ML20111B401

Safety Evaluation Supporting Amend 109 to License NPF-30

ML20111B401
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Site:	Callaway
Issue date:	04/30/1996
From:	NRC (Affiliation Not Assigned)
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p aeogk UNITED STATES NUCLEAR REGULATORY COMMISSION f

WASHINGTON. D.C. 20066 0001

\\ *****/ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.109 TO FACILITY OPERATING LICENSE NO. NPF-30 UNION ELECTRIC COMPANY Cr'_AWAY PLANT. UNIT 1 DOCKET NO. 50-483

1.0 INTRODUCTION

By letter dated September 6, 1995, as supplemented by letters dated January 30, 1996, March 27, 1996, and April 2, 1996, Union Electric Company (UE), requested changes to the Technical Specifications (Appendix A to Facility Operating License No. NPF-30) for the Callaway Plant, Unit 1.

The proposed amendment would allow for the storage of fuel with an enrichment not to exceed a nominal 5.0 weight percent (w/o) U-235, subject to certain integral fuel burnable absorber (IFBA) requirements or discharge exposures, in the spent fuel pool storage racks.

Plant operation using the higher enriched fuel will be demonstrated to be acceptable by a cycle specific reload safety evaluation performed prior to each fuel loading. TS 5.3.1, TS 5.6.1.1 and TS Figure 3.9-1 would be revised to incorporate the above changes.

The January 30, 1996, March 27, 1996, and April 2,1996, supplemental letters provided additional clarifying information and did not change the original no significant hazards consideration determination published in the Federal Reaister on November 8, 1995 (61 FR 56372).

2.0 EVALUATION 2.1 Criticality Review The Callaway spent fuel pool (SFP) is divided into two separate and distinct regions.

Region I contains unpoisoned racks and is designed to accommodate fresh (unirradiated) fuel assemblies in a two-out-of-four checkerboard array.

Therefore, from a criticality viewpoint, any type of fuel from the Callaway core can be stored in Region 1.

Region 2 is designed to accommodate only irradiated fuel assemblies which have attained sufficient burnup.

The analysis of the reactivity effects of fuel storage in the SFP storage racks was performed with the three-dimensional multi-group Monte Carlo computer code, KENO-Sa, using neutron cross sections generated by the NITAWL code package from the 27 energy group SCALE data library. The two-dimensional transport theory code, CASMO-3, was also used to determine a reference k,,

which can be used as an alternate approach for determining the acceptability of a fuel assembly for storage in the Region 1 racks.

These codes are widely used for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments. These experiments simulate the Callaway fuel storage racks as realistically as possible with 9605140402 460430 PDR ADOCK 05000483 P

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respect to parameters important to reactivity such as enrichment and assembly spacing. The intercomparison between two independent methods of analysis (KENO-Sa and CASMO-3) also provides an acceptable technique for validating calculational methods for nuclear criticality safety. To minimize the statistical uncertainty of the KENO-Sa reactivity calculations, a sufficient number of neutron histories were accumulated in each calculation to assure convergence of KENO-Sa reactivity calculations. Based on the above, the staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the Callaway spent fuel storage racks with a high degree of confidence.

The spent fuel racks are normally fully flooded by water borated to at least 2000 ppa of boron and verified weekly by plant procedures. However, to meet the criterion stated in Section 9.1.2 of the WRC Standard Review Plan (SRP),

k,'ticipated reactivity and flooded with unborated water.# must not exceed 0.95 with an The maximum calculated reactivity must include a margin for uncertainties in.eactivity calculations and in manufacturing tolerances such that the true k,,, will not exceed 0.95 at a 95/95 probability / confidence level.

The spent fuel storage racks in Region 1 were analyzed for fresh Westinghouse 17x17 Vantage-5 (V-5) fuel assemblies enriched to 4.1 w/o U-235 with no IFBA rods and moderated by pure water at 68 degrees F with a density of 1.0 gm/cc, which results in the highest reactivity.

For the nominal storage cell design in Region 1, uncertainties due to tolerances in fuel enrichment and density, storage cell spacing, and stainless steel thickness were accounted for. These uncertainties were appropriately determined at the 95/95 probability /

confidence level.

In addition, calculational and methodology biases and uncertainties due to the KENO-Sa statistics and benchmarking were included.

The resulting spent fuel rack k,, was 0.9481, including biases and uncertainties at the 95/95 level. Tnis meets the NRC acceptance criterion of 0.95 and is, therefore, acceptable.

To enable the storage of fuel assemblies with nominal enrichments greater than 4.1 w/o U-235, the concept of reactivity equivalencing was used.

In this technique, which has been previously approved by the NRC, credit is taken for the reactivity decrease due to the IFBA material coated on the outside of the U0, pellet.

Based on these calculations, 21 IFBA rods are required to maintain k,,, no greater than 0.95 for fuel initially enriched to 5.0 w/o U-235.

Since current Westinghouse IFBA patterns are limited to 16 or 32 rods per assembly, the actual limit for assemblies with enrichments greater than 4.1 w/o U-235 and less than 4.8 w/o U-235 is 16 IFBA rods, and is 32 IFBA rods for assemblies with enrichments greater than 4.8 w/o U-235. The calculations included uncertainties on the boron-10 (B-10) loading tolerance, the IFBA stack length tolerance and IFBA rod position. Although the boron concentration in the IFBA rods decreases with fuel depletion, calculations have shown that for the number of IFBA rods considered in this analysis, the fuel assembly reactivity decreases more rapidly. Therefore, the reactivity equivalencing calculations were performed at zero burnup, resulting in the maximum fuel rack reactivity.

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4 As an alternative method for determining the acceptability of fuel storage in Region 1, a reference k calculation was performed using CASMO-3. The calculation used the nominal 4.10 w/o V-5 fuel assembly with no burnable absorbers in the Callaway reactor geometry at a temperature of 68 degrees F.

The resulting k,, was 1.480 and included the 1 percent reactivity bias to account for calculational uncertainties. Thus, fuel assemblies which are to be stored in the Callaway Region I spent fual racks must either meet the initial enrichment versus IFBA requirements previously described, or have a reference k,, less thar, or equal ';o 1.480, to ensure that the final k,,, of the l

Callaway Region I spent fuel racks will be no greater than 0.95.

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Most abnormal storage conditions will not result in an increase in the k,,,

j of the racks.

However, it is possibla to postulate events, such as the i

misloading of an assembly with an enrichment and IFBA combination outside of j

the acceptabic area, which could lead to tn increase in reactivity for Region l

1.

However, for such events, credit may be taken for the presence of i

4 approximately 2000 ppe of boron in the pool water since the staff does not require the assumption of two unlikely, independent, concurrent events to I

ensure protection against a critiMity accident (Double Contingency 4

Principle).

The reduction in k

' ised by the boron more than offsets ti.:

reactivity addition caused by c,r,e,diole accidents. Therefore, the staff 4

j criterion of k,,, no greater than 0.95 for any postulated accident is met.

a Previously approved analyses performed for the Rcgion 2 racks showed that acceptable criticality limits are maintained when storing fuel enriched to a 3

maximum of 5.0 w/o U-235, providad that the fuel burnups meet the prescribed limits. However, due to thermal-hydraulic constraints, fuel enrichments only up to 4.45 w/o U-235 were allowed. These constraints have been resolved and the current spent fuel pool heat load n,ethodology can be used to support j

storage of fuel up to a maximum initial enrichment of 5.0 w/o U-235.

f 2.2 Thermal / Hydraulic Analysis of Soent Fuel Pool 2.2.1 Licensing Bases i

Aetails of the Callaway licensing bases are located in Section 9.1.2, Section i

9.1.3 and Appendix 9.lA of the Final Safety Analysis Repo[t (FSAR). These are: Case 1 - Coolant temperature limit of 140 degrees F, assuming placement of approximately 60 spent fuel assemblies in the spent fuel pool 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shttdown, and Case 2 - Coolant temperature of 160 degrees F, assuming placement of the entire core in the spent fuel pool 196.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after shutdown.

In both cases it is assumed that only one train of two available fuel pool cooling system trains as operating to cool the water in the UP.

For the design basis calculation, the licensee selected the SFP coolant temperatures for Case 1 (140 degrees F) and Case 2 (160 degrees F), and then calculated the decay heat generated in the SFP which would result in those 1

Changed from 135'F in Amendment 54 to accommodate fuel loading of 4.45 weight percent U-235.

b temperatures when operating only one of the two trains of the SFP cooling system. The resultant decay heat values for those conditions were found to be 26.40 E06 BTU /HR and 41.48 E06 BTU /HR, respectively. The maximum anticipated heat load to be removed by the fue' pool cooling system is based on the decay heat generated by a full core removed from the reactor and stored in the SFP 196.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> following a reactor shutdown, while spent fuel assemblies from previous refuelings remain in the SFP. The 196.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> consist of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> to cool down the reactor and 96.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to transport the full core to the SFP (1/2 hour per assembly). Technical Specification 3.9.3 mquires that the reactor be subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> before irradiated fuel in the core may be moved.

The licensing bases of 140 degrees F and 160 degrees F were found to be acceptable in previous licensing documents, including license amendment 54.

These continue to be acceptable.

2.2.2 Decay Heat The licensee in Mcated that the decay heat loads which were calculated for the partial and M core offloads, assuming the use of 5.0 w/o enriched fuel, are 19.948 E06 BTU /HR and 39.466 E06 BTU /HR, respectively, which are less than the licensing values of 26.40 E06 and 41.48 E06 BTU /HR.

For the purpose of calculation, it was assumed that an additional 84 spent fuel elements remain in the SFP at the end of each refueling period.

The licensee also assumed, for the purpose of calculation, that the SFP is filled with spent fuel elements.

Certain decay heat generation rates calcub.ted by the licensee were verified by the staff using values located in ANSI 5.1.

The staff's results were within 1 percent of the licensee's results. Since the licensee's calculations were less than those found in making the licensing bases calculations and the staff's calculations were in agreement with the licensec's, t.he licensee's calculations for decay heat are acceptable.

2.2.3 Coolant Temperatures The licensee stated that the maximum SFP coolant temperatun es with the increase in fuel enrichment of 5.0 w/o U-235 would be less than the design basis temperatures of 140 degrees F for a partial offload and 160 degrees F for a full core offload because the decay heat loads (19.948 E06 and 39.466 E06 BTU /HR) were less than those established for the licensing bases (26.4 E06 and 41.48 E06 BTU /HR).

Since SFP bulk coolant temperatures would not reach 140 degrees F for the partial offload case or 160 degrees F for the full core offload case, the temperatures (less than 140 degrees F and less than 160 degrees F), are acceptable.

At the staff's request, the licensee conducted an analysis that involved using the input parameters for Case 2 with the exception that two SFP cooling trains, instead of one, were assumed to be in operation. The bulk SFP coolant temperature calculated for that case was 133 degrees F.

The results of the

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. calculation indicate that in the event of necessity, the licensee can reduce the bulk temperature of the SFP coolant below the 160 degree F limit established for Case 2 when both cooling trains are available.

2.2.4 Coolant Bulk Boiling The licensee calculated the time it would take for the bulk coolant to boil starting at 140 degrees F and 160 degrees F, using the decay heat generation rates calculated for the licesing cases of partial and full core offload, respectively. The resultant elapsed times to reach boiling conditions are 8.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> for the partial offload and 4.03 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the full core offload case. Since in both cases, time would be available to utilize some source to restore the cooling process or to replace coolant which has evaporated should no other cooling source be available, this is acceptable.

2.2.5 Fuel Cladding Temperature The licensee calculated the maximuu surface heat flux for a hot assembly in the case of full core ofiload. The maximum cladding temperature resulting from this calculation (23;.6*F) is acceptable since it is much lower than the normal cladding temperature occurring during operation in the core.

2.2.6 Local Coolant Boiling The licensee noted that local boiling would not oce.ur in the spent fuel assemblies since the saturation temperatures at the locations of maximum heat flux exceeded the temperatures attained by the spent fuel cladding. 1he licensee also noted that even if local boiling were to occur, the net result would be a decrease in reactivity because of the presence of boron dissolved in the spent fuel pool coolant to the extent of a minimum concentration of 2000 ppe. This is acceptable.

2.2.7 Spent Fuel Pool Cleanup System The SFP cleanup system contains a desineralizer with resins to purify the SFP coolant.

The resins are the most temperature sensitive components of the SFP cleanup system and could become degraded at temperatures in excess of 140 degrees F.

In order to protect the desineralizing system against high temperatures, an annunciator is sounded in the control roca when the temperature of the SFP coolant reaches 130 degrees F, at which time operating procedures require that the cleanup pumps be shut down and manual isolation valves leading to and from the cleanup system be closed so that coolant with temperatures of 140 degrees F or greater will not enter the system spent fuel pcol cleanup system. The application of the annunciator and operating procedures to protect the deionizer resins in case of high coolant temperatures is acceptable.

2.3 S - ary of_Results The following tuhnical speci'ication changes have been proposed as a result of the requested enrichment increase.

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• I (1) TS 5.3.1 has been revised to reflect a change in the maximum initial enrichment to 5.0 w/o U-235 for reload fuel, subject to the IFBA requirements determined above, and to increase the maximum fuel

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enrichment not requiring IFBAs from 3.85 to 4.10 w/o U-235.

l (2)

TS 5.6.1.1 has been revised to increase the maximum reference K, from 1.455 to 1.480 for storage in Region 1.

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i (3) TS Figure 3.9-1 has been revised to reflect a maximum initial enrichment i

of 5.0 w/o U-235 for storage in Region 2.

l Based on the review discussed above, the staff finds the criticality aspects of the proposed enrichment increase to the Callaway SFP storage racks to be acceptable. The increase meets the requirements of General Design Criterion 62 for the prevention of criticality in the fuel storage and handling.

The staff also finds the thermal / hydraulic analysis of the SFP to be acceptable for storage of fuel in the SFP with an initial enrichment up to 5.0 w/o U-235.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Missouri State official was notified of the proposed issuance of the amendment.

The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environ =r.l.al assessment and finding of no significant impact was published ja the Federal Reoister on March 25, 1996 (61 FR 12112).

Accordingly, based upon the environmental assessment, the Commission has j

determined that issuance of this amendment will not have a significant effect on the quality of the human environment.

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5.0 CONCLUSION

The Commission,.s 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

L. Kopp N. Wagner Date:

April 30,1996 1

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