ML20215J206
| ML20215J206 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/13/1986 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20215J181 | List: |
| References | |
| NUDOCS 8610240351 | |
| Download: ML20215J206 (6) | |
Text
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'O OESIGN FEATURES 5.2.1.2 REACTOR BUILDING a.
Nominal annular space = 5 feet.
b.
Annulus nominal volume = 427,000 cubic feet, Nominal outside height (measured from top of foundation base to the c.
top of the done) = 177 feet.
d.
Nominal inside diameter = 125 feet.
Cylinder wall minimum thickness = 3 feet.
e.
f.
Dome minimum thickness = 2.25 feet.
g.
Dome inside radius = 87 feet.
DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a j
maximum internal pressure of 15.0 psig and a temperature of 250*F.
5.3 REACTOR CORE FUEL ASSEM8 LIES The core shall contain 193 fuel assemblies with each fuel assembly con-5.3.1 taining 264 fuel rods clad with Zircaloy-4, except that limited substitutions y
of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel, or by q
vacancies, may be made in peripheral fuel assemblies if justified by cycle-I Each fuel rod shall have a nominal active fuel length '
Reload
' k specific reload analyses.
of 144 inches and contain a maximum total weight of 1766 grams uranium.
fuel shall be simila.' in physical design to the initial core loading and shall l
have a maximum enrichment of.3rS" weight percent U-235.
4.o CONTROL R00 ASSEMBLIES The core shall contain 53 full-length and no part-length control rod 5.3.2 The full-length control rod assemblies shall contain a nominal assemblies.
142 inches of absorber material.
The nominal values of absorber material for The Unit 1 control rods shall be 80% silver, 15% indium, and 5% cadmium.
nominal values of absorber material for Unit 2 control rods shall be 100%
baron carbide (8 C) for 102 inches and 80% silver, 15% indium, and 5% cadmium All control rods shall be clad with stainless steel 4
for the 40-inch tip.
tubing.
8610240351 861013 PDR ADOCK 05000369 P
PDR Amendment No. ?(Unit 1) f McGUIRE - UNITS 1 and 2 5-6 Amendment No f(Unit 2)
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JUSTIFICATION AND SAFETY ANALYSIS Based on analyses which indicate that fuel cycle lengths longer than 12 months will be more economical for Duke Power (longer cycles can facilitate higher nuclear capacity factors and lower Duke's overall generation costs), longer (330+ EFPD) fuel cycles are planned for McGuire Nuclear Station which require high enrichments. Since future reload fuel regions will have enrichments exceeding the current reactor core technical specification reload fuel maximum enrichment limit of 3.5 weight percent U-235 (enrichments between 3.5 w/o and 4.0 w/o will be required to realize the economic advantages of these longer cycles), the proposed amendments seek to revise the T.S. 5.3.1 maximum allowable fuel enrichment to 4.0 w/o for both units.
The 3.5 weight percent reload fuel limit in Technical Specification Section 5.3.1 is based on the criticality analyses for the original new and spent fuel storage facilities at the McGuire Nuclear Station (reference FSAR sections 4.2 and 4.3).
Since that time, the original spent fuel storage racks have been replaced with high density storage racks (reference MNS license amendments 35 (Unit 1)/16 (Unit 2)). These current spent fuel racks are described in FSAR Section 9.1.2, with the-storage capabilities governed by Technical Specification sections 3.9.12 and 5.6 which limit storage to fuel assemblies having an initial enrichment less than or equal to 4.0 w/o U-235 (vs. 3.5 w/o of the original racks). The criticality aspects of these current spent fuel racks with fuel enriched up to 4.0 w/o have been evaluated and deemed acceptable by License Amendment 35/16. The storage capabilities of the new fuel storage vault are governed by Technical Specification Section 5.6, and are further discussed in FSAR Section 9.1.1 (which indicates that the criticality evaluation performed for the new fuel storage vault assumes assemblies to be 3.5 w/o enrichment U-235 and unirradiated). In light of the planned higher enrichments for future reload fuel regions, the new fuel vault has been reanalyzed for normal storage and accident conditions to enable reception of fuel enriched to 4.0 w/o U-235 including a conservative margin for enrichment variance. This criticality reanalysis (which was performed in accordance with the criteria of T.S. 5.6) is summarized in Attachment 2A, and demonstrates a 4.0 w/o (4.1 w/o maximum) storage capability.
The reactor core is similarly capable of handling 4.0 (and higher) weight percent reload fuel.
This capability will be demonstrated as necessary in the cycle-specific reload, safety evaluations (RSE) which are performed prior to fuel loading (the RSE's consider the standard reload design methods described in WCAP-9272 and 9273, " Westinghouse Reload Safety Evaluation Methodology", and/or other appropriate criteria to demonstrate that the core reload will not adversely affect the safety of the plant). Criticality accidents during refueling operations are precluded by stringent administrative procedures which require count rate monitoring and sufficient boric acid concentration in the reactor vessel.
Consequently, based on the above evaluation it is concluded that use of reload fuel enriched up to 4.0 w/o U-235 at McGuire Nuclear Station is acceptable, and T.S. 5.3.1 is revised accordingly (the 4.0 w/o limit was chosen so that the new fuel vault / reactor core / spent fuel storage rack would have consistent limits). No other Technical Specification changes (e.g. to T.S. 3.9.12, 5.6, or Bases) are required for this enrichment upgrade.
- A NEW FUEL STORAGE VAULT CRITICALITY ANALYSIS
SUMMARY
Criticality analyses allowing up to 4.1 w/o U-235 enriched Westinghouse STD or OFA fuel in McGuire Nuclear Station's new fuel storage vaults have been performed in accordance with ANSI N16.9-1975.
Calculated values of K for the storage arrays, including the effects of calculationalandgeomekhicaluncertainties,arelessthanthoserequiredbyANSI N18.2-1973, Section 5.7.4.1 when a full loading of fuel assemblies is considered.
The computer codes and techniques utilized in the analysis have been validated against experimental data for water moderated UO lattices with characteristics 2
similar to the fuel analyzed.
The analysis method which ensures the criticality safety of 4.1 w/o fuel assemblies in the new fuel vault uses the CSAS2 criticality safety analysis sequency and the 123GROUPGMTH master cross-section library included in the SCALE-3 system of codes. CSAS2 consists of two cross-section processing codes (NITAWL and BONAMI), a 1-D transport code for cell-weighting cross-section data (XSDRNPM), and a 3-D monte-carlo code (KEND-IV) for calculating the ef fective multiplication factor of a system.
The following assumptions were made in evaluating criticality safety:
a.
No credit was taken for the inherent neutron-absorbing effect of the new fuel storage rack materials.
b.
No burnable poisons, control rods, or supplemental neutron poisons are assumed to be present.
c.
Effects of reflectors other than water are included if their neglect would have been nonconservative.
This includes the storage vault's concrete walls, ceiling, and floor.
d.
All assemblies are assumed to be 4.1 w/o U-235 enriched and unirradiated.
e.
The new fuel storage vault is modeled as 2 rooms which are separated by a 2-foot thick concrete wall.
Each room contains 3 infinite rows of 12-foot high fuel assemblies.
(See FSAR Figures 9.1.1-1 and 9.1.1-2) f.
Each fuel assembly is treated as a heterogeneous system with the fuel pins, control rod guide tubes, and instrumentation thimble guide tube modeled explicitly.
The following accidents are considered in the criticality design of the new fuel storage vault:
Oak Ridge National Laboratories, " SCALE-3: A Modular Code System for Performing Standardized Computer Analysis for Licensing Evaluation," NUREG/CR-0200-Vols.
1-2-3-Bk.
4, Pevision 3 December 1984,
]
-- ' A Page 2 a.
Flooding: complete immersion of the entire array in pure, unborated room temperature water, b.
Envelopment of the entire array in a uniform density aqueous foam of optimum moderation density (that density which maximizes the reactivity of the array). This accident could occur as a result of fire fighting.
Accidents resulting in an increase in K because of geometrical changes of the ff racks or fuel handling accidents are nol considered credible due to the following design bases:
a.
The facility is designed in accordance with GDC's 2 and 4.
b.
The racks are designed to Seismic Category 1 requirements, c.
The only Category 1 structure that could disrupt the array should it fail during a seismic event is the crane trolley. Administrative procedure prohibits the trolley from being parked over the new fuel storage vault.
d.
The runway conductors for the trolley are divided and power to each section is provided through separate circuit breakers. Power to the conductors in the area of the new fuel storage vault is provided only during handling operations. The conductors are divided at a point which will prohibit the trolley being positioned over the vault when power to that end is interrupted, The racks and anchorages can withstand the maximum uplift force e.
available without a significant change in geometry.
f.
The design of the Fuel Handling System and administrative procedures insure suberitical spacing of fuel assemblies.
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ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION As required by 10CFR50.91, this analysis is provided concerning whether the proposed amendments involve significant hazards considerations, as defined by 10CFR50.92. Standards for determination that a proposed amendment involves no significant hazards considerations are if operation of the facility in accordance with the proposed amendment would not:
- 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the
. possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.
The proposed amendments would allow use of higher enrichment (up to 4.0 w/o U-235) fuel for future McGuire unit 1 and 2 core reloads. The current technical speci-fication limit of 3.5 w/o is based on the criticality analyses for the original new and spent fuel storage facilities at the McGuire Nuclear Station.
Since the original spent fuel storage racks have previously been replaced with high density storage racks which allow use of fuel having an initial enrichment less than or equal to 4.0 w/o U-235 (reference McGuire license amendments 35 (Unit 1)/16 (Unit 2)), only the criticality analyses for the new fuel vault are affected (use of the-higher enrichment fuel in the reactor core will be demonstrated to be acceptable via the cycle-specific reload safety evaluations which are performed prior to fuel loading - criticality accidents during refueling operations are precluded by stringent administrative procedures). A reanalysis (see Attachment 2A) of the new fuel vault normal storage and accident conditions performed to support the revised enrichment limits similarly demonstrates a 4.0 w/o storage capability.
No significant hazards considerations need be addressed with respect to the spent fuel storage facility since use of the higher enrichments is already allowed for it.
The above referenced new fuel vault criticality evaluation demonstrates that use of the higher enrichment fuel does not involve a significant increase in the consequences of an accident previously evaluated or involve a significant reduction in a margin of safety (as the reactor core is capable of handling 4.0 (and higher) weight percent reload fuel similar conclusions are anticipated to.be documented in the cycle-specific RSE's).
Since only the fuel enrichment is being changed no accident causa1' mechanisms are affected or created and consequently the probability of an accident previously evaluated is unaffected and no new or different kinds of accidents from any accident previously evaluated can be created.
The commission has provided examples of amendments likely to involve no signifi-cant hazards considerations (48FR14870). One example of this type is (iii) "For a Nuclear Power Reactor', a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously accep-table to the NRC for a previous core at the facility in question are involved.
This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are not signifi-cantly changed, and that NRC has previously found such methods acceptable." Since the proposed reload fuel assemblies differ from previously approved reload core I
assemblies only in a slight increase in enrichment levels, the acceptance criteria 1
a Page 2 for the technical specifications is met (e.g. compliance with T.S. 5.6.1 criteria was demonstrated by the new fuel vault reanalysis) with the exception of the proposed amendment itself, and the analytical methods used to demonstrate con-formance of the new fuel vault with the technical specifications and regulations (as well as the methods anticipated to be used in the cycle-specific RSE's) are those previously used/ acceptable to the NRC, the above cited example can be applied to this amendment. In addition, another example of actions not likely to involve a significant hazards consideration is (vi), "A change which either may result in some increase to the probability or consequences of a previously analy-zed accident or may reduce in some way a safety margin, but where results of the change are clearly within all acceptable criteria with respect to the system or component specified in the standard review plan:
for example, a change resulting form the application of a small refinement of a previously used calculational model or design method". Because the evaluations previously referenced for the new and spent fuel storage facilities show that all acceptable criteria are met, and similar conclusions are expected from the cycle specific RSE's, this example also applies.
Based upon the preceding analyses, Duke Power Company concludes that the proposed amendments do not involve a significant hazards consideration.
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