ML20207M103
| ML20207M103 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/12/1988 |
| From: | Roberts J NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Mcconaghy W NUTEC, INC. |
| References | |
| REF-PROJ-M-49 NUDOCS 8810180112 | |
| Download: ML20207M103 (4) | |
Text
-
00T 131988 Project No. M-49 Docket No. 72-4
- i NUTECH, Inc.
ATTN: Mr. William J. McConaghy.
Vice President Waste Management Business Group 145 Martinvale Lane San Jose, CA 95119
Dear Mr. McConaghy:
Per our telephone conversation of October 11, 1988, this is a request for information to complete our review of NUTECH's Topical Report (TR), NUH-002, Revision 1 (see enclosure).
If you have any questions on this matter, call me l
at (301) 492-0608, f
Sincerely,-
Ot W Signed sy:
John P. Roberts, Section Leader j
Irradiated Fuel Section Fuel Cycle Safety Branch
Enclosure:
Request for Information cc: Mr. H. B. Tucker, Duke Power Com any
[
Service List i
i i
I
~
DISTRIBUTION: w/ enclosure l
o Wetet No. A49"4" 72-47""PDR A' LPORO NRC File"Centir7 IMAF R/F IMSB R/F l
N c)
I o
NHSS R/F' " LCRouse' ~~JPRoberts
'FLBrown HPastis MRood AHodgdon f
o) 0)
bec:
JStokley h
.:0
,,n 0
FC :
IN i
- MIF
\\
- )RM
(.V.......... ;
i P
........X LS NAME: JQtoberts:lg rown gcai DATE:
10//l/88 041/88 l
l 1
OFFICIAL RECORD COPY
0 l
o Mr. H. B. Tucker Oconee Nuclear Station Duke Power Company Units Nos. 1, 2 and 3 cc:
Mr. A. V. Carr, Esq.
Mr. Paul Guill Duke Power Company Dukt Power Company Pre
- Office Box 33189 P. O. Box 33189 422 South Church Street 4;J South Church Street Charlotte, North Carolina 28242 Charlotte, North Carolina 28242 J. Michael McGarry, III, Esq.
Bishop, Liberman, Cook, Purcell & Reynolds 1200 Seventeenth Strcet, N.W.
Washington, D.C.
20036 Mr. Robert 8. Borsus Babcock & Wilcox Nuclear Power Generation Division Suite 525 1700 Rockville Pike Rockville, Maryland 20852 Manager, LIS NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Senior Resident Inspector U.S. Nuclear Regulatory Comission Route 2, Box 610 Seneca, South Carolina 29678 Regional Administrator, Region II U.S. Nuclear Regulatory Comission 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 3:'323 Mr. Heyward G. Shealy Chief Bureau of Radiological Health South Carolina 06partmint of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 Offico of Intergovernmental Relations 116 West Jones Street Raleigh, North Carolina 27603 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621
OCT 121988 i
REQUESY FOR INFORMATION 1.
Table 8.1 Provide methodology for determination of load due to creep.
2.
Provide additional information on the HSH sleeve design.
3.
Provide additional discussion on thermal movements in the longitudinc1 direction for the DSC support rails.
4.
Provide reinforcing requirements for HSM front, rear walls.
5.
Clarify that door loads are due to "normal" operation and delete reference to drop.
6.
Cask drop analysis - peak or plateau acceleration - confirm design basis.
7 HSM STRUDL analysis - Run I, p. 9 Loading 3F - Provide description of teraperature gradient methodology.
8.
DSC STRUDL analysis - p. 5 - Explain loading combination in respect to Table 8.2-12, Accident Condition, and Table 8.1-9, Normal DSC Handling Loads.
9.
DSC support analysis - Acceleration values appear inconsistent with
- p. 8.2-19 of the TR.
10.
The assurance of nuclear criticality safety of the DSC design depends in part on the burnup of the fuel. A relationship between irradiated fuel reactivity and equivalent fresh fuel enrichment was determined through a series of detailed criticality calculations which are referenced in the Topical Report. ^Although there is no doubt that the reactivity of irradiated fuel is reduced from that of unirradiated fuel, there are a nurrber of questions regarding the accuracy with which the reactivity of irradiated fuel can be estimated. An alternative to providing assurance of nuclear criticality safety which avoids these questions is to consider loading tha DSC in a high neutron absorbing solution and to base the criticality safety analysis on fresh fuel enrichment.
If nuclear criticality safety can be assured during loading of fresh fuel into the DSC and subsequent drying 03erations, and if assurance can be provided to preclude reflooding of the )SC, there appears to be a basis for accepting the design.
Therefore, what is the boron concentration required to provide assurance of nuclear criticality safety during the loading of unirradiated 4 percent enriched fuel into the DSC7 Can you identify other design or operational issues which preclude this option? Describe procedures which will assure maintenance of the required boron concentration in solution in the DSC during fuel loading and unloading.
i 2
OCT 121988
- 11. There are several questions or uncertainties centering on the variation in burnup along the length of the irradiated fuel (axial burnup dependence).
For a given irradiated fuel assembly, how accurately is the isotopic composition of fuel known? What are the uncertainties in fuel composition, and how do these uncertainties translate to uncertainties in reactivity of a finite array of irradiated fuel assemblies? If the isotopic composition of each fuel assembly is known with precision, what errors result from the i
approximate models of the axial variation in isotopic composition? Can it be demonstrated that the model selected is conservative (in the sense of reactivity)?
i Finally, there is the question regarding the availability or applicability i
of benchmarks for reactivity calculations with irradiated fuel.
Please defend that the benchmarks used for validation are representative of the t
irradiated fuel asserblies to be placed in the DSC, and that the method bias determined from these benchmarks can be appropriately used in the criticality safety analysis.
- 12. Thestatementgivenonthetopofpage1.3-3gftheTopicalReport indicates that the maximum temperature of 570 C during short-term operational and postulated accident conditions will not affect fuel cladding integrity.
The references cited indicate that cladding failure was not found. However, cladding creep can and does occur at this temperature. Please provide a response that confirms that cladding creep will not adversely affect cladding integrity.
[
l i
t I
i