ML20134C926
| ML20134C926 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 10/03/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20134C925 | List: |
| References | |
| NUDOCS 9610110134 | |
| Download: ML20134C926 (8) | |
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UNITED STATES g
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WASHINGTON, D.C. 20566-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.131 TO FACILITY OPERATING LICENSE NO. NPF-10 AND AMENDMENT NO.120 TO FACILITY OPERATING LICENSE NO. NPF-15 SOUTHERN CALIFORNIA EDIS0N COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE. CALIFORNIA THE CITY OF ANAHEIM. CA'LIFORNIA 1
SAN ON0FRE NUCLEAR GENERATING STATION. UNITS 2 AND 3 DOCKET NOS. 50-361 AND 50-362
1.0 INTRODUCTION
By application dated December 6, 1995, Southern California Edison Company (SCE or the licensee) requested changes to the Technical Specifications (Appendix A to Facility Operating License Nos. NPF-10 and NPF-15) for San Onofre Nuclear Generating Station (SONGS), Unit Nos. 2 and 3.
The proposed changes would allow fuel assemblies having a maximum U-235 enrichment of 4.8 weight percent (w/o) to be stored in both the spent fuel racks and the new fuel racks.
The August 30, 1996, and September 20, 1996, supplemental letters provided additional clarifying information and did not change the initial no significant hazards consideration determination which was published in the i
Federal Reoister on April 10, 1996 (61 FR 15997).
2.0 BACKGROUND
The licensee proposes to increase the allowable maximum fuel-pin enrichment from 4.1 w/o to 4.8 w/o U-235 for SONGS Units 2 and 3.
This will allow an increase of the current cycle length from approximately 520 Effective Full Power Days (EFPD) to approximately 600 EFPD.
The proposed TS amendments will revise TS Section 4.3 " Fuel Storage" to allow fuel assemblies having a maximum U-235 anrichment of 4.8 weight percent (w/o) to be stored in both the spent fuel racks and the new fuel racks.
Additionally, TS Section 3.7.18 " Spent Fuel Assembly Storage," Figures 3.7.18-1 " Unit 1 Fuel Minimum Burnup vs Initial Enrichment for Region II Racks," and 3.7.18-2 " Units 2 and 3 Fuel Minimum Burnup vs. Initial Enrichment for Region II Racks," would be revised and relabeled. A single value is being provided as a burnup limit for unrestricted storage of Unit I spent fuel 9610110134 961003 PDR ADOCK 05000361 P
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- assemblies in Region 11 rack locations. Another single value is being provided as a burnup limit for storage of Unit I spent fuel in the Region II peripheral rack locations. Therefore, the current Figure 3.7.18-1 is being replaced with a curve applicable to Units 2 and 3 fuel assemblies and relabeled appropriately.
3.0 EVALUATION The staff's evaluation of the proposed increase in fuel enrichment is separated into its effects on criticality, decay heat generation, and radiological dose consequences.
3.1 Criticality Analysis The analysis of the reactivity effects of fuel storage in the new and spent fuel storage racks was performed with the three-dimensional multi-group Monte Carlo computer code, KEN 0 V.a, using 27 energy group neutron cross sections generated by the NITAWL-II code package. Since the KEN 0 V.a code package does not have depletion capability, burnup analyses were performed with the two-dimensional transport theory code, CASM0-3. CASM0-3 was also used to determine the reactivity effects of material and manufacturing tolerances.
These codes are widely used for N analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments. These experiments simulate the SONGS Units 2 and 3 fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment, assembly spacing, and absorber worth.
The intercomparison between two independent methods of analysis (KENO V.a and CASMO-3) also provides an acceptable technique for validating calculational methods for nuclear criticality safety. To minimize the statistical uncertainty of the KEN 0 V.a reactivity calculations, a minimum of 500 neutron generations with 2000 neutrons per generation were accumulated in each calculation.
Experience has shown that this number of histories is quite sufficient to assure convergence of KENO V.a reactivity calculations. Based on the above, the staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the SONGS Units 2 and 3 new and spent fuel storage racks with a high degree of confidence.
The fresh fuel storage racks provide dry storage for 80 assemblies at a nominal center-to-center spacing of 29 inches and 38 inches.
The storage racks are intended for the receipt and storage of fresh fuel under dry (air) conditions. However, to assure the criticality safety under normal and accident conditions and to conform to the requirements of General Design Criterion 62 for the prevantion of criticality in fuel storage and handling, two separate criteria must be satisfied as defined in NRC Standard Review Plan (SRP), Section 9.1.1.
These criteria state that the maximum reactivity of the fully loaded fuel racks shall not exceed a k,,, of 0.95 if fully flooded with unborated water or a k of 0.98 assuming the optimum hypothetical low density moderation (e.,,.,, fog or foam). The maximum calculated reactivity g
must include a margin for uncertainties in reactivity calculations and in manufacturing tolerances such that the true k,,, will not exceed the
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calculated maximum value at a 95 percent probability, 95 percent confidence level (95/95).
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Unirradiated ABB Combustion Engineering (ABB/CE), Zircaloy-clad, 16x16 fuel assemblies will be stored in these racks. Although the requested maximum fuel rod enrichment for SONGS 2 and 3 is 4.8 w/o U-235, the criticality analyses I
were performed for up to 5.1 w/o U-235 and thus bound the requested value.
The maximum k 'u' lated to be 0.904 under fully flooded conditionsfor a fully l U-235 was calc For the hypothetical low-density optimum moderation condition, the maximum calculated i
k was 0.856 at a moderator density of approximately 4.5 percent of full d,e,n,sity for a fully loaded rack of 5.1 w/o fuel. The methodology bias i
obtained from benchmark results was included, but since the maximus k,, for both adverse conditions was less than 0.91, the manufacturing tolerances and calculational uncertainties (typically less than 0.01 ok) can be neglected.
The results conform to the acceptance criteria of SRP 9.1.1 and are, 4
l therefore, acceptable.
The storage racks in the spent fuel pool are divided into two regions.
Region j
1 contains 312 stainless steel storage cells with each cell surrounded on all i
four sides by Boraflex neutron absorber panels. The cells are spaced 10.4 inches a panels. part with a 1.1-inch water flux-trap between two adjacent Boraflex j
Region 2 consists of 1230 storage cells surrounded with Boraflex panels and assembled in a checkerboard pattern, producing a honeycomb structure. The cells are stainless steel with an inside dimension of 8.63 inches. The spent fuel racks are normally fully flooded by water borated to
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at least 1850 ppo of boron as required by the TS. However, to meet the i
criterion stated in SRP 9.1.2, k,,,ticipated reactivity and flooded with must not exceed 0.95 with the racks fully loaded with fuel of the highest an unborated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity must include a margin for uncertainties in reactivity calculations and in manufacturing tolerances such that the true k,,, will not exceed 0.95 at a 95/95 probability / confidence level.
Although the requested maximum fuel rod enrichment is 4.8 w/o U-235, the spent fuel storage racks in Region I were reevaluated for 5.1 w/o U-235 enriched fuel moderated by pure water at 20*C with a density of 1.0 gm/cc, which I
results in the highest reactivity.
For the nominal storage cell design in Region 1, uncertainties due to tolerances in storage cell inner dimension, i
stainless steel thickness, water gap thickness, Boraflex wrapper thickness, and Boraflex density, width and thickness were accounted for, as well as 4
I eccentric fuel positioning. These uncertainties were appropriately determined j
at the 95/95 probability / confidence level.
In addition, calculational and methodology biases and uncertainties obtained from benchmarking were included.
The reactivity calculations also considered the effects of Boraflex shrinkage and gap formation. All Boraflex panels were modeled with 4 percent width f
shrinkage.
In addition, a 6-inch gap (4 percent axial shrinkage) was randomly placed in every Boraflex panel.
Based on the results of blackness testing i
performed at other storage facilities, and on upper bound values recommended by Electric Power Research Institute (EPRI), the staff concurs that these
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i assumptions bound the current measured data and future development of i
additional shrinkage and gaps.
The final Region 1 design, when fully loaded with fuel enriched to 5.1 w/o U-235, resulted in a k,,, of 0.9413 when 4
combined with all known uncertainties and the Boraflex gap assumption. This meets the staff's criterion of k no greater than 0.95 including all uncertainties at the 95/95 probaN11ty/ confidence level and is, therefore, j
acceptable.
j Elevated silica concentrations have been observed in the SONGS spent fuel j
pool, indicating possible Boraflex erosion. Therefore, calculations were i
J performed to investigate the reactivity consequences of loss of Boraflex thickness in the Region 1 storage racks. KEN 0 V.a calculations assuming unirradiated 5.1 w/o U-235 fuel and a 6-inch random gap in every Boraflex panel show that about 20 percent of the Boraflex thickness can be lost i
uniformly before k,, reaches the 0.95 criterion in pure water. The current spentfuelpoolwalersilicalevelatSONGSindicatesthatthelossof i
Boraflex has been less than 3 ppm in five years and therefore, a 20 percent thickness loss appears to be reasonably conservative. However, SCE will continue to monitor the Boraflex integrity through their coupon surveillance program and the monitoring of pool silica levels. Based on this, the staff concurs that reasonable assumptions have been made to account for Boraflex deterioration in the SONGS storage racks.
The Region 2 spent fuel storage racks were reanalyzed for storage of ABB/CE 16x16 fuel assemblies with nominal enrichments up to 5.1 w/o U-235 (thus bounding the requested 4.8 w/o U-235 maximum fuel rod enrichment) using the concept of burnup reactivity equivalencing. The same initial assumptions, biases and uncertainties as used for the Region 1 analyses were included. A depletion uncertainty of 5 percent is applied to the total reactivity decrement calculated by CASMO-3 to account for the fact that the burnup history (and thus the isotopic content and distribution) is not known exactly for the discharged fuel assemblies.
This uncertainty is consistent with current practice and is acceptable.
The equivalencing showed that fresh fuel enriched to 1.85 w/o U-235 was equivalent to 5.1 w/o U-235 fuel irradiated to 38.6 gigawatt-days per ton (GWD/T), yielding a 95/95 rack reactivity (k,,,) of 0.948. These values meet the NRC acceptance criterion of 0.95 and are acceptable.
Because of the high neutron leakage from the peripheral cells which face the pool walls in Region 2, additional calculations were performed to determine
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the burnup required for safe storage in these cells.
Fresh 2.30 w/o U-235 enriched fuel or 5.10 w/o U-235 enriched fuel irradiated to at least 28.3 l
i GWD/T can be stored in these peripheral locations and meet the 0.95 k,,,
criterion.
Most abnormal storage conditions will not result in an increase in the k,,, of the racks. However, it is possible to postulate events, such as a fuel assembly dropped between a rack module and the pool wall, or the misloading of an assembly in Region ?. with an enrichment and burnup combination outside of the acceptable area, which could lead to an increase in reactivity. However, for such events credit may be taken for the presence of approximately 1850 ppm
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. of boron in the pool water required by TS 3.9.17 since the staff does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident (Double Contingency Principle). The reduction in k,,, caused by the boron more than offsets the reactivity addition caused by credible accidents. Therefore, the staff criterion of k,, no greater than 0.95 for any postulated accident is met.
Based on the review described above, the staff finds the criticality aspects of the proposed enrichment increase on the SONGS Units 2 and 3 new and spent fuel pool storage racks are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling.
1 This criticality evaluation is limited to the storage of fuel enriched up to 4.8 w/o U-235 in the new and spent fuel racks.
Evaluations of reload core designs (using any enrichment) will, of course, be performed on a cycle-by-cycle basis as part of the reload safety evaluation process.
Each reload design is evt.luated to confirm that the cycle core design adheres to the limits that exist in the accident analyses and TS to ensure that reactor operation is acceptable.
The staff issued an exemption to 10 CFR 70.24, " Criticality Accident Requirements," for SONGS Units 2 and 3 on August 20, 1990. This exemption allows irradiated or unirradiated fuel assemblies to be handled and stored in the fuel handling building without having any criticality monitoring systems operable. The bases for the exemption is the potential of accidental criticality is precluded.
The staff's evaluation of the criticality analysis as detailed above, for fuel enriched to 4.8 w/o U-235 reconfirms that accidental criticality will be precluded. Therefore, the exemption remains valid.
3.2 Decay Heat Analysis New fuel, either in the dry storage racks or in the spent fuel pool, does not produce significant amounts of heat. The decay heat analysis presented herein is for spent fuel.
The two relevant heat load cases were considered in this safety evaluation report (SER). The normal full core offload case assumes the decay heat from a full core after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of decay concurrent with a full spent fuel pool, and assumes the failure of a single spent fuel pool cooling pump. The maximum abnormal heat load case is defined as the decay heat from one full core offload after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> decay, plus the decay heat frem one full core offload after 36 days decay, in addition to the decay heat of spent fuel from past offloads.
Single failure is not considered in the maximum abnormal case.
The staff's SE accompanying the May 1, 1990, issuance of Amendments 87 and 77 for SONGS Units 2 and 3, respectively, approved the increase in spent fuel pool capacity that is currently part of the SONGS Units 2 and 3 design basis.
The staff concluded that the assumptions used by the licensee to characterize
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- the maximum heat loads on the spent fuel pool cooling system were acceptable.
The staff further concluded that the spent fuel pool and associated cooling system were adequately sized to handle the maximum heat loads for the increased spent fuel pool capacity. The results of this decay heat analysis is contained in the Updated Final Safety Analysis Report (UFSAR) and is presented in the table below. The licensee utilized the same methodology and assumptions, with the exceptions noted below, in analyzing the two heat load cases describe above for 4.8 w/o U-235 fuel reloads, and the results are also presented in this table.
Case Maximum Abnormal 49.9 MBtu/hr Heat Load Maximum Refueling 42.0 MBtu/hr Full Core Offload Heat Load
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As can been seen from the table, the 4.8 w/o U-235 fuel heat loads are less than the UFSAR heat loads and therefore the UFSAR maximum allowable temperatures are still applicable. The 4.8 w/o U-235 fuel heat load cases are less than the UFSAR heat load cases because of the following changed assumptions.
1.
The UFSAR assumed 1572 assemblies in the spent fuel pool (30 more assemblies than the 1542 storage locations in the pool), whereas the 4.8 w/o U-235 analysis assumed 1545 assemblies were in the spent fuel pool.
Fewer assemblies result in lower heat loads.
2.
The UFSAR assumes a cycle length of 570 EFPD and a batch size of 108 assemblies being discharged per fuel cycle. The analysis for 4.8 w/o U-235 fuel assumes a cycle length of 635 EFPD and a batch size of 104 assemblies discharged per fuel cycle.
Filling the pool at a slower rate results in a lower calculated heat load in the pool for the 4.8 w/o U-235 fuel case.
3.
The UFSAR decay heat loads include the heat load from 52 Unit I assemblies transshipped on a regular basis, some of which are assumed to have decayed for 120 days.
Transshipment is no longer occurring, since Unit 1 is no longer operating.
The staff finds the changes to the assumptions are reasonable and acceptable.
The total heat loads calculated for the 4.8 w/o U-235 fuel cases more i
accurately reflect the licensee's current fuel management plans and, as a result, lower the total anticipated heat load in the pool.
In its September 20, 1996, letter, the licensee has committed to update the UFSAR to show that l
the calculated heat loads for the 4.8 w/o U-235 fuel are lower than the design l
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_7 basis heat loads, and to clarify the description of the heat load cases in the UFSAR. The licensee further committed to perform a Condition Specific Analysis if the UFSAR assumptions are deviated from in a non-conservative manner, and to perform the analysis specified in Section A.4.3 of American Concrete Institute Standard 349 for nomal operations (full core offload) above 150*F before the next core offload. This analysis will confirm the operability of the spent fuel pool at temperatures up to 160*F.
Since the heat load cases for the 4.8 w/o U-235 fuel reloads are lower than the UFSAR design basis heat load analyses previously accepted by the staff in its May 1,1990, SER, the staff concludes that the spent fuel pool and its associated cooling system are capable of safely storing the 4.8 w/o U-235 spent fuel assemblies assumed in the licensee's analysis.
3.3 Radioloaical Dose Analysis The May 1,1990, SE reanalyzed the fuel handling design basis accident assuming that the fuel burnup was 60,000 MWD /T.
This burnup is equivalent to fuel initially enriched to about 5.3 w/o U-235, which is higher than the fuel enrichment cf 4.8 w/o U-235 proposed by the licensee.
Since the radiological dose consequences for fuel enriched to about 5.3 w/o U-235 was found acceptable by the staff in itt May 1,1990, SE, it follows that this conclusion also applies to the current case of 4.8 w/o U-235 fuel. The licensee comitted not to move a spent fuel pool gate except in a safe load path (i.e., no fuel assemblies in the drop zone) in its letter dated August 30, 1996.
This commitment removes the possibility of damaging multiple fuel assemblies due to an inadvertent drop of a fuel pool gate.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the California State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Reaister on September 26, 1996 (61 FR 50513).
Accordingly, based 1oon the environmental assessment, the Commission has determined that i'wnce of the amendments will not have a significant effect on the quality ot.ne human environment.
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6.0 CONCLUSION
The Commission has 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
- l Principal Contributors:
Larry Kopp, SRXB Norman Wagner, SPLB Date:
October 3, 1996 2
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