ML14183A090
| ML14183A090 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/20/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14183A089 | List: |
| References | |
| NUDOCS 8702030386 | |
| Download: ML14183A090 (4) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 112 TO-FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER AND LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 I. INTRODUCTION By letter to L. S. Rubenstein (NRC) from A. B. Cutter (CP&L) dated October 13, 1986, Carolina Power and Light Company (the licensee) requested revision of the Technical Specifications for H. B. Robinson Unit 2. The changes concern the storage and handling of fuel with an increased enrichment to 3.9 w/o of U-235.
At our request, the licensee revised the content of one of the changes to specifically designate acceptable fuel storage locations.in the new fuel storage vault. This was done in a letter to L. S. Rubenstein (NRC) from A. B. Cutter dated December 11, 1986. In support of the requested chances the licensee provided the following Exxon reports with his initial submittal:
- 1) "Final Report, Criticality Safety Analysis, H. B. Robinson New Fuel Storage Vault with 4.2 Percent Enriched 15 x 15 Fuel Assemblies," August 1986, Exxon Report No. XN-NF-86-100.
- 2) "Final Report,.Criticality Safety Analysis, H. B. Robinson Spent Fuel Storage Rack (Unpoisoned, Low Density) with 4.2 Percent Enriched 15 x 15 Fuel Assemblies," August 1986, Exxon Report No. XN-NF-86-107.
II. EVALUATION The analyses contained in the above Exxon Reports were performed for an enrichment of 4.2 w/o U-235. The value in the proposed Technical Specification change, however, is a more limiting enrichment of 3.9 w/o, which was previously submitted and approved for the high-density, poisoned spent fuel racks. The scope of the requested changes is limited to the handling and storage of more highly enriched new fuel elements. Operati-on with the increased enrichment will be addressed in subsequent reload analyses.
The specific proposed Technical Specification changes are:
6_702030386 670120 PDR ABORK 05000261 P
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- 1. Page 5.3.7, Section 5.3.13 - Reactor Core Desian Specification. The enrichment of the reload fuel is changed from 3.5 w/o to 3.9 w/o.
- 2. Pages 5.4-1 and 2, Section 5.4 -
Fuel Storage Design Specification. This section has been rewritten and reformatted to include mentionof the nominal 21-inch center-to-center spacing of the new fuel storage racks, allow storage of fuel with a maximum axial plane enrichment of 3.9 w/o of U-235 in both the new and spent storage racks, the design k for worst accident conditions, and boron concentration for the spent u 1 pit during fuel handling.
- 3. Page 4.1-10, Table 4.1 Frequency for Sampling Tests. Changes include correction of a typographical error and specification of a sampling requirement prior to new fuel movement in the spent fuel storage pit.
The analysis supporting the proposed changes for the new fuel storage vault is contained in Exxon Topical Report XN-NF-86-100 (referenced above).
The report describes the model used for the new storage vault analysis, the assumed inout parameter values, the methods used for the analysis, and presents some of the results of the methods verification.
The values of fuel parameters selected for the analysis were chosen in the conservative direction. Thus, fuel pellet density was chosen to be slightly greater than the design value, the fuel pellet dish volume was neglected, and the fuel stack length was taken as 144 inches whereas the fuel design stack length is 132 inches (enriched). There is a minimum of 12 inches of natural uranium in all fuel rods, and some have a larger section of natural uranium.
Most importantly, no Gd 0 content was assumed in the model fuel.
Nominal values were used for remafning fuel geometry and composition parameters.
Because of the conservative assumptions indicated above, we conclude the fuel model used in the calculations is acceptable.
Conservative assumptions were made concerning the fuel storage rack, geometry and composition. The modeled storage rack pitch is 20.857 inches, whereas the actual rack value is nominally 21 inches. The vault was reflected within 30 cm of concrete at the 4 walls, the floor and at 14 feet above the ceilino.
All rack materials of construction were neglected in the model. Thus the model is conservative in geometry, reflection and neutron absorption effects, and is therefore acceptable.
The calculation methods use KENO-IV or XSDRNPM for k. and k calculations. Suitable cross section libraries wereied. TA report presents the results of comparison of the criticality factors for four sets of critical experiments. The results show good agreement with the measured criticalities. We therefore conclude the calculation model used is acceptable.
The calculation of the actual new fuel vault criticality with fuel bundles modeled in all 105 locations indicated that the criterion of k
=0.98 with optimum moderation of the fuel rack would not be met. This crf[rion and one requiring k to be =0.95 for the rack fully flooded (or for the worst credible acilaent) must be met according to the Standard Review Plan, NUREG-0800. In view of unacceptability of the criticality of the new fuel storage racks when fully loaded, the report presents the results of four
3 alternative loadings of fuel in the rack with empty locations interspersed between fuel locations. These allow loadino of 69-73 fuel bundles.
The alternative loading patterns all show an acceptable k for optimum moderation. At our request the licensee amended his orig l submittal to specify which fuel locations must be locked out in order to ensure conformance with the optimum moderation criticality requirements.. With these changes, we find the proposed changesto Specification 5.4.1, New Fuel Storage Racks, acceptable. The Specification also defines the maximum enrichment to be stored as 3.9 w/o U-235. This is conservative because the analysis shows acceptable criticality results with an enrichment of 4.2 w/o U-235.
The change to Specification 5.3.1.3 (Ttem 1 above) is acceptable because it merely indicates that fuel enrichments up to 3.9 w/o U-235 can be used in the core design. Determination of the acceptability of an actual core design must be verified in the calculation of physics parameters and accident analysis in the reload design evaluation.
Topical Report XN-NF-86-107 (referenced above) provides the results of a criticality analysis of the spent fuel pit using fuel with a maximum enrichment of 4.2 w/o U-235. The conservative assumptions concerning fuel and rack geometry described above for the new fuel vault calculations were also used for the spent fuel pit calculations except the more conservative.
assumption of an infinite array of infinite length assemblies was used for the spent fuel pit.calculations. The same computer code methods were used. The results indicate that keff =0.95 for the worst condition of the pit fully flooded with pure water. A spectrum of accidents was evaluated which show that the above case is limiting except for closer edge to edge fuel assembly placement during a fuel handling accident. The analysis shows that a minimum boron concentration of 500 ppm during fuel handling will prevent exceeding the k
criterion of =0.95. The proposed change to Specification 5.4.3, BORON C 6ENTRATION-SPENT FUEL STORAGE PIT, which requires a boron concentration of
=1500 ppm during refueling operations or new fuel movement in the spent fuel storage pit is much more conservative than the value used in the analysis and is, therefore, acceptable.
III.
SUMMARY
We conclude that the Technical Specification changes proposed by CP&L for H. B.
Robinson Unit 2 above are acceptable. The change in item 1 above is acceptable because the cycle specific-reload analysis will demonstrate the safety of the actual reload enrichment. The changes in item 2 above are acceptable because.
they conform with our requirements for fuel storage criticality and were calculated as discussed above with conservative fuel parameter and storage rack assumptions and with acceptable computer models.
The changes in item 3 above are acceptable because the first is an administrative change (correction o* a typographical error) and the other represents a suitable requirement to prevent an approach to criticality when moving fuel in the spent fuel pit.
4 IV.
ENVIRONMENTAL CONSIDERATION This amendment involves a change in the installation or use of a facility component located within the restricted area as defined in 10 CFR.Part 20.
The staff has determined that the amendment involves no significartt increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant ha7ards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
V. CONCLUSION We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: January 20, 1987 Principal Contributors:
M. Dunenfeld