be noted that th'e
., ~ '
maximum grid impact forces obtained from the two accidents did not' q
- m, occur at the same grid elevations.
lC x
A >
\\
4 4
x a
l INSEM A (C0fj M J With respect to the guidelines of Appendix A of SRP Section 4.2, Osstinghouse has detonstrated that a simultaneous SSE and LOCA event is
~
' highly unlikely.
Thi fatigue cycles, crack initiation and crack growth due to normal operating 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 ccasidered 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.
Non-Gnd Ano/yes The sfresses mduced m >%e n2rmus fuel assembly non-gnd componen/s sre assessed based on +he mad hmiJmg seisme and locA cond Jrons. 7Ae fuel assembly oxial forces sesullmg from a 20CA are y%e pnnt2ry Souce of /he Ghesses in
/he /bimble guide Jube and fue/ asseuAly nozzles.
The fue/ rod acciden/ mduced,s/resses ahich are generally mry sma//, are caused by bendin due lo /k fue/ ossemb/y deflec+ ions darmg e
?
, seismic. acciden+. A summary of /he esmbmed seisme aid L6cn mduced,s/resses, aAich is expre.ssed m /erms of a. peraenkee af a/ low 1sle y
s/ress hmih for /Ae lu' e/ assemsly asjor canposerr/s, is given m 7a~ble 496./- /.
7Ee esirponen/ s/resse,s adiah ine/ade norma / cperasmg s/resses are, cub -
1'
,s/snha//y be/ow /de eslas AsAed a//owab/e hmi/s.
6msequen-Ny -/he s/ruc/ ural designs of 4he fuel i
. pn/u/a/y conponen& are accep/asle andered occid assems/
ihe.
1 svupas plan 4.
~
SNUPPS 6
t 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-erence 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 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.
Reference 3.Balfour,J.
Insert B E 4
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.
490.1-7 Rev.
INSERT B q*
1.
Gesinski, L. and Chi g, D.i " Safety Analysis of the 17 x 17 Fuel Assembly for, Combined Seistric and Loss-of-Coolant Accident,"
WCAP-9236 (Proprietary) and WCAP-8288 (Non-Proprietary),
December 1973.
s the 17 x 17 Op@tTmized Fuel Assembly,"
Beaumont, M. D
- t. al., " Verification Testing and Analyses of
=
2.
WCAP-9401-P-A (Proprietary)
.and WCAP-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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