ML17249A894
| ML17249A894 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/15/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17249A891 | List: |
| References | |
| NUDOCS 8005130233 | |
| Download: ML17249A894 (11) | |
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 32 TO PROVISIONAL OPERATING LICENSE NO.
DPR-18 ROCHESTER GAS AND ELECTRIC CORPORATION R.
E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244
- 1. 0 INTRODUCTION By application (Reference
- 2) dated December 14, 1979 (transmitted by 'letter dated December 20, 1979),
Rochester Gas and Electric Corporation (RGt'E)
(the licensee) requested an amendment to License No.
DPR-18 for the R.
E. Ginna Nuclear Power Plant to allow plant operation with four plutonium mixed oxide (MOX) fuel assemblies.
By letters (Reference
- 6) dated February 20,
- 1980, and March 5, 1980, RGSE provided additional information responsive to".our questions.
The staff has previously evaluated generically the ability of nuclear reactors to operate with MOX fuel in excess of the four bundles now being considered for Ginna.
After discussions with and reviewing submittals from the domestic nuclear fuel manufacturers, the staff issued its findings in Chapter IV, Section C-3 of NUREG-0002 (Reference 1).
This section adequately discusses the differences in nuclear and material properties of MOX and UO fuel and the impact of these differences on reactor safety.
These genera differences will not be included in this safety evaluation for Cycle 10 operation with four MOX fuel assemblies, since these generic differences are not significant with respect to the proposed amendment to use four (4)
MOX fuel assemblies.
Our evaluation concerns the specific effects on reactor safety of loading four MOX assemblies and 32 new U02 assemblies in Ginna core beginning with Cycle 10 operation.
MOX fuel has been,irradiated in other U. S. light water reactors.
This experience up to 1975 is discussed in Reference 1.
The experience of Exxon and Westinghouse with MOX fuel is given in Tables 1
and 2 (attached).
- 2. 0 EVALUATION A description of the fuel to be irradiated during Cycle 10 in Ginna is provided in Table 3.1 of Reference 2.
The mechanical design of the fuel assemblies containing the MOX fuel is similar to fuel already irradiated
at Ginna (designated Region 7).
No problems have occurred with this fuel batch except for excessive fuel rod bowing.
Westinghouse, the manufacturer of, the Region 7 fuel and the 4
MOX fuel assemblies, in discussions with the staff, stated that this was traced back to the cladding material used for the Region 7.
The licensee has stated that none of this material was used for the MOX fuel rod cladding.
Based on previous operating experience of Westinghouse 14xl4 fuel, and specifically with the Region 7 fuel irradiated in Ginna, we anticipate no problems with the use of four MOX fuel assemblies.
The licensee, in Reference 2, notes that the densification of Westinghouse MOX fuel is less than or equal to that of UO An Electric Power Research Institute (EPRI) Study (Reference
- 3) showed that, in general, the behavior of MOX fuel is comparable with that of UO fuel, i.e.,
Pu0 additions to U02 typical of plutonium recycle fuel( do not create any limitations on performance in terms of densification.
Also, like U02 fuels, it was demonstrated in this EPRI Study that stability towards densification of the MOX fuel types studied is related to micro-structural characteristics, i.e., grain size, pore size, and volume percent of submicron porosity.
The licensee has presented data (Reference
- 6) which indicate that MOX fuel manufactured by Westinghouse does not densify any differently from U02 fuel manufactured by Westinghouse.
This conclusion is important in justifying the use of the standard Westinghouse densification model for LOCA analyses and other postulated accident analyses.
Data from MOX fuel irradiated in San Onofre, Saxton and Beznau were compared with the data base for UO fuel. given in WCAP 8218 (Reference 4) to show that no difference in densification would be expected.
2.2 Nuclear Desi n and Safet Anal sis Because only four MOX fuel bundles are to be included in the Cycle 10 reloading, and these four assemblies will be located symmetrically at the core periphery, the effect on the core properties will be minimal.
The values of the kinetics parameters for Cycle 10 and a calculation of shutdown margin are reported in Reference 2.
Cycle 10 with MOX fuel has slightly lower control rod worths and shutdown margin than without MOX.
The licensee reports that the differences are less than approximately 0.55 ak/k.
According to Reference 1, the uncertainty associated with the calculation of local power peaking in MOX fuel may be greater than that currently used for U02-fuel.
This effect was not considered by the licensee since the MOX fuel bundles will be in the periphery of the core at a power level below the core average.
We understand that the current plan for the next cycle is to continue to keep these bundles below the core average power.
However, after the second cycle these MOX assemblies might be placed in core positions where the bundle power will be greater than core average.
To assure that the power is being adequately calculated for the MOX assemblies, the licensee will compare the measured and predicted powers in the instrumented MOX assemblies with the measured and predicted powers in adjacent U02 assemblies.
The data will be reported to the NRC at each refueling outage following Cycle 10.
Exxon Nuclear Company performed the physics calculations for expected Cycle 10 core configuration.
Comparisons of Exxon calculational methods for HOX fuel with data are given in Reference 5.
In particular, Tables 4.2-1 and 4.2-2, 4.2-3 and 4.2-4 give comparisons with critical experiments which contained~UO~
rods and Pu02 rods.
These comparisons are an indication of the ability of Exxon's physics methods to calculate power distributions and related quantities such as neutron multiplication factors and buckling.
In general, the comparison is good.
For Cycle 10, because of the addition of the four MOX assemblies, the reactivity worth of the boric acid will slightly decrease and the BOC delayed neutron fraction will slightly decrease so that the values assumed for safety analyses for the postulated accidents listed below in Table 3 were reevaluated.
These accidents are the most limiting'ith respect to the above two parameters.
The results of the analyses show that the applicable safety criteria for each event were met.
The reference analyses for Cycle 10 are given in References 8 and 9.
il TABLE 3 Steam Line Break (Large and Small)
Fast Rod Withdrawal Rod Ejection Although boron worth decreases, the safety criterion for the steam line break will still be met since the minimum Departure from Nucleate Boiling Ratio (DNBR) of the reference analysis is above the safety limit of 1.3 and the change in the delayed neutron fraction (8) from the reference analysis would result in only a slight increase in fuel rod power and a negligible change in DNBR.
A recalculation of the Rod Ejection Accident showed that the maximum total peaking factor (F~) after ejection was less than that for the reference cycle!
The LOCA analysis was not redone for Cycle 10.
The licensee stated (Reference
- 6) that the volumetric average temperature (stored energy) for the MOX fuel (at the same power and burnup) will be lower than for UO fuels The staff has performed an independent calculation to verify this result.
Our calculations show only a slight difference between the volumetric average temperature calculated with MOX fuel at 3.1$
Pu0~
and UO fuel with U-235 enrichment of 3.45% (the enrichment of the Regi5n 12 fuel).
The calculated UO volumetric average temperature is slightly higher.
These calculations ktilized the NRC code GAPCON THERMAL-2.
Densification and fuel relocation were both considered.
The confirmatory NRC calculations were done for a peak power fuel rod and a fuel rod at slightly above the average core power to a burnup of 5000 MWd/MTU to account for densification effects.
The flux depression for the MOX fuel was based on calculations done for the EPRI densification study (Reference 3). It is noted that although the MOX fuel has a lower thermal. conductivity than the UOq, more of the heat is generated in the outside area of the fuel pellet and Tess at the center due to the neutron flux depression in the MOX fuel rod interior.
As part of the calculation of Fn, the licensee must include the effects of fuel rod bowing.
As a fuel Pod bows, the local moderation will increase and may result in power peaking.
In Reference 7, Westinghouse presents calculations which show that this effect can be adequately accounted for within the existing uncertainty allowance.
However, this calculation was for U02 fuel only.
The MOX fuel bundles, like all the Westinghouse fuel used in Ginna, is HIPAR, meaning that the reactor cluster control guide tubes are stainless steel.
Westinghouse has previously presented data to the staff to show that the amount of fuel rod bowing in HIPAR fuel is negligible.
Therefore, the effect of any power peaking due to fuel rod bowing in the MOX fuel assemblies will be negligible.
3.0
SUMMARY
The addition of four MOX fuel assemblies results in negligible changes to the Ginna Cycle 10 core.
The licensee has taken the differences in fuel material properties into account in evaluating Cycle 10 performance.
The fuel bundles are identical in design to Westinghouse fuel bundles previously irradiated satisfactorily at Ginna.
Two parameters, the boron worth and the delayed neutron fraction are outside of the range of values used for previous accident analyses.
The licensee reevaluated the most limiting postulated accidents for which these parameters have a significant effect and concluded that the applicable safety criteria are still met.
Hased on the-above, we have concluded that the Ginna reactor can be operated safely during Cycle 10 operation with four MOX fuel assemblies.
However, the licensee must determine the nuclear uncertainty on power peaking for the MOX fuel rods before operation for future cycles.
This uncertainty; after review and approval by the staff, should be applied
to the MOX fuel assembly irradiation beyond Cycle 10.
4.0 COMMISSION POLICY - MOX FUEL The proposed action to amend the Provisional Operating License No.
DPR-18 for Ginna is consistent with the Commission's Memorandum of Decision, dated Myy 8, 1978, (In the Matter of Mixed Oxide Fuel, CL1-78-10, 7
NRC 711 (1978))'/,
and the Commission's Order of December 23, 1977, concerning its proceeding on the Generic Environmental Statement on Mixed Oxide Fuel (GESMO) and matters related to reprocessing and the recycling of uranium and plutonium in mixed oxide fuel.
42 FR 65334 (December 30, 1977); CLI-77-33, 6
The proposed action is consistent with the Coranission's policy on the use of MOX fuel in that the proposed use of the four MOX assemblies in the Ginna reactor involves the use of a small quantity of MOX fuel for experimental, demonstration, and feasibility purposes on a noncommercial basis.
The proposed use does not involve wide-scale commercial reprocessing.
Our conclusion is based on the following factors.
The proposed action to use four MOX fuel assemblies involves the use of less than 50 kg of plutonium.
Rochester Gas and Electric does not presently have other contracts in existence for the purchase of MOX fuel.
Nor does it now plan to use MOX fuel in the future in addition to these four MOX assemblies.
As stated in the application submitted by Rochester Gas and Electric, the proposed insertion of the MOX fuel into the Ginna reactor is the culmination of the experimental work carried out as part of a Research, Demonstration, and Development
("RDD") program initiated by Rochester Gas and Electric approximately six years ago (Application, Attachment D, "Research, Demonstration and Development Aspects of the Proposed Use of Mixed Oxide Fuel Assemblies," at page 1).
Completion of this experimental RDD program by the use of the four MOX assemblies will allow Rochester Gas and Electric to:
1/
The Commission's Memorandum of Decision of May 8,
- 1978, and its Order of December 30, 1977 concerning its GESMO proceedings was upheld by the Third Circuit Court of Appeals in'Hestinjhouse Electric
~di I
d II 598 f. 2d 759 (3r
. Cir, 1979
(a)
(b)
(c)
(d)
(e)
(g)
Verify current neutronic methodology applied to mixed oxide assemblies; Verify current capabilities to calculate incore detector responses in mixed oxide assemblies relative to all-uranium assemblies; Obtain a degree of mathematical confidence relative to the capability to predict mixed oxide assembly reactivity and migration area as a function of burnup; Compare calculated and measured control rod worths for these mixed oxide assemblies when used in future cycles if desired; Make visual comparisons of fuel assemblies during refueling outages to determine if there are any differences in mixed oxide assemblies, as opposed to the all-uranium assemblies; Analyze the actual power distribution for two MOX assemblies on a regular basis and validate existing PtlR design codes by comparison with actual operating data.
Participate in post-irradiation programs.
(h)
Obtain information that is not currently available and achieve, a substantial advance in state-of-the-art knowledge concerning the use of mixed oxide fuel in commercial nuclear reactors.
Obtain energy spectrum data on available based on mixed oxide rods fabricated with processes the identification of the fuel fuel densification not currently fuel in pressurized, zircaloy clad developed and approved following densification problem.
Use of the fuel converts it to a form that is less vulnerable to safeguards risks for two reasons.
l<hile in the reactor the fuel is virtually inaccessible.
Once used, MOX spent fuel is virtually indistinguishable from the normal highly radioactive discharged U02 spent fuel.
The Commission has also obtained the views of the Administration in connection with this matter:
"Several considerations are relevant in this connection:
First, it is our understanding that the MOX fuel was fabricated well before the 1977 announcement of President Carter's policy, and no fabrication of MOX fuel is now going on or contemplated.
I would note that we have, in the past, permitted
~ex ort of MOX fuel for recycle RSD in three cases where commitments had been made
prior to our April 1977 policy change.
I also understand the present holders of plutonium generally want to divest themselves, to avoid the need for associated special physical protection measures.
In fact, another of the general non-proliferation efforts has been to move toward a situation in which presence of unirradiated plutonium, more vulnerable to theft or diversion, is minimized.
These considerations would indicate that, from a foreign policy standpoint, approval of the RG8E license would not be seen as a new thermal recycle
- program, but rather as cleaning up or minimizing an old problem."
(Letter from Stuart E. Eizenstat, Assistant to the President for Domestic Affairs and Policy to Chairman Ahearne, dated April 4, 1980).
Since the four mixed oxide fuel assemblies with a total. quantity of less than 50 kg of plutonium constitute a very small quantity of plutonium and fuel assemblies to be used as part of an experimental
- program, there will be no foreclosure of future safeguards options or future operational alternatives,
- 5. 0 ENVIRONMENTAL CONSIDERATION t
Essentially the only aspect of the change in assemblies that will affect offsite releases from postulated accidents is the change in fission product inventory in the core, and any resulting changes in the fission product concentration in the reactor coolant.
The change in the core inventory is very small for the nuclides that are of highest significance during a release, with the greatest changes being those of I-131 and I-132.
For the GESMO model MOX reactor (Reference 1), the increase over' U-only reactor is about 3.6g for I-131 and 8.3% for I-132, compared to the cori-centrations that exist just prior to refueling.
However, the Pu added in the four assemblies planned for insertion in the Ginna reactor results in a total heavy-metal percentage of new Pu in the core of only 0.09Ã (Exxon Nuclear Inc. XN-NF-79-103).
This is about 5'4 of the initial Pu fraction that was used in the GESMO model, so the change in fission product inventories would also be 5X of the changes in the GESMO report.
Thus, there is only a negligible (at most 0.18 to 0.425) increase expected in the concentration of any fission product that could contribute to accidental offsite doses, and therefore no significant change from our previous accident analyses.
We have further determined that the proposed amendment does not authorize a change in effluent types, increase in total amounts of effluents, or an increase in power level, and will not result in any significant environmental impact.
Having made this determination, we have concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact, and, pursuant to 10 CFR 51.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance -of this amendment.
- 6. 0 CONCLUSION We also conclude, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the pr'obability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) 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 the health and safety of the public.
Attachments:
Tables 1 and 2
Date:
References 1.
Fuel Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in Light 1<ater Cooled Reactors:
Health, Safety and Environment, NUREG-0002, Vol. 3, Chapter IV, Section C-3, August 1976.
2.
Letter to H. Denton, USNRC, from H. Voigt of LeBoeuf, Lamb, Leiby and
- MacRae, December 20, 1979 (enclosing application dated December 14, 1979) which contains as addenda the following documents:
Mestinghouse Fuel and LOCA Evaluation of R.
E. Ginna Mixed Oxide Fuel Assemblies Plant Transient Analysis for the R.
E. Ginna Unit 1 Nuclear Power Plant Addendum to the Criticality Analysis for the Ginna Nuclear Plant Fuel Storage Racks to Address the Storage of Mixed Oxide Fuel Assemblies Radiological Impact of Mixed Oxide Fuel Assemblies R.
E. Ginna Nuclear Plant Cycle 10 Safety Analysis Report with Mixed Oxide Assemblies 3.
Plutonia Fuel Study, Electric Power Research Institute (EPRI), January 1978.
4.
Helluan, J. M., et. al.,
"Fuel Densification Experimental Results and Model for Reactor Applications," Westinghouse Electric Corporation, WCAP 8218, October 1973.
5.
- Skogen, F. K., "Exxon Nuclear Neutronic Design Methods for Pressurized Mater Reactors,"
Exxon Nuclear Company, Inc., XN-75-27 June 1975.
6.
Letter to Director of Nuclear Reactor Regulation, USNRC, from L. D. >lhite, Rochester Gas and Electric Corporation, February 20,
- 1980, and letter dated March 5, 1980 from L.
D. i<hite enclosing a better copy of page 16 to Attachment A of the February 20, 1980 submittal.
7.
- Reavis, J. R., et. al., "Fuel Rod Bowing," Mestinghouse Electric Corporation, MCAP 8691, December 1975.
8.
Markowski, F. J., et. al., "Plant Transient Analysis for the R.
E. Ginna Unit 1 Nuclear Power Plant,"
Exxon Nuclear Company, Inc., XN-NF-77-40.
9.
Markowski, F. J., et. al., "Plant Transient Analysis for the R.
E. Ginna Unit 1 Nuclear Power Plant,"
Exxon Nuclear Company, Inc., XN-NF-77-40, Rev.
1 July 3, 1979.
TABLf 1 EXXON NUCLEAR COMPANY MIXED OXIDE FUEL PERFORMANCE
~
~
RECTOR NUNBER OF ASSEMBLIES NTRIX EXPOSunE <engr)
NERAGE NXINUN BIG ROCK POINT llxll llxll llxll llxll llxll 6xe 30,ii00 25,000 20,300 17,100 15,700 11,600 30,F00 25,000 30,800 17,800 17,900
'>,200
'ISCHARGED
TABLE 2 WESTINGHOUSE HIXED OXIDE IAAADIATIONEXPERIENCE IN PLIRs Reactor Core C cle Number of Rods Power (kw/ft)
~Burnu
~klMD/HTU Dates Of 0 eration Saxton Saxton San Onofre San Onofre eeznau Core II Core III Cycle 2
Cycle 3
Cycle 8 Cycle 9 638 250 720 716 716 716 18 7( '
21.2(
,,(1) 6.1(
)
(4) 28,000 51,000(')
12,600(')
25,200(2) 11,200 20,900 5) at.EOC 9
Dec.
1965 to Oct. 1968 Oec.
1969 to Hay 1972 Nov. 1970 to Oec.
1971 Harch 1972 to June 1973
. June 1978-June 1979 Presently operating (1)
Peak pellet power achieved during the cycle.
(2)
Peak pellet burnup at the end of life.
(3)
Two mixed oxide fuel rods achieved 18.7 kw/ft during a special overpower test.
However, the peak power for the remainder of rods was 13.7 kw/ft..
(4)
Assembly average power.
(5)
Assembly average burnup.