ML20137Z311
| ML20137Z311 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/02/1985 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20137Z277 | List: |
| References | |
| TAC-59743, TAC-59744, NUDOCS 8512110269 | |
| Download: ML20137Z311 (9) | |
Text
_.
@ Mr o
UNITED STATES 8"
NUCLEAR REGULATORY COMMISSION o
5
- j WASHINGTON, D. C. 20656
~
SAFETY EVALUATION AMENDMENT NO. 39 TO NPF-10 AMENDMENT NO. 28 TO NPF-15 SAN ONOFRE NUCLEAR GENERATING STATION, U F '; 2 & 3 DOCKET NOS. 50-361 AND 50-362 INTRODUCTION Southern California Edison Company (SCE), on behalf of itself and the other licensees, San Diego Gas and Electric Company, The City of Riverside, California, and The City of Anaheim, California, has submitted several applications for license amendments for San Onofre Nuclear Generating Station, Units 2 and 3.
One such request, Proposed Change PCN-199, is evaluated he min. This change would revise the technical specifications relating to the maximum enrichment of the fuel assemblies and the criticality requirements for storage of fuel in the fuel storage areas (reference PCN-199). These technical specifications are being changed because the Cycle 3 fuel enrichment is being changed from 3.7% to 4.1% to acconmodate an 18-month refueling cycle. Other amendments have been requested to modify the technical specifications associated with reactor operation with the revised enrichment, and are now being evaluated by the NRC staff. The amendments associated with PCN-199 were requested by the licensee's letters of August 23, October 10, and October 16, 1985. The request includes the results of analyses on the effect of the increased 11-235 fuel enrichment on the criticality aspects of both the new and spent fuel storage racks at SONGS 2 and 3.
The sta f evaluation of the proposed change is given below.
ANALYSIS METHODS The analysis of the criticality aspects of the storage of SONGS 2 and 3 fuel assemblies having a fuel enriclunent of 4.1 wt% U-235 was performed by SCE.
The previous analysis for the SONGS 2 and 3 new and spent fuel storage racks was performed by Nuclear Energy Services (NES) for the storage of fuel assemblies having a fuel enrichment of 3.7 wt% U-235. The SCE analysis methods consist of the N-KENO-IV/S and NITAWL-S codes which are part of the SCALE-2 (Ref. 3) code packaoe and of the EPRICELL-2 and DANC0FF codes which are part of the ARMP code package (Ref. 4). The KENO-IV/S code is a multigroup criticality program for the This code has the capability of modeling complex, three-dimensional syI b) detennination of a system's effective neutron multiplication factor (K The NITAWL-S code is used to perform the resonance self-shielding calculations for those nuclides with resonance paran.'0ers that are important to the criticality analysis. The Nordheim integral treatment (Ref. 5) is used by NITAWL-S.
1 Both NITAWL-S and KENO-IV/S codes were used with the 27 group neutron cross 8512110269 851202 PDR ADOCK 050003h1 P
PUR
7.,,
....... ~.....--...~.. ~... - -.
.... ~....
.., -. -... -.=..... ~..- ~-
. ~
-9 W
e k
y s
- f-c 9
5
.h (i.
M
+
4
' F l
l
+ _
4 I
E 4
9 4
f k
I ie ',
s 1
4..-
f t
i g.
1
- 1. -
s g
f T
f f
i 3 _.^
t 2a' i
.s d
i 4.
i.
4 f
4 -
1' 8
r i
i,.
4.
4 i
T m.
4 i
I f
7 a
Z t
.4-0 4
d.
T i
S 1-0 I
k :.
a I
b.
. section library furnished with the SCALE-2 code package. The EPRICELL-2 code is a fuel pin cell code used to determine fuel material concentrations for the 4.1 wt% fuel and Dancoff factors for input to the NITAWL-S code. The DANC0FF code was used to calculate Dancoff factors with Sauer's method for checking values computed using EPRICELL-2 and to calculate Dancoff factors for water densities for which the use of EPRICELL-2 is inappropriate.
~
The SCE analysis methods were benchmarked against ' critical experiments described in Reference 6.
Six experiments were analyzed, that is, experiments number 001, 004, 007, 008, 013, and 014. These experiments used arrays of U0 fuel p
rods with a fuel enrichment of 4.29 wt% U-235. Other physical characteristics of these experiments made them suitable for benchmarking analysis methods for analysing the SONGS 2 and 3 new and spent fuel storage racks. The SCE benchmarking results are presented in a document (Ref. 7) transmitted by Reference 2.
These results indicate that the SCE analysis methods underpredict (N.withabiasof-0.01322atthe95%confidencelevel.8) transmitted by Reference 2, Torrey In another document K
stated, in a letter attached to Reference 8, that SCE results for the nominal K
of a reference case with 4.1 wt% fuel were in good agreement with a value o8 bins. by extrapolating the NES results for 3.7 and 3.9 wt% fuel to 4.1 wt%
fuel i:E stated that the Torrey Pines extrapolation was erroneous and that the
' NES extrapolated value for 4.1 wt% fuel should have been 0.8890 as compared to the SCE value of 0.90111 (Ref. 9). The agreement is still acceptable considering the different analysis methods used by SCE and NES.
Although the SCE benchmarking is not extensive, the results obtained and the documentation provided indicate that the use of these methods by SCE for the analysis of the SONGS 2 and 3 new and spent fuel storage racks is acceptable.
SPENT FUEL STORAGE RACK ANALYSIS The criticality of fuel assemblies in the SONGS 2 and 3 spent fuel racks is prevented, primarily, by limiting the U-235 enrichment of the uranium in the U0 fuel rods. The SCE racks consist of stainless steel cans of square cross 7
section having an outer dimension of 8.81 inches and a nominal wall thickness of 0.125 inches. These storage cans are arranged in a square array with a pitch of 12.75 inches. Songs 2 and 3 are provided with separate spent fuel pools. Each pool has a present capacity to store 800 fuel assemblies.
The criticality criterion that the fuel assemblies with a fuel enrichment of 4.1 wt% U-235 must meet is that the effective neutron multiplication factor, K
, shall be less than or equal to 0.95 for normal and postulated accident c8bitions. The K shall include all biases and uncertainties at least at a95/95probabilithonfidencelevel. Even though the pools contain borated water' th'e analysis must assume unborated water when normal conditions are being comidered.
3-i SCE: performed a number of analyses to detemine the sensitivity of a reference i
K based on nominal rack dimensions and a pool water temperature of 68*F, to VIffa,tions in pool water temperature, rack pitch, eccentric fuel storaae in the racks, and water in the fu'el pin gap for 1% failed fuel. The fuel assemblies Jwere assumed to consist of a 16 X 16 array of fuel rods with appropriate spaces for 5 quide tubes, with a fuel enrichment of 4.1 wt% U-235, and with the U0 at 2
94.75% of theoretical density. The effect of a dropped fuel assembly on K was also determined. The dropped fuel assembly accident was postulated to,ff occur by dropping a fuel assembly on top of a loaded fuel storage location.
All.other possible dropped fuel assembly accidents result in greater than 24 inches of water between the dropped fuel assembly and the assemblies in the racks. SCE states that these dropped fuel assembly. accidents are not considered further since the _ dropped fuel assembly is, in effect, isolated from a criticality standpoint from the fuel-assemblies stored in the racks. This is an acceptable assumption to make in these analyses.
The SCE analyses detemined that the K of the reference case is equal to 0.90111 forfreshfuelassembliescontIfkinguraniumenrichedto4.1wt%in U-235. The uncertainties and biases are:
[
(1) 0.00266 Calculation uncertainty from benchmarking (2s)
(2) 0.01332 Calculation to measured bias from benchmarking (3) 0.01503 Minimum rack pitch and eccentric fuel load (4) 0.00192 Most reactive temperature (39*F)
(5) 0.00214 Dropped fuel assembly accident (6) 0.00008 Waterlogged fuel pins (1% failed) 0.03515 Total rack biases and uncertainties i
The total ancertainties and biases on the K associated with SCE's analysis of the Songs 2 and 3 spent fuel racks is 0.6N15 and, therefore, the maximum K
is equal to 0.936 for the 4.1 wt% enrichment fuel assemblies. Since the cIfbulationaluncertaintyistakenastwicethestandarddeviationofthedata and other uncertainties are for worst case conditions, SCE meets the intent of staff quidance in the determination of the biases and uncertainties at least at a 95/95 probability / confidence level.
SCE considered the dropping of a fuel assembly and included its reactivity I
effect in its uncertainty allowance.. Other postulated accidents that were considered were judged to have an insignificant reactivity effect.
In any case, the staff believes that-no other postulated accident would cause a criticality accident because credit may be taken for the boron in the pool water by invoking the Double Contingency Principle. This is acceptable.
t h
5
..,-g
. - -, - =,.,., -,,. -. - -.,..,
y 7--
-,--w
,..,,.,,-,--.,,----.---,,---..n-
-.,.v~
n
4 The preitous calculations for SONGS 2 and 3 were perfomed by NES for 3.7 wt%
enrichment fuel assemblies.
NES obtained a worst case K 0.946.
Since the worst case SCE value for 4.1 wt% enrichment fuel assembl N is 0.936, there is an apparent discrepancy between these two results. SCE states (Ref. 9) that this difference can be explained on the basis of its analysis model enhancements..To evaluate this contention, we can extend the SCE and NES results for 4.1% fuel which was discussed previously for nominal K s for ff the reference case. UsingNESresultsprovidedbySCE(Ref.9),w$ construct the following table:
SCE NES Nominal K for 4.1 wt% fuel
.9011
.8890*
Calculati8MtoMeasurementbias
.0133
.0487 Uncertainties (including pool
.0219
.0265 temperature and various worst case conditions)
Total Worst Case K,ff.
.9363
.9642 The comparison shown in the table demonstrates that the bulk of the reactivity difference is caused by the calculation to measurement bias term. Since the methods used by SCE are clearly an improvement over those used by NES (mainly diffusion theory analysis), the smaller SCE value for the calculation to measurement bias is reasonable as compared to the very conservative NES value.
Therefore, we conclude that there is no apparent discrepancy between SCE and NES results based on our review of the available information.
meetsthestaffcriterionof0.95,includinguncertainti$[fequalto'0.936 Based on'our review, we conclude that the SCE value of K and biases, for the storage of 4.1 wt% enrichment fuel assemblies in the SONGS 2 and 3 spent
- 1. fuel storane racks.
NEW FUEL STORAGE RACK ANALYSIS The' criticality of fuel assemblies in the SONGS 2 and 3 new fuel racks is prevented, primarily, by limiting the U-235 enrichment of the uranium in the U0 fuel rods. The SCE new fuel racks consist of stainless steel structural bers that fom square cans. The structural material is neglected in the analysis except for the four corner angle pieces. The storage locations are arranged in a square array with two pitches. The first pitch has a nominal value of 29 inches and the second a nominal value of 38 inches. SONGS 2 and 3
.are provided with separate new fuel storage racks. Each new fuel storage facility has a present capacity to store 80 fresh (unirradiated) fuel assemblies.
The new fuel storage racks are nomally in an air (dry) configuration.
i
- The NES value was extrapolated by SCE from NES results from 3.7 and 3.9 wt%
fuel. The NES nominal results included, according to SCE, an arbitrary NES added value of 0.01, 3
i 4
. The criticality criteria that the fuel assemblies with a fuel-enrichment of 4.1 wt% U-235 must meet are that the effective neutron multiplication factor, K
, shall be less than or equal to 0.95 for the dry storage racks, that K sRN1 be less than or equal to 0.95 when the racks are flooded with pure
'If water, and the K shall be less than or equal to 0.98 when the. racks are immersed with lo$$$ensity hydrogenous material due to such causes as, for example, mist, fog, or fire-fighting foam. The K shall include all biases anduncertaintiesatleastata95/95probabilityN$nfidencelevel. SCE has performed a comparison of calculational results to criteria in a manner that_is different than staff practices.
In this evaluation, we will adapt the available SCE results, as required, to compare to the listed criteria.
SCE performed calculations for the dry configuration modeled with a finite lateral geometry including the effect of the concrete walls. The racks are completely enclosed by concrete walls and does not have an open wall in the modeling. SCE calculated K for this configuration as 0.384, which includes twicethestandarddeviatio$$ftheKENO-IV/Sstatisticaluncertainty. SCE has l
f not established uncertainties in.its benchmarking for this type of configuration.
l However, its use of the calculational methods in the documentation presented clearly establishes SCE's ability to calculate these configurations. Even assuming large errors and. uncertainties, SCE easily meets the criterion on K,ff for this dry configuration.
SCE perfomed calculations for the fully flooded new fuel storage racks with the same finite lateral geometry as for the dry configuration.
For this
-configuration, the maximum K as a function of water density (0 to 1.0 gm/ccwaterdensity)occursIffthe maximum water density and was calculated to be 0.86294. Since the flooded new fuel storage racks are similar to the spent fuel storage racks, the spent fuel pool uncertainty in K of 0.0352 forthefT$$dednew may be used as a good approximation.
TheworstK'fkefullyfloodedracks storage. racks is thus equal to 0.898.
Therefore, meet the criterion on K,ff of being equal to or less than 0.95.
For the extreme, low-density hydrogenous moderator conditions, the SCE finite lateral geometry calculations indicate a second peak in the K versus water density curve. This K is equal to 0.813 at a water 9Iksity of about 0.05 gm/cc. Even though unc$ffainties have not been clearly established for this configuration and assuming even large uncertainties, SCE meets the criterion
- of the worst K,ff being less than or equal to 0.98 for this configuration.
A fuel handling accident for the dry configuration would increase the K,ff of the dry configuration by a delta-K of 0.08527. Adding this to the K of the dry configuration would give a K of 0.469. AgainthisfuelhIkbling f
accident would remain, including assume 0 uncertainties and biases, well below the criterion on X of 0.95.
This fuel handling accident was modeled conservatively in Ikinfinite lateral geometry and by misloading of every storage location with fuel assemblies.
.. Based on our review and interpretation of SCE calculations, we conclude that SCE meets all the criteria for the storage of 4.1 wt% fuel assemblies in the new fuel storage racks and we also conclude that the uncertainty analysis assumed meets the intent of the 95/95 probability / confidence level requirement.
FUEL TRANSFER CARRIER ANALYSIS Although the fuel transfer carrier is not normally a part of our review of new and spent fuel pool racks, we have reviewed the SCE results for the 4.1 wtf enriched fuel assemblies. The calculations were performed for two assemblies and.for a water density of 1 gm/cc at 68 F.
The SCE results indicate that the K is equal to 0.908 including a bias term. This result is certainly-accep@lesince.itpredictsamargintocriticality.
TECHNICAL SPECIFICATIONS Technical Specification 5.3 REACTOR CORE Based on our review, it is acceptable to change this specification from a maximum enrichment of 3.7 to 4.1 wt% U-235.
Technical Specification 5.6.1 CRITICALITY Our review indicates that the licensee needs-to revise this specification to include the new value of the uncertainty of 0.035 delta K/K.
SUMMARY
OF EVALUATION Based on our review, we conclude that SONGS 2 and 3 16X16 fuel assemblies having a maximum enrichment of 4.1 wt% uranium-235 may be stored in the new and spent fuel racks. Our conclusion is based on the followino:
1.
The criticality calculations have been perfomed with acceptable methods and have been benchmarked, 2.
Uncertainties have been accounted for, 3.
Postulated accidents have been considered, and 4.
The effective neutron multiplication factor, including a consideration of the various uncertainties and biases, meets our acceptance criteria.
CONTACT WITH STATE OFFICIAL The NRC staff has advised the Chief of the Radiological Health Branch, State Department of Health Services, State of Californic, of the proposed determinatins of no significant hazards consideration. No consnents were received.
. ENVIRONMENTAL CONSIDERATION These amendments involve changes in the installation or use of facility
~
components located within the restricted area. The licensee has stated that the average level of irradiation of the irradiated fuel discharged from the reactor will not exceed 33,000 megawatt-per metric ton. The staff has determined that the amendments involve no significant increase in the amounts of any effluents that may be released offsite and that there is no significant increase in individual or commulative occupation radiation exposure. The Commission has previously issued proposed findings that the amendments involve no significant hazards consideration, and there has been no public comment on categorical exclusion set forth in 10 CFR Sec. 51.22(c)gibility criteria for such findings. Accordingly, the amendments meet the eli (9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need to be prepared in connection with the issuance of these amendments.
CONCLUSION Based upon our evaluation of the proposed changes to the San Onofre Units 2 and 3 Technical Specifications, we have concluded that:
there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendments t
will not be inimical to the common defense and security or to the health and safety of the public. We, therefore, conclude that the proposed chances are acceptable, and are hereby incorporated into the San Onofre 2 and 3 Technical Specifications.
Dated: DEC2 1985
REFERENCES 1.
Letter from M. O. Medford (SCE) to Director (NRR), August 23, 1985.
2.
Letter from M. O. Medford (SCE) to Director (NRR), October 10, 1985.
3.
" SCALE-2: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation," NUREG/CR-0200, Oak Ridge National Laboratory.
4.
" Advanced Recycle Methodology Program System Documentation," Electric Power Research Institute, CCM-3, January 1976.
5.
L. W. Nordheim, "The Theory of Resonance Absorption," Proceedings of Symposia in Applied Mathematics, Vol. XI, 58, G. Birkhoff and E. P.
Wigner, Eds., Am. Math. Soc. (1961).
6.
" Critical Separation Between Suberitical Clusters of 4.29 wt% U-235 Enriched U0,ific Northwest Laboratories, May 1978.
Rods in Water with Fixed Neutron Poisons," NUREG/CR-0073, Battelle Pat 7.
" KENO-IV Benchmark Calculations for Criticality Analysis," C. W. Gabel, DC-1859, May 29, 1985.
8.
" Criticality Calculations for Fuel Storage Systems," C. W. Gabel, DC-1858, June 22, 1985.
9.
Letter from M. O. Medford (SCE) to G. W. Knighton (NRC) October 16, 1985.
l.
l l
z-ISSUANCE OF AMENDMENT NO. 39 TO FACILITY OPERATING LICENSE NPF-10 AND AMENDMENT NO. 28 TO FACILITY OPERATING LICENSE NPF-15 SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 DISTRIBUTION L 0diiCFili 50-361/362il?
NRC PDR Local PDR PRC System NSIC LB#3 Reading JLee(20)
HRood TMNovak JSaltzman, SAB LChandler, OELD CMiles HRDanton DGEisenhut JRutberg AToalston RDiggs LFMB EJordan
- LHarmon MVirigilo TBarnhart (8)
JPartlow BGrimes
. _,.