ML18227C737
| ML18227C737 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 08/18/1976 |
| From: | Lear G Office of Nuclear Reactor Regulation |
| To: | Robert E. Uhrig Florida Power & Light Co |
| References | |
| Download: ML18227C737 (7) | |
Text
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/QG 0~4 Docket Nos. 50-250 and 50-251 Florida Power and Light Company ATTN:
Dr. Robert E. Uhrfg Vice President P. 0.
Box 013100 ttfamf, Florida 33101 Gentlemen:
Distribution NRC PDR Local PDR ORB83 File KRGoller TJCarter OELD OI8IE (3)
GLear DElliott
- CParrish, DEisenhut TBAbernathy, DTIE JRBuchanan ACRS (16)
DRoss By letters dated January 28, 1976 and April 30, 1976 you subnftted proposed changes to the Technical Specifications which relate to your proposed modification of the spent fuel storage facility for each unit.
Based on our continuing review of your submittals, we have determined that we need additional information.
Please provide the items of additional information listed in the enclosure.
In many cases the enclosed request for information refers to questions which were included in our request for fnformatfon dated April 6. 1976 and which were not adequately discussed in your submittal of April 30, 1976.
In order'for us to continue our review in a timely manner please submit the requested information as soon as possible.
If you have any questions regarding our request, please contact us.
Sincerely, Encl osur e:
Request for Additional Information Qtlg4>All s~,'%IPd big George Lear, Chief Operating Reactors Branch 83 Division of Operating Reactors cc:
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Docket Nos. 50-250 and 50-251 Florida Power and Light Company ATTN:
Dr. Robert E. Uhrig Vice President P. 0. Box 013100 Nami, Florida 33101 Gentlemen:
Distribution
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NRC PDR Local PDR ORB¹3 File KRGoller TJCarter OELD OI8E (5)
- GLear, DE1liott CParrish DEisenhut TBAbernathy, DTIE JRBuchanan ACRS (16)
DRoss j)
By letters dated January 28, 1976 and April 30, 1976 you submitted proposed changes to.the Technical Specifications which relate to your proposed modification of the spent fuel storage facility for each unit.
Based on our continuing review of your submittals, we have determined that we need additional information.
Please provide the items of additional information listed in the enclosure.
In many cases the enclosed request for information refers to questions which were included in our request for information dated April 6, 1976 and which were not adequately discussed in your submittal of April 30, 1976.
In order for us to continue our review in a timely manner please submit the requested information as soon as possible.
If you have any questi'ons regarding our request, please contact us.
Sincerely, Qg$ ~>'IIs~ak wI~skwlrk WJ'nclosure:
Request for Additional Information George Lear, Chief Operating Reactors Branch ¹3 Division of Operating Reactors cc:
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, P REqUEST FOR ADDITIONAL INFORMATION TURKEY POINT UNITS 3 AND 4 SPENT FUEL POOL -
INCREASED STORAGE CAPACITY DOCKET NOS. 50-250 5, 251 The response to question 18* indicates that the faulted limits of the ASME Boiler and Pressure Vessel Code,Section III, Division I Subsection NF will be used for the load combination which includes the effects of the safe shutdown earthquake (SSE).
This will permit stress limits approximately 20 percent greater than those permitted in paragraph II.5 of Standard Review Plan 3.8.4.
It is the staff's position that the structural acceptance criteria be similar to those given in paragraph II.5 of the Standard Review Plan.
The yield stress values recommended by Section III of the ASME Code should be used in conjunction with the AISC requirements.
Describe the design procedure and permissible stresses for the stainless steel welded joints.
The response to question 21 with regard to the design of the fuel storage racks, is not acceptable.
It is the staff's position that the design and analysis of the new rack system be performed using either the existing input paramaters including the old damping values or new parameters in accordance with Regulatory Guides 1.60 and 1.61.
The use of existing input with the damping values in Regulatory Guide 1.61 is not acceptable.
Furthermore, seismic excitation along three orthogonal directions should be imposed simultaneously for the design of the new rack system.
Section 3.2" indicates that the time history method will be used in the design of the racks.
Please provide a comparison of the response spectra derived from the time history and the design response spectra for the damping values that are used in the design.
With regard to question 22, only the mathematical model of the fuel pool has been provided.
Please provide the information requested for the storage rack and fuel assembly system.
The effects of
- gaps, sloshing water, and increase of effective mass and damping due to submergence in water should be quantified.
With regard to question 23, if the racks are laterally supported from the fuel pool walls, describe how these loads were included in the re-analysis of the pool structure.
It is not clear if the racks will be anchored to the base slab of the pool. If so, describe the design of the anchorage and the materials used.
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2-
'8.
"9.
'10.
11.-
'12.
Section 3.2 indicates that the time history analysis is performed with the SHOCK computer code.-
Describe the procedure used to validate this program.
Acceptable procedures for validation are delineated in paragraph II.4.e of Standard Review Plan 3.8.1.
The response to question 31{a) is inadequate.
Please provide the radionuclides concentrations as specified in the question.
Section 4.4 states that the spent fuel pit area is a radiation zone 1 classification with a maximum dose rate of 0.5 mrem/hr.
Section 5.6 states that "the measured dose rate to personnel from the radionuclides in the spent fuel pool water is only a few mrem/hr."
Please clarify this anomoly.
Also, provide the dose rates that
=have been measured in the areas specified in question 31(b),
namely
~the center and edge of the pool.
Section 5.6 does not respond to questions 31(e) and (g) in an acceptable manner.
= Please address your estimate of'he increase in the annual man-rem burden 'from all operations in the spent fuel pool as a result of the pool midification.
Include those exposures from more frequent changing of the demineralizer resin and filter cartridges.
Section 5.6 'states that "the radiation levels along the sides of the pool and over the center of the pool are essentially the same.
Therefore there has been on special build-up of crud around the sides of the pool..."
The conclusions drawn with respect to the measurements appear to be incorrect.
The dose rate at the edge of the pool, if there were clean pool walls {i.e., no crud build-up), should be less than the dose rate at the pool center based on source-detector geometrical considerations (i.e.,
a cylindrical source at the center and half-cylinder at the edge).
Please pro-vide mathematical models and calculations to justify your conclu-sions with respect to crud build-up around the sides.of the pool.
13.
Please provide the maximum and average volume of water in the
- spent fuel pool as requested in question 27(a).
- 14. 'sing realistic initial conditions that will exist for the fuel pool and the fuel assemblies within the pool, please show that the calculated neutron multiplication in the pool will be less than 0.95 under all accident conditions (including the cask drop and tip in
- the fuel pool accident).
- guestion and section numbers refer. to FPL submittal of April 30,
- 1976, Spent Fuel Storage Facility Modification Safety Analysis Report-Rev. l.