ML19347D140
| ML19347D140 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 02/18/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Jens W DETROIT EDISON CO. |
| References | |
| NUDOCS 8103110211 | |
| Download: ML19347D140 (8) | |
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Docket No. S0-341 D
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Dr. Wayne s'. Jens
- fu Assistant Vne President Engineering & Constmetion The Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226
Dear Dr. Jens:
SUBJECT:
REQUESTS FOR ADDITIONAL INFORMATION IN FEFdiI 2 FSAR As a result of our continuing review of the Final Safety Analysis Repcrt (FSAR) for the Enrico Femi Atomic Power Plant Unit No. 2, we have developed the enclosed requests for additional ~ infomation.
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Please amend your FSAR to comply with the requirements listed in the enclosure.
Our review schedule 1: based on the assu=ption that the additional information will be available for or review by March 10, 1981.
If you cannot meet this date, please infom us within 7 days after receipt of this letter so that we may revise our scheduling. -
Sincerely, 0%
Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
Requests for Additional Information cc w/ encl:
See next page O
810811b80
Dr. Wayne H. Jens Assistant Vice President Engineering & Construction Detroit Edis?n ',ompany FEB18199; 2000 Second Avenue Detroit, Michigan 48226 Eugene S. Thomas, Jr., Esq.
Oavid E. Howell, Esq.
cc:
LeBoeuf, Lamb, Leiby & MacRae 21916 John R 1333 New Hampshire Avenue, N. W.
Hazel Park, Michigan 48030 1
Washington, D. C.
20036 Mr. Bruce Little Peter A. Marquardt, Esq.
Co-Counsel U. S. Nuclear Regulatory Ccanissien The Detroit Edison Ccmpany Resident inspector's Office 2000 Second Avenue 6450 W. Div.e Highway Newocet, Michigan 48165 Detroit, Michigan 48225 Mr. William J. Fahrner Profect Manager - Fermi 2 The Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 Mr. Larry E. Schuerman
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Licensing Engineer - Fermi 2 "stroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 Charles Bechhoefer, Esq., Chairman Atomic Safety & Licensing Board Panel i
U. S. Nuclear Regulatori Commission Washington,-D. C.
20555 Dr. David R. Schink Department of Oceanography
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Texas A & M University 1
College Station, Texas 77840 Mr. Frederick J. Shon Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C.
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ENCLOSURE REQUESTS FOR ADDITIONAL INFORMATION ENRICO FERMI ATOMIC POWER PLANT, UNIT NO. 2 Occket No. 50-347, Requests by the following branches in NRC are included in this erclosure.
Requests and paget are numbered sequentially with respect to pr sviously transmitted requests.
Branch Page No.
Materials Engineering Branch 121-9; 121-10 Reactor Systems Branch 212-53 Core Performance Branch 241-4
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, 121.0 Materials Engineering Branch - Materials Application Section
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121.14 Due to a long history of problems dealing with inoperable and incorrectly installed snutbers, and due to the potential safety signifimance of failed snebbers in safety related systems and components, it is requested that maintenance records for snubbers be documented as follows:
Pre-service Examination A pre-service examination should be made on all snubbers listed in tables 3.7-4a and 3.7-4b of Standard Technical Specifications 3/4.7.9 This exami-nation should be made after snubber installation but not more than six months prior to initial system pre-operational testing, and should as a mimimum verify the following:
(1) There are no visible signs of damage or impaired operabil.ity as a result of storage, handling, or installation.
(2) The snubber location, orientation, position setting, and configuration (attachments, extensions, etc.) are according to design drawings and specifictions.
(3) Snubbers are not seized, frozen or jammed.
(4) Adequate swing clearance is provided to allow snubber movecpy.
(5)
If applicable, fluid is to the recommended level and is not leaking from the snubber system.
(6) Structural connections such r.s pins, fasteners and other connecting.
hardware such as lock nutst tabs, wire, cotter pins are installed correctly.
If the period between the initial pre-servica examination and initial system pre-operational test exceeds six months due to unexpected situations, re-examination of items 1,4, and 5 shall be performed. Snubbers which are installed incorrectly or otherwise fail to meet the above requirements must be repaired or replaced and re-examined in accordance with the above criteria.
Pre-Operational Testing During pre-operational testing, snubber thermal movements for systems whose operating temperature exceeds 250* F.should be verified as follows:
(a) During initial system heatup and cooldoen, at specified temperature intervals for any system which attains operating temperature, vdrify the snubber expected thermal movemat.
(b) For those systems 3hich do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement.
(c) Verify the snubber swing clearance at specified heatup and cooldown intervals. Any discrepencies or inconsistencies shall be evaluated for cause and corrected prior to p.oceeding to the next specified interval.
121-10
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The above described operability program for snubbers should ce included and documented by tr.e pre-service inspection and pre-o::eraticnal test progrees.
The pre-service inspection must be a prerequisite for tre pre-c;eraticnal testing of snubber thermal motion. This test p.'ogram should be specified in Chapter 14 of the FSAR.
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I 212-53 212.0 Reactor Systems Branch 212.180 Review procedure III.20 of SRP 6.3 requires that icng term cooling capability fallowing a LOCA should be adequate in the event of failure of any single active or passive component of tre ECCS. We require the applicant to discuss how leakage from the first isolation valve in an ECCS suction line from the suppression pool durin, post' LOCA long term cooling will be contained. Our concern is drainage of the suppression pool (heat sink) in view of the possible inaccesscbility for repair of a leaking valve due to local contamination.
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241-4 241.0 Reactor Fuels Section, Core Performance Branch 241.6 Since we last worked on your FSAR, several significant changes have taken place that will affect our review of Section 4.2,
" Fuel System Design." The most fundamental change deals with I
the format and content of Section 4.2 as they relate to the Standard Review Plan; the other changes dcai.t th technical i
issues that have arisen recently. All cf these changes are discussed below. The questions (Qls) that were raised earlier on your FSAR are entirely compatible with the SRP requirements, so information that you have already provided is still applicable.
Standard Review Plan The basic fuels sections of the Standard Format (Rev. 3), the Standard Review Plan (Rev.1,1978), and your FSAR are all the same:
4.2.1 Design Bases, 4.2.2 Description,,and Design Drawings, and 4.2.3 Design Evaluation. Unfortshately, 4.2.1 of the Standard Format (and, hence, of your FSAR) does not clearly call for a quantitative (usually numerical) statement of all the design bases as does the Standard Review Plan.
Similarly, the other sections of the Standard Format and your FSAR mix up design bases, design descriptions, and design evaluations, but that information is sorted out clearly in the Standard Review Plan.
Because of improvements in clarity and completeness in this 1978 version of the Standard Review Plan, we will conduct our review ar.d prepare the SER according to the SRP. Our questions, then, will not bo:open-end, but they will simply ask for the residual information called for in the SRP but not present in your FSAR. There are, thus, two options at this stage of the review.
- 0) tion 1 - You could revise Section 4.2 of your FSAR to follow
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tie details of the SRP (remember, the basic organization structure would be unchanged). This would automatically bring l
out all of the information that is needed.
Option 2 - A cross reference could be provided to link each item in the SRP with a paragraph in your FSAR. This method would leava Section 4.2 of your. FSAR in its p esent format.
but might lead'to additional questions since all of the information is not present.
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241-5 We recomend Option 1.
Revision 1 of the SRP, to which we refer, was formally issued more than two years ago and our original questions (Q1s) were written (compatably) during its preparation. Therefore, we do not view this change as either precipitous or disruptive. Furthermore, it is likely that you will have to identify and justify all deviations frem the SRP under the provisions of a proposed rule (Federal Reaister 45, p 67099, October 9,1980) since your SER w'Til be issued affe~r January 1, 1982.
We urge you to provide the information that would be needed to demonstrate compliance with the SRP at your earliest convenience.
To help you anticipate an iminent revision to SRP-4.2, the following coments are provided.
Revision 1 - This revision was issued in October 1978 and contains all of the basic requirements that you need to address.
It will not be changed significantly by the planned revision.
Revision 2 - This revision is planned for April,.,fblemaking 1981 and is the revision alluded to in the notice of propose on SRP compliance.
In SRP-4.2 this revision will (a) add acceptance criteria for mechanical response to seismic and LOCA loads, and (b) make editorial changes largely confined to adding and correcting citations to regulations and regulatory guides that are already addressed in Rev.1. The acceptance criteria for mechanical response were recently implemented as part of the resolution of Unresolved Safety Issue, Task A-2 and _are given in Appendix E of NUREG-0609. Therefore, you can base your FSAR revisions on SRP-4.2 Rev.1 (current version) plus Appendix E of NUREG-0609, and last-minute changes in referencing can be made in April prior to your submittal of the additional' fuel-related information.
Recent Technical Issues Following is a list of current technical issues that should be given special attention in your ESAR.
For BWRs, recent SER " outstanding issues" include:
1.
Supplemental ECCS calculations with NUREG-0630 models.
2.
Periodic channel box deflection tests.
3.
Combined seismic LOCA loads analysis.
4.
Fission gas release analysis at high burnup.
Other. issues requiring special attention include:
5.
Pellet / Cladding Interaction (PCI).
6.
Water rod end-plug wear.
7.
Waterside corrosion.
8.
Rod bowing.
'9.
Control Blade stress corrosion cracking.
- 10. High burnup.
- 11. Fuel assembly design shoulder gap analysis'.
12.
_ Fnd-of-life fuel rod. internal pressure analysis.
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