ML20004B934
| ML20004B934 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/22/1981 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.K.2.13, TASK-TM 718, TAC-45198, NUDOCS 8106010328 | |
| Download: ML20004B934 (2) | |
Text
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TOLEDO EDISON Docket No. 50-346 AcHanoP CaousE u.p,una License No. NPF-3 sc=.r 14131259-5221 Serial No. 718 m
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May 22, 1981 d
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'""[*. J 0 3,N # 7 Director of Nuclear Reactor Regulation
.M Attention:
Mr. John F. Stolz M '- f)
Operating Reactor Branch No. 4
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m Division of Operating Reactors (V-,
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United States Nuclear Regulatory Commission
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Washington, D.C.
20555
Dear Mr. Stolz:
On May 12, 1981, the Chairman of the Babcock & Wilcox Regulatory Response Group, Mr. John J. Mattimoe, submitted a report " Letter Report o_n Reactor n
Vessel Brittle Fracture Concerns in B&W Operating Plants".
Mr. Mattimoe stated in his letter that each licensee owning a Babcock & Wilcox (B&W)
Nuclear Steam Supply System would make a submittal related to this subject.
This letter is Toledo Edison's transmittal related to the Davis-Besse Nuclear Power Station Unit No. 1 (DB-1).
The report provided by Babcock & Wilcox is generic in nature and, as such, has combined conservative assumptions such that all B&W systems are bounded by the results. However, due to certain design features of our facility, the generic report provides such an extremely conservative evaluation that thermal shock of reactor vessel is not a concern requiring any remedial action by DB-1.
Specifically, the design features of interest include:
1.
The absence of longitudinal welds.
DB-1 reactor vessel fabrication included no longitudinal welding. Therefore, the location of longi-tudinal welds is not an issue at DB-1.
2.
Atypical weld material. As a result of previous concerns over impurity content of weld material for reactor vessels, DB-1 was found to not have significant problems in this area.
3.
Shutoff head of the high pressure injection (HPI) system. The HPI pumps have a shutoff head of approximately 1600 psi.
This is well below that of other B&W facilities and therefore minimizes the effect of the HPI pressurization.
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THE TCLEDO ECISCN CCMPANY ECISCN PLAZA 300 MAC!SCN AVENUE TOLECO, CHIO 43652 8100 n1 n32 8
e Docket No. 50-346 License No. NPF-3 Serial No. 718 May 22, 1981 4.
The initiating event. The assumed scenario of the generic report is the complete loss of main feedwater as well as auxiliary feedwater.
DB-1 has a redundant safety grade auxiliary feedwater (AFW) system.
This includes initiation and control. One of the redundant trains also provides independence of A.C. power. Therefore, the initiating event scenario is extremely conservative for DB-1.
5.
The secondary side isolation capability. The steam and feedwater rupture control system (SFRCS) provides a rapid safety grade isolation of the steam generator during loss of main feedwater conditions. The effect of the SFRCS is to maintain secondary side pressure, and limit the maximum reactor coolant system cooldown at DB-1 to one much less severe than evaluated in the generic case.
Toledo Edison feels that the report provided in the May 12, 1981 submittal combined with the above listed features of the DB-1 provide sufficient information to identify that the reactor vessel thermal shock issue is not an immediate safety concern for cur facility.
Very truly yours, i
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/W&
RPC/RFP: lab cc:
DB-1 NRC Resident Inspector J
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