ML19294A933
| ML19294A933 | |
| Person / Time | |
|---|---|
| Site: | 07001100 |
| Issue date: | 11/09/1979 |
| From: | Crow W, Ketzlach N NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 8002260158 | |
| Download: ML19294A933 (6) | |
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UNITED STATES y
p, NUCLEAR REGULATORY COMMISSION 3.g l
WASHINGTON, D C. 20555
+....$
NOV S 1979 DOCKET NO.:
70-1100 APPLICANT:
Combustion Engineering, Inc.
FACILITY:
Windsor, Connecticut
SUBJECT:
REVIEW 0F LICENSE AMENDMENT APPLICATION DATED APRIL 23, 1979, AND SUPPLEMENTS DATED JUNE 27, AUGUST 7, OCTOBER 17, AND OCTOBER 29, 1979 I.
Background
Combustion Engineering (CE) by application dated April 23,1979, and supplements dated June 27, August 7, October 17, and October 29,1979, has requested authori;;qgU to 4.1% 235U. All fuel fabrication process steps ion to increase the maximum enrichment of uranium from the present 3.5% "
are to remain the same. Section 19 of the license, the principal nuclear criticality safety license condition section, was revised in its entirety.
Exhibit C, parts of which contained license condition criteria beyond the general criteria in Section 19, was revised in gU enriched feel fabrication s entirety to include the more extensive criteria required for the 4.1% 2 operations. The entire revised Exhibit C will become a license condition section. Exhibit D, a demonstration section in support of the new license conditions, has also been revised in its entirety.
II.
Discussion A.
Nuclear Criticality Safety 1.
CE Independent Review CE, in its supplement dated October 17, 1979, provided the review by its Nuclear Safety Committee (NSC) of the amendaent application dated April 23,1979, and its supplements dated June 27 and August 7,1979.
The NSC review revealed no errors in the original calculations provided in the amendment application which would significantly influence conclusions about the safety of the facility operation at the higher enrichment limits. The 8002260\\5%,
2 review provided a critique of the application and the results of the safety review. Had this been performed before the submittal of the application to the NRC, an improved application would have resulted that would have facilitated NRC review.
2.
NRC Independent Review a.
Section 19 Section 19, " Nuclear Criticality Scfety Limits," provides the bases for safety of all simple geometry units and for all individual units that are safe independent of container geometry.
Spacing criteria between units are provided that are based on the surface density technique. The safety of the criteria has been confirmed by comparison with criticality data presented in ARH-600 (" Criticality Handbook"), DP-1014 (" Critical and Safe Masses and Dimensions of Lattices of U and U02 Rods in Water"), N16.1-1975
(" Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors"), and in WCAP-2999 (" Criticality Control Parameters and Supporting Calculations for Uranium Bearing Nuclear Fuel at Low Enrichment"). The safety margins for the mass, geometry and surface density limits meet the criteria approved in the present license and are consistent with good industrial practices.
b.
Exhibit C Exhibit C, " Nuclear Products Manufacturing Operations and Processes," contains the criticality safety criteria for the more complex geometry operations and for the interaction between maximum safe individual units that do not meet the surface density criteria. As mentioned above, the results of an independent criticality safety review made under the direction of the CE Nuclear Safety Committee are included in the application.
The Nuclear Safety Committee review includes a review of the methods of analysis, calculational models and the calculated results.
The NRC staff evaluated both the CE primary analyses and those of the independent review.
The staff review of the nuclear criticality safety of the operations can be divided into those that are based on (1) controlled moderation within a unit and (2) optimum water moderation within an individual unit, (3) neutron poisoning (8 C) as a secondary criticality control and 4
(4) interspersed water moderation.
Optimum interspersed water moderation was assumed whenever it could not be excluded.
(1) Controlled Water Moderation Although the moisture content of the U02 powder is <5 wt. %,
a maximum of 7% moisture content was assumed for all operations in the process flow from the receipt of the virgin U02 powder through the powder
3 batch make-up hood and its subsequent transfer to the U02 blenders.
This is conservative.
The reactivity for the 7% moisture systems is greater than that for 5%. Within the make acod itself, optimum water moderation was assumed since the containers are first opened in this hood. All undamaged containers of UO2 are stored in the virgin U02 storage vault and are sealed against moisture by the supplier. There is no source of water in the storage vault and administrative controls prevent water from sprinklers in adjacent areas from entering the vault. The staff determined the maximum reactivity of the storage array is 0.924 compared to the CE calculated reactivity of 0.934. The difference between the results of the two analyses are within the statistical limits of error in the analyses.
CE determined the maximum reactivity in the batch make-up area.
The reactivity of the array in the batch make-up array should be lower than that in the vault due to the greater spacing between units in the array.
This is confirmed by the licensee who calculated a maximum keff=0.860. Therefore, the fuel in both the virgin powder storage and batch make-up areas is sa fe.
(2) Optimum Water Moderation The nuclear safety of the powder blending, drying and grar.ulation operation is based on optimum water moderation within and between process units.
The analysis of this operation is divided into two parts:
(1) the blender-dryer section (back end) and (2) the dryer-granulation-hopper section (front end). The licensee determined the maximum keff=0.85 for the back end.
The staff confirmed the nuclear safety of the back end section:
(1) the blender spreader funnel is a safe volume for 4.1% enriched U02 and the dryer plenum powder depth is limited to 0.5 inches compared to 3.5 inches for the 3.5% enriched U0. The 3.0 inches difference in slab thickness more 2
than compensates for the 0.6% increase in enrichment.
The licensee's experience indicates very little accumulation of powder in the plenum. He will inspect the plenum weekly and clean it out if necessary.
Therefore, the controls are adequate to ensure the nuclear safety of the back end section.
To ensure the nuclear safety of the front end section, the licensee proposes to replace the front end hopper by one t!.at is safe by volume for 4.1% enriched UO2 powder wher, processing oxide having an enrichment >3.5%
2350.
The smaller hopper and the decrease in the maximum thickness of U0g in the plenum from 3.5 inches to 0.5 inches should provide an assembly that is no more reactive than the present assembly for 3.5% U0 enriched fuels.
This is confirmed by the calculations of the licensee and b his independent reviewer.
For processing oxides having an enrichment <3.5% 23 0, the larger hopper previously approved for axides <3.5% 235U wTil be used.
The larger hopper will be stored under lock and key prTor to processing oxides with an enrichment
>3.5% 235 U Slabs, safe independent of the degree of water moderation and reflection, are used in a number of process operations.
Maximum safe volumes and masses, independent of the degree of water moderation and reflection, are also used.
4 (3) Neutron Poisons CE proposes to use 8 C powder clad in ss tubes as a 4
secondary control to ensure nuclear safety in the pellet drying furnace.
In the absence of water in the furnace, an infinite quantity of UO2 cannot be made critical.
To ensure nuclear safety under accident conditions,16 B4C tubes are positioned in fixed locations within the drying furnaces. A quality assurance program has been developed to assure the presence of B4C in the tubes and that the 8 C meets the design criteria for nuclear safety.
4 Aluminum plates are welded to each end of the furnace positions holding the poison tubes to prevent their removal at any time. The tubes are inspected periodically to provide assurance of the continued presence of the B C.
CE has determined the maximum keff=0.87 under conditions 4
of interspersed moderation within and between drying furnaces. Although the model used in the analysis neglecting the aluminum pellet troughs in the furnaces and replacing them with the variable density water used in the analyses to determine the keff as a function of water density may be non-conservative over part of the range of water densities, the maximum calculated keff (0.87) at optimum water density is a measure of the subcriticality of the system at optimum moderation. The CE calculated maximum keff of 0.87 compares with a keff=0.88 calculated by the staff.
(4)
Interspersed Water Moderation Optimum interspersed water moderation between storage units in an array has been considered whenever it cannot be excluded.
It has been excluded only in the dry virgin powder storage and in the fuel rod storage areas. The former area is enclosed by concrete block walls, a double steel roof and a metal fire door. Althousn the fire door is normally in the open position, it is automatically closed upon activation of the fire alarm or on failure of electrical power.
The operation of the fire door under the emergency conditions specified is verified quarterly. The fuel rod storage area is enclosed on all four sides by sheet metal and by a corrugated fiberglass roof to assure the exclusion of water. As an added precaution against water flooding the boxes, no more than one rod storage box is in the open position at a time, and then only when attended by personnel, for the addition or removal of rods. Therefore, water moderation within and between boxes may be neglected.
CE calculated a keff=0.80 for the array.
This compares with a keff=0.89 calculated by the staff for a closely-packed infinite hexagonal lattice of rods flooded with water. Therefore, the array is safe.
The double shelf rod storage array is exposed to the water sprinkler system. Therefore, interspersed moderation is considered. The CE calculated keff=0.89 with no moderation within the boxes compares witn a keff=0.91 calculated by the staff ith water moderation within the boxes loaded with rods in a close-packed hexar aal array.
Therefore, the array is safe.
5 The pellet storage shelves, each safe by geometry (4.1 inches high) for 3.5% enriched UO, will be replaced by shelves at the 2
same center-to-center spacing 3.7 inches high that are safe by geometry for 4.1% enriched U0. CE calculated a reactivity of 0.82 for the array with 2
optimum interspersed moderation between shelves.
The staff concluded the array to be safe based on the similarity of the two fuels (3.5% and 4.1%
enriched UO ), the same center-to-center spacing between the shelves, and 2
the same safety factor in the thickness of the slab shelves for the two enrichmen ts.
The autoclaves contain a maximum of 32 rods each. The staff determined this number is less than 45% of the minimum critical number at optimum water moderation and reflection.
The surface density criteria confirm the safety of the array of autoclaves.
The nuclear safety of the 111 fuel assembly storage array has also been evaluated assuming optimum interspersed water moderation within assemblies and between assemblies.
CE calculated a maximum kgff=0.90.
An independent study made earlier by the staff indicates 130 assemblies cannot be made critical at the indicated spacing under optimum interspersed water moderation and reflection.
The shipping container for the 4.1% 235U enriched fuel has not been approved for the transport of fuel.
Since it has not been submitted as part of the SNM-1067 amendment application for use in fuel storage, it is recommended tMt - license condition be added to preclude the use of the containers for f al storas: "ntil approved by the Nuclear Regulatory Commission for the transport of tne fuel.
B.
Radiation Safety since the licensee does not propose any change in process operations, no significant effect in raciological safety should result. During our review of the amendment application, it was noted the requirement for a quarterly caitoration of the detectcrs for the criticality alarm system was not in a license condition section.
It has now been incorporated in a license condition section.
C.
Environmental Effi. cts Since the license does not propose any change in process operations, no significant change in environmental effect should result.
D.
General The amendment application was discussed with Mr. Jerome Roth, Region I (IE) Inspector of the CE facility on June 28, 1979, and on November 5,1979.
6 He foresaw no safety or environmental related problem with authorization to increase the maximum enrichr.c..t of uranium from the present 3.5% to 4.1% 235,
0 III.
Conclusion The controls associated with the processing of 4.1% 235U enriched fuel are adequate for the protection of the health and safety of the operating personnel, the public and the environment.
Issuance of the license amendment is recommended subject to the following condition:
Condition 26.
Notwithstanding the statements on page C-27 of the amendment application, the storage of fuel assemblies in shipping containers shall be made only in containers approved by the Nuclear Regulatory Commission.
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Norman Ketzlach Uranium Fuel Licensing Branch Division of Fuel Cycle and
', Y A' aterial Safety l0,,
(
Approved by:
W. T. Crow, Section Leader Uranium Process Licensing Section
.