ML17209A857
| ML17209A857 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 03/18/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| References | |
| NUDOCS 8103240749 | |
| Download: ML17209A857 (11) | |
Text
MAR 18 1981 Docket No./ 50=389~
Florida Power 8 Light Company ATTN:
Dr. Robert E. Uhrig, Vice President Advanced Systems 8 Technology P. 0.
Box 529100 Miami, Florida 33152
- Dist, I,Docket File LB¹1 Rdg DEisenhut BJYoungblood VNerses MRushbrook RTedesco SHanauer RVollmer DPowers LKopps TMurley DRoss RHartfield, MPA OELD OIE (3) bcc:
TERA NRC/PDR L/PDR NSIC TIC ACRS (16)
Dear Dr. Uhrig:
Subject:
St. Lucie Plant, Unit 2 FSAR - Request for Additional Information As a result of our review of your application for an operating license; we find that we need additional information regarding the St. Lucie Plant, Unit 2 FSAR.
The specific information required is as a result of the Reactor Fuels and Reactor Physics Section of the Core Performance Branch's review and is
'isted in the Enclosure.
Responses to the enclosed request should be submitted by May 1, 1981.
If you cannot meet this date, please inform us within seven days after receipt of this letter of the date you plan to submit your responses.
Please contact Mr. Nerses, St Lucie 2 Licensing Project Manager, if you desire any discussion or clarification of the enclosed report.
Sincerely, QHgmg ~~.~
Rtdvee' T8d8300 Robert L.,Tedesco, Assistant Director for Licensing Division of Licensing
Enclosure:
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Dr. Robert E. Uhrig, Vice President Advanced Systems and Technology Fl'orida Power 5 Light Company P. 0.
Box 529100
'iami, Florida 3~152 CC:
Harold F. Reis, Esq.
Lowenstein,
- Newman, Reis', Axelrad 5 Toll 1025 Connecticut
- Avenue, N: M.
Mashington, D.
C.
20036 Norman A. Coll, Esq.
HcCarthy, Steel, Hectory 5 Davis 14th Floor, First National Bank Bui lding-Niami, Florida 33131.
Hr. Hartin H. Hodder 1131 N.
E.
86 Street Niami,'lorida 33138 Or; David L. Hetrick Professor of Nuclear Engineering The Uni vers ity of Arizona Tucson, Arizona 8572-1 Dr. Frank F.
- Hooper, School of Natural Resources University of Michigan Ann Arbor, Nichigan 48104 Res.ident Inspector St. Lucie Nuclear Power Station c/o U.
S. Nuclear Regulatory Commission,.
P. 0.
Box 400 Jensen. Beach, Florida 33452 Edward Luton, Esq.,
Chariman Atomic Safety 5'icensing
,US Nuclear Regulatory Commission
~
Mashingtpn, DC 20555
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Enclosure ST.,UCIEE, UNIT 2, FSAR Reactor Fuels Section, Core Performance Branch Several significant changes have taken place that will affect our review of Section 4.2, "Fuel System Design."
The most fundamental change deals with the format and content of Section 4.2 as they relate to the Standard Review Plan; the other changes deal with technical issues that have arisen recently.
All of these changes are discussed below.
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.
Unfortunately, 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 and prepare the SER according to the SRP.
Our questions, then, will not be open-ended, 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 thi s stage of the review.
Option 1 - You could revise Section 4.2 of your FSAR to follow the
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s of the SRP (remember, the basic organization structure would be unchanged).
This would automatically bring out all of the information that is needed.
Option 2 - A cross reference could be provided to link each item iin lie eRP with a paragraph in your FSAR.
This method would leave Section 4.2 of your FSAR in its present format, but might lead to additional questions since all of the information is not present.
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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 imminent revision to SRP-4.2, the following corments 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 1981 and is the revision alluded to in the notice of proposed rulemaking 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.
r Recent Technical Issues The following is a list of current technical issues that have frequently been noted as outstanding issues in recent SERs and that should be given special attention in your FSAR.
1.
Supplemental ECCS analysis with NUREG-0630.
2.
Combined seismic and LOCA loads analysis.
3.
Enhanced fission gas release analysis at high burnups.
4.
Fuel rod bowing analysis.
'. " Fuel assembly control rod guide tube wear analysis.
6.
Fuel assembly design shoulder gap analysis.
7.
End-of-life fuel rod internal pressure analysis.
We recommend Option 1.
Revision 1 of the SRP, to which we
- refer, was formally issued, more than two years, ago.,There-
- fore, we do not view this change as either precipitous'r disruptive.
Furthermore, it is likely that you will have to identify and justify all deviations from the SRP under the provisions of a proposed rule (Federal Re ister 45, p 67099, October 9, 1980) since your ER wi be issued after January 1, 1982.
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Reactor Ph sics Section, Core Performance Branch There appears to be an incons.istency, or at least confusion, between the values of CEA worth available for trip as given in Section 4.3.2.5, Table 4.3-5, Table 4.3-7, and Section 15.0.3.2.3.
Please clarify this item.
In the past, CE has performed analyses of various control rod mis-operation events including single part length CEA drop and part length CEA subgroup drop.
The results of these analyses have always met the NRC acceptance guidelines that critical heat flux should not be exceeded and centerline fuel melting should not occur.
A new
- event, the part length CEA group drop is now presented in the St.
Lucie 2 FSAR as the most limiting initiating event in the uncontrolled positive reactivity insertion event group.
CE categorizes it as an infrequent event and allows a very small amount of failed fuel.
In view of the fact that this is a
new event presentation and 0.5 per cent of the fuel pins do experience DNB, present additional information describing the failure or failures which would lead to a part length CEA group drop as well as the rationale used to place it in the in-frequent event probability group category.
. GeneralDesign Criterion 25. for. nuclear power plants requll es th the protection system be. designed. to, assure. that specified -acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal of control rods; In view of this, justify the. categorization of events such as the sequential CEA withdrawal as an infrequent event in which a, small amount of fuel failure is accepted.
The inadvertent loading and operation of a fuel assembly iri an im-proper position does not appear in the Event Type/Frequency Matrix of Table 15.0-2, Provide additional information describing and justifying the probability group category CE considers this event to fall within.
Discuss the effect of loading one or more fuel assemblies into improper.locations or misorientation (particularly in the case of reload cores} of fuel assemblies on nuclear design para-
- meters, power distribution, and local power density as well as the radiological implications to justify that it is not a limiting event.
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