ML19320A188
| ML19320A188 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/05/1972 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Phillips J ARKANSAS POWER & LIGHT CO. |
| References | |
| NUDOCS 8004210492 | |
| Download: ML19320A188 (3) | |
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Docket No. 50-313 g
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D Mr. J. D. Phillipe i
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Vice Presidut & Chief Engineer D
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,v, Sixth and Pine Streets Pine Biaff, Arkansas 71601 Deer Mr. Phitilper On the basis of our ese*4==1mp review of the Final Safety Analysis Report for the Arkansas Deelear One - Unit No.1, we find that we need additiemal j
information to semplete our evaluaties. The speciffe informaties required l
is listad la the analeeure.
In order to maintain our 11m==m4=3 review schedule we will need a sempletely I
adequete reopease by September 15, 1972. Your representative has heard these requests and indicated that respease by this~date is reassemble. Please infers us withis saves (7) days after reseipt of this letter of your esafirmaties of the aahad=1o.
If yes sassot meet our spesified date er if your reply is met fully reopensive to our requests it is highly likely that the overall schedule for completing the licensing review for this project will have to be extended. Sinee reassignment of the staff's efforts will require sempletian of the new assignment prior to returning to this project, the estaat of extension will moet likely be greater them the extent of delay la your response.
Sincerely.
Original signed by R. C. DeYoung '
j R. C. DeYoung, Assistant Direeter for Pressurized Water Reactors Direstorate of Licensing E=ataeure:
Request for Additional Informaties
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2-SEP 5 1972 Mr. J. D. Phillips cc w/ enc 1:.
Horace Jewell, Esq.
House, Holms & Jewell 1550 Tower Building Little Rock, Arkansas 72201 DISTRIBUTION AEC PDR Local PDR Docket RP Reading PWR-4 Reading SHHanauer RSBoyd RCDeYoung DSKovholt FSchroeder RRMaccary
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REOUEST FOR ADDITIONAL INFORMATION ARKANSAS' POWER AND' LIGHT COMPANY DOCKET NO. 50-313 5.0 STRUCTURES 5.83 General Design Criterion No. 57 (GDC 57) in 10 CFR '50, Appendix A (published in the Federal Register on July 7, 1971) states that a simple check valve may not be used as an automatic isolation valve. Describe the type of check valve that is used as an isolation valve in the main feedwater line to each steam generator and explain how its use complies with GDC 57.
5.84 With respect to isolation valve arrangement No. 8 (see Figure 5-6) for the instrument and air supply lines (penetration Nos. 43 and
- 46) explain the basis for the following actions:
a.
a manual (locked closed) valve is used inside the reactor building.
b.
the lines are capped.
c.
the valves are classified as Type IV, and d.
a simple check valve is used outside of the reactor building.
Figure 9-14 also shows the isolation valving arrangements for these lines. Are they shown correctly?
5.85 There appears to be a discrepancy between the valve arrangement No. 2 schematic (see Figure 5-6) and the isolation valving description for penetration Nos. 31 and 32 (see Table 5-1).
Please make the necessary corrections to resolve this discrepancy.
5.86 With respect to the penetration numbers in 5-1, identify the FSAR system schematics which show the penetrations and the corresponding isolation valve arrangements.
5.87 Identify the penetration numbers in Table 5-1 that correspond to the reactor building purge system supply and exhaust ducts.
5.88 Discuss the design provisions to assure the integrity of the main steam isolation valves under tl.a dynamic forces resulting from inadvertent valve closure during reactor operation.
5.89 Figure 5-7 refers to Figure 9-1(b) for the legend nomenclature.
Please provide Figure 9-1(b).
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