ML17249A370
| ML17249A370 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 12/31/1979 |
| From: | Skaritka J WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17249A368 | List: |
| References | |
| NUDOCS 7912280213 | |
| Download: ML17249A370 (20) | |
Text
WESTINGHOUSE FUEL AND LOCA EVALUATION OF R.
E.
GINNA MIXED OXIDE FUEL ASSEMBLIES December 1979 Edited by J. Skaritka M.~G. Arlotti, Manager Fuel Licensing 8 Coordination Nuclear Fuel Di visi on
y I Q
1.0 INTRODUCTION
'AND
SUMMARY
Rochester Gas and Electric plans to have four mixed oxide (Pu02-U02) assemblies fabricated and inserted into the R.E.Ginna reactor for Cycle 10 as demonstrated assembles;
=.The mixed oxide fuel rods were originally fabricated by Westinghouse for Cycle 7 and have been in storage since 1974.
Each 14x14 fuel assembly would contain 179 mixed oxide fuel rods.
This Westinghouse report to RGE addresses the fuel assembly design, fuel rod design and the loss-of-coolant-accident (LOCA) evaluations.
This report is planned to be part of an RGE submittal to the NRC.
The evaluations confirm that the standard U02 fuel design and LOCA safety criteria are. satisfied by the mixed oxide assemblies.
As part of the evaluations, the mixed oxide fuel rod design and anticipated duty are compared with the Westinghouse Region 7* standard U02 fuel manufactured in the same time period (1974) as the mixed oxide fuel.
The evaluations show that the mixed oxide fuel assemblies satisfy the LOCA safety criteria and also satisfy the fuel rod des'ign bases for burnups exceeding the anticipated three cycles of irradiation (Section 2.2).
There are no restrictions in core position as a result of the evaluations in this report.
- The mechanical design of Region 7 is the same as the most recently supplied
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2.0 DESIGN DESCRIPTION AND EVALUATION 2.1 Fuel Assembly The mechanical design of the mixed oxide (Pu02-U02) fuel asserrblies is the same as the Westinghouse supplied Region 7 standard U02 fuel assembly, except for the fuel isotopics and their positions in the assembly.
The basic design features and design bases of the standard fuel assemblies are described in Reference 1.
The mixed oxide assemblies have the same design bases as the standard U02 assemblies.
The fuel rods are positioned in the 14xl4 array as shown in Figure l.
The mixed oxide fuel rods are installed in a standard U02 assembly skeleton.
The 179 mixed oxide fuel rods are designed with three different nominal en-richments (3.20, 3.00 and 2.6 w/o)*, with the higher enrichments located in the center and the lower enrichments on the outside rows and in the corners.
This arrangement flattens the power peaking across the fuel assembly to an estimated peak/average power of about 1.06 at the beginning-of-life.
This aids the core designer in the prevention of excessive power peaking in the mi xed oxi de assemblies.
The mixed oxide fuel assemblies have an average reactivity equivalent U-235 enrichment of 3.1 w/o.
Table 1 gives the "typical" U-Pu isotopic weights for the three enrichment fuel rods.
2.2 Fuel Rod The mechanical design of the R.E.Ginna mixed oxide rods is nearly identical to the standard uranium dioxide rods in Region 7.
The mechanical components (tubing, end plugs, and plenum spring) are identical.. The pellet dimensions (length, diameter, dish geometry) are also identical.
Fuel den-sities are equal.
The mixed oxide. fuel rods are pressurized with He gas.
PuO Mixed oxide enrichment definition:
w/o =
p 0
+
UO x100
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I 0
T 0
0
The mixed oxide internal rod pressures are within the range of current practice and satisfy the fuel rod design criteria fol" core average burnups of at least 27,000 EFPH.
This is well in excess of the anticipated 21,000 EFPH accumu-lated during 3 cycles. with the mixed oxide asse'mblies in the core.
In particular, the rods were-pressurized to assure that clad flattening will not occur,-
and the rod internal gas pressure remains less than the system coolant pressure (2250 psi a nominal).
The material properties of Pu02-U02 and U02 are similar for the enrichments in the R.E.Ginna fuel.
A comparison of the characteristics of key properties is given.in Table 2.
The basis for these comparisons is described in Reference'.
As discussed in Section 3 of Reference 2, the Westinghouse fabri cation specification limits the maximum Pu02 particle size to 400'.
The average particle size is 10 to 20',
and 99.9X of the sizes are less than 100'.
This assures that potential. surface heat flux peaking due to large particles does not cause adverse effects on mixed oxide fuel performance under normal operation and accident conditions.
Based on the geometric similarities and the similarity in properties, the fuel performance characteristics of the mixed-oxide fuel will be almost the sana as the standard Region 7 U02 fuel.
This is supported by the Westinghouse experience with mixed oxide fuel rods in the Saxton and San Onofre
- reactors, described in Reference 2.
The mixed oxide fuel rods are designed to satisfy the design criteria described in Section 3.1.2 of the FSAR plus the additional criterion of no clad flattening.
The fuel rod design is evaluated with an approved fuel performance code to assure that the design criteria are met.
The models in the fuel performance code that are applicable'o uranium dioxide fuel are also applicable to mixed oxide fuel with two exceptions; the thermal con-ductivity and radial power depression factor of mixed-oxide fuel are slightly less than for U02 fuel.
Appropriate expressions or methods to evaluate these parameters are given in Reference 2, and these were used in the fuel performance code for this evaluation.
The FSAR design criteria limits address fuel tempera-
- tures, rod pressures, clad stress and clad strain.
The design criteria limits for the mixed-oxide fuel are the same as for the Region 7 standard fuel.
A t
reduction in the melting temperature of mixed oxide fuel was accounted for in the calculation.
~ l el 0
0
0
'Calculations were'erformed using the approved fuel performance code (3) with appropriate equations for mixed oxide fuel, and it was determined that all design bases were met for the R.E.Ginna mixed-oxide fuel.
Calculations were also made to determine the tim for clad flattening, using an approved model As previously discussed in this section, the predicted clad flattening tine greater than 27,000 EFPH assures that the no flattening criterion is satisfied during the mixed oxide irradiation in the Ginna re-actor.
The mixed oxide fuel rods have been in storage for approximately 5 years since they were manufactured in 1974.
The rods have been stored in sealed steel containers under atmospheric (air) conditions.
Effects of this storage has a negligible effect on the fuel rod materials and internal gas pressure compared to the as-fabricated rods.
The Zircaloy cladding is virtually immune to corrosion in air at the peak storage temperature (less than 230'F) which is much less than the oxidation corrosion temperature of concern (well above 500'F).
Table 1 gives the mixed oxide is'otopics as of November 1978 for typical fuel rod enrichments.
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3.0 LOCA EVALUATION The ECCS performance of plutonium oxide fuel loaded in the R.E.Ginna Nuclear Plant core is bounded by existing analyses considering standard uranium oxide fuel.
Since the fuel assembly designs are equivalent, the thermal-hydraulic transients calculated for uranium oxide fuel apply to plutonium oxide fuel as well.
Fuel performance parameters
[pellet tempera-ture, fuel dimensions, gap fill pressure, core peaking] are similar for the two fuels, so the stored energy present during a
LOCA will be basically the same.
Due to the difference in the fission products form d-by Pu02 compared to U02, the decay heat of plutonium oxi de in the first 200 seconds after shut-down is 8-20% lower than the corresponding uranium oxide value.
Comparisons are given in the proposed ANS Standard 5.1 Given the thermal-hydraulic (5) compatibility and equi valent fuel peramters, the significantly lower decay heat of plutonium oxide fuel makes it less limiting in any ECCS performance evaluation than the uranium oxide fuel whi ch has been previously analyzed for Ginna.
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- 4. 0 REFERENCES l.
USAEC Docket Number 50-244, "Final Facility Description and Safety Analysis Report",
Robert Eranet Ginna Nuclear Power Plant, Unit Number 1, January 1968.
2.
- Rim, C. S, et. al, "Performance.Characteristics of Mixed Pu02-U02 Fuels in Pressurized Water Reactors, WCAP 8349-P,
- February, 1975.
3.
Miller, J. V., ed.,
"Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations",
WCAP 8785, October 1976 4.
- George, R. A., et. al., "Revised Clad Flattening Model",
WCAP 8381, July 1974 I
5.
ANS Standard 5.1, "Decay Heat Power in Light Water Reactors",
October 1978.
Final Approval by ANSI - August 29, 1979.
I
~8 M
RCC GUIDE TUBES LEG END:
R 1
IHSTRUMEHTATIOH TUBE
~
~
'I medium-enri chment 1 ow-enri chment FIGURE ENRICHMENT POSITION FOR PuOp-UO>
FUEL ASSEMBLIES
I
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0 0
0 TABLE R. E.
Ginna Mixed Oxide Fuel Rods
- Typical Isotopics-
- Per Fuel Rod-Weight of Isotopic Material (Grams)
Isotopic Materi al Pu238 p 239 Pu240 Pu241 Pu242 Total Pu U235 Total U
Total (Pu02 U 2 Hi gh Enri chmen t 0.2 54.5
- 11. 0 2.3 0.5
- 68. 5 15.2 2143 2510. 5 Medium Enri chment 0.2
- 50. 7
- 10. 7 2.3 0.6
- 64. 5
- 15. 3
~- 2154 2517 Low Enri chment 0.5 41.0 8.2 3.5 0.8 54
- 15. 3 2161 2515 8-
~8
TABLE 2-CHARACTERISTICS OF Pu02-U02 COMPARED TO U02 Parade te r Yalue Pu02-U02 Melting Point Pu02-U02 Thermal Conductivity Pu02-U02 Radial Power Depression PU02 U02 Specific Heat Pu02-U02 Thermal Expansion DNBR and Clad Failure Threshold Energy (With Pu02 Particle Size 400 microns)
Slightly lower than UO Slightly lower than UO Greater than U02 Equivalent to UO Equivalent to UO Equivalent to U02 Pu02-U02 Fuel Densification Less than or equal to UO Pu02-U02 Fuel Swel ling Equivalent to U02 Pu02-U02 Fission Gas Release Equi val ent to '02
t i