ML20085F875
| ML20085F875 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/18/1991 |
| From: | Hairston W ALABAMA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9110230084 | |
| Download: ML20085F875 (20) | |
Text
,
Alabama Power Company 40 frwrnas Center Parkway Porit Othee Dox 1295 Dermingham, A!ubama 35201 Telept one 205 808 5581 i
W. G, Hairston, lit sannr vo P,esent Nudem Opereous fM{jf)[jj));}f()g[
twwm mn wiem Docket Nos. 50-348 10 CFR 50.90 50-364 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, DC 20555 Joseph M. Farley Nuclear Plant Fuel Enrichment Increase and VANTAGE-5 Fuel Design Amendments Senlemental Information Gentlemen:
By letters dated July 1,1991, and July 15, 1991, Alabama Power Company proposed to amend the Farley Nuclear Plant (FNP) Units 1 and 2 Technical Specifications. The proposed amendments include provisions which allow for continued extended burnup levels to 60 GWD/HTU, an increase in reload fuel enrichment fiom 4.3 weight percent U-235 to a nominal 5 weight percent U-235 (corresponds to 5.05 w/o maximum, which includes a 0.05 w/o manufacturing uncertainty), and a change in reactor core fuel type from Westinghouse LOPAR fuel to Westinghouse VANiiAGE-5 fuel. The environmental effects of extended burnup and higher initi.il enrichments to these levels have previously been addressed by the NRC [e.g., Millstone Unit 3 (54 FR 38772)].
As discussed in a telephone call between Alabama Power Company and Steve Hoffman of your staff on September 11, 1991, Alabama Power Company hereby supplements the information contained in the above referenced submittals, as requested by the NRC, with the additional information contained in Attachment 1 to this letter.
In addition. Attachments 2 and 3 contain a revision to the July 1, 1991 sutmittal, updating the Technical Specifications for Units 1 and 2, and the Significant Hazards Consideration Analysis, respectively.
The revisions were hiade to reflect the nominal enrichment limit in the Technical f;pecifications, rather than the maximum.
The Westinghouse report,
<t,O Q
\\
'~' ' ' " \\ ' U 9110230084 911010 PDR ADOCK 05000348 p
U. S. Nuclear Regulatory Cot,nission Page 2
" Criticality Analysis of the Farley Units 1 and 2 Fresh and Spent Fuel Racks," originally submitted with the July 1,1991 package, remains applicable and is not updated.
Respectfully submitted, ALABAMA POWER COMPANY Id,
/dru ND 'IE~
W. G. Hairston,-III WGH, III: RAH / map 0984 Attachments 1,2,3-SWORN TO AND SUBSCRIBE 0 BEFORE ME cc: Mr. S. D. Ebneter THIS
/N' DAY OF O(/2/cc1991 htL& $wU'i$ll h G.' F.'
1 Dr. C. E. Fox-Wotary Public" My Commission Expires: /A/J!94L l
l t
ATTACHMENT 1 ENVIRONMENTAL EVALUATION i
I
JOSEPH M. FARLEY NUCLEAR PLANT rHEL ENRICHMutT INCREASE ANC v AGE-5 FUEL DESIGN AMENDMENTS
[SyJB@ MENTAL EVAlyK[10]
The generic environmental effects of the uranium fuel cycle are provided in Table S-3 of 10 CFR 51.51, " Uranium fuel Cycle Environmental Data," and the environmental impact of transportation of the fuel and waste to and from a reference reactor i.; providad in Table S-4 of 10 CFR 51.52,
" Environmental Effects of Transprtaton of fuel and Waste." These environmental assessments provided in Tables S-3 and S-4 are applicable for a fuel burnup to 33 GVD/MTU and fuel enrichment to 4 weight percent U-235.
The use of fuel enriched to 4.3 weight percent U-235 to accommodate fuel management schemes associated with 18 month fuel cycles a.t-Farley Units 1 and 2 was approved by the NRC in License Amendments 52 and 43, respectively, dated November 8, 1984, in conjunction with the conversion to VANTAGE-5 fuel, Alabama Poiver Company anticipates extended fuel burnups to 60 GWD/MTU and the use of fuel with initial nominal enrichments to 5 weight percent U-235.
The environmental effects of extended burnup and higher initial enrich-ments to these levels have previously been addressed by the NRC.
A hutice published in the Federal Register on February 29, 1988 (53 FR 6054),
states that the NRC's environmental assessment of extended fuel burnup is complete and that the environmental impacts summarized in Table S-3 of 10 CFR 51.51 and in Table S-4 of 10 CFR 51.5? bound the corresponding impacts for burnup levels up to 60 GWD/HTU and enrichments up to 5 weight percent U-235.
The environmental impacts of transportation resulting from the use of extended burnup and higher enrichment fuel were further addressed in the Federal Register on August 11, 1988 (53 FR 30355). This notice reiterated the conclusion stated in 53 FR 6054 and further concluded that there are no significant adverse radiological or non-radiological impacts associated with the use of extended burnup and/or increased enrichment, and that burnup levels to 60 GWD/MTV and enrichments to 5 weight percent U-235 will not significantly affect the quality of the human environment. Moreover, pursuant to 10 CFR 51.31. " Determinations Based on Environmental Assessment," the NRC determined that an environmental impact statement need not be prepared for this action.
As stated in 53 FR 30355, the NRC is in the process of revising the regulations in 10 CFR 51.52 to reflect the findings published in 53 FR 6054.
In the interim, in connection with its review of proposed licensed amendments'to permit use of fuel enriched to 5 weight percent U-235 and
.burnup levels to 60 GWD/MT4J, and pursuant to 10 CFR 51.52(b), the NRC has elected to accept the analysis of environmental effects of the transportation of such fuel and waste provided in 53 FR 30355 until such time as the revision to the rule is issued. 1 l
l J
a
1 i
i Alabama Power Company has determined'that the proposed license amendments are bounded by the NRC generic environmental assessment provided in 53 FR 6054 and 53 FR 30355.
Accordingly, there are no significant radiological or non-radiological environmental impacts associated with the proposed amendments ~.
Therefore, it can be concluded that a plant specific environmental assessment need not be prepared for this action.
-2
ATTACHMENT 2 Unit 1 Revision Page 5-6 Replace Page 5-7 Repl ace Unit 2 RevisioD.
Page 5-6 Replace Page 5-7 Replace e
I
(
_.. ---_-_ __- _ _ _ _ - _ _ ____ - _ __.___- - __.- _ _ - --____- -- _ -_._ - _- -. _ --_ _ _ w
DESIGN FEATURLS 5.3 REAf10R CORE
[ VEL ASSEMBLIES 5.3.1 The reactor core shall contair; 157 fuel assemblies with each fuel assembly containing 264 fuel rods ;1ad with Zircaloy-4.
Each fuel rod shall have a nominal active fuel length of 144 inches.
The initial core loading shall have a maximum nominal enrichment of 3.15 weight percent U-235.
Reload fuel shall be similar in physical design to the initial core loading and s'all have a maximum nominal enrichment of A.25 weight percent U-235 n
for Westinghouse LOPAR fuel ar.d a maximum nominal enrichment of 5.0 weight percent U-235 for Westinghouse OfA and VANTAGE-5 fuel.
Westinghouse OfA-and VANTAGE-5 fuel with maximum nominal enrichments greater than 3.9 weight percent 'J-235 shall contbin suf ficient integral burnable aosorbers such that the requirements of specifications 5.6.1.1.c and 5.6.1.2.c 3rc met.
Westinghouse LOPAR fuel does not require integral burnablo absorbers.
CONTROL ROD ASSEMBLIES 5.3.2 1he reactor core shall contain 48 full length and no part length control rcd assemblies.
The full length cor. trol rod assemblies shall contain a nominal 14211ches of absorber material.
The nominal value, of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium.
All control rods shall be clad with stainless steel tubing.
LLRfACTOR 200t ANT SYS1ff1 QLSIGN PRESSE AND TEME[MIi/EE 5.4.1 The reactor coolant system is 'iesigned and shall be maintained:
a.
In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a temperature of 6500F, except for the pressurizer which is
- 6800f, MME 5.4.2 The total water and steam volume of the reactor coolant system is 9723 1 100 cubic feet at a nominal Tavg of 5250F, 5.5 HETEOR010GICAL TOWER IQ W 10f]
5.5.1 The meteorological tower shall be located as shown on figure 5.1-).
FARLEY< UNIT 1 5-6 AMENDMENT N0.
4 NI I
E Ir YN+ A
l ELilG!LLLAIMLS -
5.6 FULL STOR40[
LRL11CALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with a Ke f less than or equal to 0.95 when flooded with unborated water, e
which includes conservative allowances for uncertainties and biases.
This is assured by maintaining:
a.
A nominal 10.75 inch center-to-center distance between fuel assemblies olaced in the storage racks, b.
A maximum nominal enrichment of 4.25 veight percent U-235 for Westinghouse LOPAR fuel assemblies.
c.
A maximum reference fuel assembly Ke less inan or equal to 1.455 at 680F for Westinghouse OfA e.nd VANIAGE-5 fuel assemblies.
5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a Keff lest than or equal to 0.98, assuming aqueous foam moderation.
This is assured by maintaining:
a.
A nominal 21 inch center-to-center distance between new fuel assemblies placed in the storage racks, b.
A maximum nominal enrichment of 4.25 weight percent U-235 for Westinghouse LOPAR fuel assemblies, c.
A maximum reference fuel assembly K less than or equal to 1.455 at 680f for Westinghnuse OfA and VANTAGE-5 fuel assemblies.
ERAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 149.
CA"allll 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1407 fuel assemblies.
17 COMPONENT CYCllC OR TBA!iE11NT LlHIT 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of lable 5.7-1.
FARLEY-UNIT 1 5-7 AMENCMi'h
'0.
QESIGN FEAE R[$_
5,J REACTOR COR[
FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy-4.
Each fuel rod shall 4
have a nominal active fuel length of 144 inches, lhe inillal core loading shall have a maximum nominal enrichment of 3.15 weight percent U-235.
Reload fuel shall be similar in physical design to the initial core loading and shall heve a maximum nominal enrichment of 4.25 weight percent U-235 for Westinghouse LOPAR fuel and a maximum nominal enrichment of 5.0 weight percent U-235 for Westinghouse OfA and VANTAGE-5 fuel. Westinghouse OfA and VANTAGE-5 fuel with maximum nominal enrichments greater than 3.9 weight percent U-235 shall contain sufficient integral burnable absorbers such that the reauirements of specifications 5.6.1.1.c and 5.6.1.2.c are met.
Westinghouse LOPAR fuel does not require integral burnable absorbers.
CONTROL ROD ASSEMBLI[5 5.3.2 The reactor core shall contain 48 full length and no put length control rod assemblies. The full length control rod assciublies shall contain a nominal 142 inches of absorber material.
The nominal values of absorber material shall be 80 percent silver.15 percent indium and 5 percent cadmium. All control-rods shall be clad Jith stainless steel tubing.
5.4 REACTOR COOLANT _SYSTE.M QESIGN PRESSURE AND TEMPERA 10E
- 5.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 5.2 of the FSAR With allowance for normal degradation pursuant to ti i applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a temperature of 6500F, except for the pressurizer wMch is
- 6800F, VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9723 i 100 cubic feet at a nominal Tavg of 5250f.
5.5 MET [QROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on figure 5.1-1.
i' FARLEY-UNIT 2 5-6 AMENDMENT N0.
. ~
.. m
QLSIGN F(MVRES L6 fVEL SlD MG[
ER111[A. Lily 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with a Keft less than or equal to 0.95 when flooded with unborated water, which includes conservative allowances for uncertainties and biases.
This ia assured by maintaining:
a.
A nominal 10.75 inch center-to-center distance between fuel assemblies placed in the storage racks, b.
A maximum nominal enrichment of 4.25 weight percent U-235 for Westinghouse LOPAR fuel assemblies.
c.
A maximum reference fuel assembly K~ less than or equal to 1.455 at 680f for Westinghouse OfA and VANTAGE-5 fuoi assemblies.
5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a Keff less than or equal to 0.98, assuming aqueous foam moderation.
1his is assured by maintaining:
a.
A nominal 21 inch center-to-center distance between new fuel assemblies placed in the storage racks, b.
A maximam nominal enrichment of 4.25 weight percent U-235 for Westinghouse LOPAR fuel assemblies, c.
A maximum reference fuel assembly K. less than or equal to 1.455 at 680f for Westinghouse OfA and VANTAGE-5 fuel assemblies.
DRaitt%E 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 149.
CAPAC11Y 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capa city limited to no more than 1407 fuel assemblies.
5.7 COMPONENT CY1LIC OR_]RANSlENT LIMil 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.
I FARLEY-UNIT ?
5-7 AMF.NOMENT NO.
unwnexArpma -
1 4
l ATTACHMENT 3 r
F. VALUATION OF NO SIGNIFICANT HAZARDS CONS 10 ERAT.10N I
t l
i r
6 l
h l:
I
{
~
p i
L l
1
-t i
a
i JOSEPH H. FARLEY UNilS 1 & 2 CRITICAll1Y ANALYSIS FOR NEW AND SPINT FUEL RACKS SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS PROPOSED CHANGE It is propoeed to update Technical Specification 5.3 (Reactor Core) and Technical Spixifie tion 5.6 (fuel Storage) to increase enrichments to a nominal 5.0 n ye for W-l ed fuel Assemblics (OfA) and for VANTAGE-5 fuel assemolics,.akir.9 c,ed81 f or the presence of Integral fuel Burnable Absorbers (Ift'O.
Ibn ;orichment limits for initial core loading and for low Parasitic (Le$ fuel are changed to reficct the nominal enrichment to be consistent with the new OfA/ VANTAGE-5 lim a.
The enrichment manUfactt; ring une:rtainty is 0.05 w/o.
Thus, the current initial core loading limit of 3.2 w/o maximum enrichment" is changed to 3.15 w/o
" maximum nominal enrichment" and the current reload fuel (LOPAR) limit of 4.3 w/o " maximum enrichment' is changed to 4.25 w/o " maximum nominal enrichment."
BACKGROUND Currently, Josepn H. farley Nuclear Plant Uu;ts 1 and 2 use Westinghouse 17x17 Low Parasitic (LOPAR) fuel. As part of a long-term fuel management strategy for f arley Nuclear Plant Units 1 and 2, Alabama Power Company plans to use the Westinghouse VANTAGE-5 fuel design in both units starting with Unit 2 Cycle 9 and Unit 1 Cycle 12.
In support of this effort, plant-specific evaluations and analyses have been performed to qualify the new fuel and spent fuel storage racks for storage of VANTAGE-5 fuel.
These evaluations and analyses also qualify the use of OfA fuel in the new and spent fuel racks for possible future use of OfA fuel.
These evaluations and analyses include OfA and VANTAGE-5 fuel nominal enrichments of up to 5.0 w/o in order to provide the maximum flexibility in core designs.
Support for implementing the r2 analysis is ba ed on a plant-specific criticality analysis of the spent fuel racks taking credit for the presence of Integral fuel Burnable Absorber (ifBA in OfA or VANTAGE-5 fuel.
- Thus, Alabama Power Company proposes that the )arley Nuclear Plant Units 1 and 2 f
lechnical Specifications be amended by revising the design features requirements to allow for nominal enrichments of up to 5.0 e/o U-235 for OfA and VANTAGE-5 fuel in the reactor core and spent fuel and new fuel pit storage racks. The enrichment limit for initial core fuel and reload LOPAR fuel is changed to reflect the nominal enrichment to be consistent with the new OfA/ VANTAGE-5 limit.
No other change to the initial core loading limit or reload LOPAR fuel limit is made.
There is no requirement for ifBA for LOPAR fuel.
A criticality analysis of the spent fuel racks at the Joseph H. f arley Nuclear Plant Units 1 & 2 was performed taking credit for the presence of Page 1 of 9
1 Integral fuel Burnable Absorber (IfBA) in OfA and VANTAGE-5 fuel.
The analysis of the spent fuel racks shows that Westinghouse 17x17 OfA and 17x17 VANTAGE-5 fuel assemblies with nominal enrichments of up to 3.9 w/o can be safely stored in the Joseph H. farley spent fuel racks utilizing all locations.
OfA and VANTAGE-5 fuel with nominal enrichments above 3.9 w/o and up to 5.0 w/o can also be stored in the spent fuel racks provided credit for ifBA is taken.
1he fuel assembly ifBAs consist of neutron absorbing material applied as a thin ZrB2 coating on the outside of the 002 fuel pellet. As a result, the neutron absorbing material is a non-removabic or integral part of the fuel assembly once it is manufactured. No IfBAs are required to support the t0 PAR enrichment limit.
It has also been demonstrated that Westinghouse 17x17 OfA and 17x17 VAN 1 AGE-5 fuel assemblies with nominal enrichments of up to 4.8 w/o can be safely stored in the Joseph H. f arley new fuel racks utilizing all locations.
Storage of OfA and VANTAGE-5 fuel assemblies in the new fuel racks with nominal enrichments above 4.8 w/o and up to 5.0 w/o U-235 is shown to be acceptable by taking credit for the same ifBAs which are required to satisfy the spent fuel rack limit.
Na IFBAs are required to support the LOPAR enrichment limit.
As a result of the above enrichment change, the Joseph H. f arley Units 1 and 2 Technical Specifications Section 5.3, Reactor Core (fuel Assemblies), and Section 5.6, fuel Storage (Criticality), are proposed to be updated.
The Joseph M. f arley Units 1 & 2 spent fuel rack criticality analysis is based on maintaining Keff less than or equal to 0.95 for storage of fuel assemblies. The new fuel rack analysis is based on maintaining Keff less than or equal to 0.95 for storage of fuel assemblies under full water density conditions and less than or equal to 0.98 under low water density (optimum moderation) conditions.
The new and spent fuel rack criticality analyses form the analysis basis for the enrichment limit in the " Reactor Core" section (Section 5.3.1) of-the Technical Specifications.
Enrichment is not a key safety parameter with respect to core operation, for core operation, reload designs (including fuel enrichments) are evaluated to confirm that the cycle core design adheres to the safety limits that exist in the current accident analysis and plant Technical Specifications. This is the only impact of enrichment on core operation.
Thus, it is proposed that this section is changed to reflect the new rack criticality analyses.
As rquired by 10 CFR 50.91(a)(1), this analysis is provided to demonstrate that a proposed license amendment to implement the criticality analysis for new and saent fuel racks and to modify the reactor core enrichment limit for Josep1 H. f arley Nuclear Plant, represents a no significant hazards consideration, in accordance with the three-factor test of 10 CFR 50.92(c), implementation of the proposed license amendment was analyzed using the following standards and found not to 1) involve a significant increase in the probability or consequences for an accident previously evaluated 2) create the possibil;ty of a new or different kind of accident from any accident previously evaluated, or 3) involve a significant reduction in a margin of safety.
Page 2 of 9
ANALYSIS Criticality of fuel assemblies in a fuel storage rack is prevented by the.
design of the rack wHch limits fuel assembly interaction.
This is done by l
establishing the m0 num separation between fuel assemblies and, for the spent fuel racks, aerting neutron poison between fuel assemblies.
l The design basis for preventing criticality outside the reactor is that, ine'uo'ing uncertainties, there is a 95 percent probability at a 95 percent con.%ence level that the effective neutron multiplication factor, Keff, of I
the fuel assembly array will be less than 0.95 as recommended b," ANSI 57.2-1983, ANSI $7.3-1983 and the NRC letter to power reactor licensaes,
" Position for Review and Acceptance of Spent fuel Storage and llandling Appilcations," April 14, 1978.
The 0.95 Keff limit applies to both the new
- and spent fuel racks under all conditions, except for the new fuel rack under low water density (optimum moderation) conditions, where the Keff limit is 0.98 as recommended by NUREG-0800 Criticalitv Caleulation Methodoloav The criticality calculation method and cross-section values are verified by comparison with critical experiment data for fuel assemblies similar to those for which the racks are designed.
This benchmarking data is sufficiently diverse to establish that the method bias and uncertainly will apply to rack conditions which include strong neutron absorbers, large 1
water gaps, and low moderator densities, If,BA Credit Reactivity Eouivalencing l
A base case analysis was performed to determine the maximum OfA and VANTAGE-5 fuel enrichment that may be stored while meeting the requirements above.
Storage of spent 0FA and VANTAGE-5 fuel assemblies with initial enrichments higher than that obtained in the base case is achievable by means of the concept of reactivity equivalencing.
Reactivity equivalencing is predicated upon the reactivity decrease associated with the-addition of IFBA fuel rods and fuel depletion. A series of reactivity calculations were performed to generate a set of IFBA-rod-number versus enrichment ordered pairs which all yield an equivalent Keff when the OfA or VANTAGE-5 fuel is stored in the Joseph M. Farley Units 1 & 2 spent fuel racks.
Criticality Analysis of jinent Fugl_RAGk l
Storage of 17x17 OfA and VANTAGE-5 fuel assemblies in the spent fuel rack is acceptable with nominal enrichments of up to 3.9 w/o U-235, and storage of 17x17 0FA and VANTAGE-5 IFBA assemblies is acceptabic with nominal enrichments above 3.9 w/o and up to a nominal 5.0 w/o U-235.
The analysis included sensitivity calculations for enrichment, cell center-to-center spacing, and poison loading.
No IfMs are required to support the LOPAR enrichment limit.
Page 3 of 9
. ~.
The following accident conditions will act result in an increase in Keff of the rack: loss of cooling systems (reactivity decreases with decreasing water density), dropping a fuel assembly on top of the rack (the rack structure pertinent for critichlity is not excessively deformed and the dropped assembly has more than ten inches of water separating it from the active fuel height of stored assemblies which precludes interaction), and dropping a fuel assembly between rack modules or between a rack module and wall (the design cf the s)ent fuel rack is such that it precludes the insertion of a fuel assemaly into other than prescribed locations).
However, acciocrts can be postulated which would increase reactivity (i.e.,
misplacing an assembly in an unqualified position in the spent fuel rack, misloading an assembly with an enrichment and IFBA combination outside of the acceptable limits, or dropping a fuel assembly into an already loaded cell). Hisplacing an assembly in the spent fuel rack is not considered credible for the Joseph M. Farley 1_ & 2 s)ent fuel racks, since the same rack design and limits apply throughout tie entire spent fuel pool.
The requirements of the spent fuel rack IFBA limit for 0FA and VANTAGE-5 fuel will become a design constraint on future Joseph H. Farley reload core designs, and multi-layered fuel vendor quality assurance controls on design, manufacturing, and shipment provide assurances that a potentially violating fuel assembly will not be delivered to the site.
For the accident of dropping of a fuel assembly into an already loaded cell, the upward axial leakage of that cell will be reduced; however, the overall effect of rack reactivity will be insignificant.
Criticality Analysis of New fuel Racks Since the new fuel racks are maintained in a dry condition, the criticality analysis shows that the rack Keff is less than 0.95 for the full water density condition and less than 0.98 for the low water density (optimum moderation) conditions.
Under nor.nal conditions, the new fuel racks are maintained in a dry environment.
The introduction of water into the new fuel rack area is the worst case accident scenario. The full density and low density optimum moderation cases are bounding accident situations which result in the most conservative fuel rack Keff.
For both cases, Keff retains within the acceptance limits.
Other accidents can be postulated which would increase reactivity (i.e.,
dropping a fuel assembly between the rack and wall or on top of the rack).
For these accident conditions, the double contingency principle is applied.
This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident.
Thus, for these accident conditions, the absence of a moderator in the new fuel storage racks can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event.
Westinghouse generic studies have shown the maximum reactivity increase for postulated accidents (such as those mentioned above) would t'e less than 10 Page 4 of 9
percent delta-k.
Furthermore, the normal, dry, new fuel rack reactivity is less than 0.70.
As a result, for postulated accidents, the maximum rack Keff will be less than 0.95.
epp_1.lcation of soent Fuel RJtek IFBA Credit limit to New FueLRack The acceptance criteria for criticality is satisfied for the storage of i
Westinghouse _17x17 0FA and VANTAGE-5 fuel assemblies with nominal enrichments of up to 3.9 w/o in the spent fuel storage racks.
lo allow storage of 0FA and VANTAGE-5 fuel assemblies with higher enrichments, minimum IFBA requirements were developed based on a reactivity equivalencing analysis.
For the new fuel storage racks, the acceptance criteria for criticality is satisfied for the storage of Westinghous217x17 0FA and VANTAGE-5 fuel assemblies with nominal enrichments up to 4.8 w/o.
This limit can be increased to-nominal enrichmonts of 5.0 w/o for 0Fli and VANTAGE-5-fuel in the new fuel racks oy taking credit for the :ame !FBA that is present-and taken credit for in the spent fuel racks.
Of the two storage rack limits (new and pent), the spent fuel rack li' nit is the more restrictive since OFA and VANTAGE-5 fuel assemblies are limited to the equivalent reactivity of an 0FA or VANTAGE-5 nominal 3.9 w/o fuel assembly.
Hence, it is conservative to use the spent fuel rack enrichment-IFBA limit for the new fuel storage rack.
Other Considert.tions The storage of 0FA or VANTAGE-5 fuel in the new and spent fuel racks will not adversely affect the seismic response of the racks.
17x17 LOPAR, OFA, and VANTAGE-5 fuel assemblies are structurally equivalent for the fuel storage rack seismic analyses.
Heat. load calculations performed for 0FA and VANTAGE-5 fuel show that the spent fuel pool decay heat load assumed in the current analytis remains bounding for 0FA_and VANTAGE-5 fuel. The clad temperature will increase by at mnst lof due to increased heat flux with the smaller rod diameter. This small increat i will not challenge fuel integrity.
Storage of 0FA and VANTAGE-5 fuel in the new and spent iuel racks does not involve any changes that would affect the releases of radiological effluent during normal operation nor the radiological consequences of postulated accidents reported in the FSAR.
The VANTAGE-5 fuel design does include
-features to accommodate extended fuel burnup; however, extended fuel burnup has previously been incorporated at Farley Nuclear Plant Units 1 & 2.
-Extended fuel _ burnup was evaluated for both_ units _, and _ fuel containing extended burnup features was incorporated into operation in Cycle 10 of Unit I and Cycle 7 of Unit 2.
Implementation of extended fuel burnup on Farley Units 1 & 2 considered the potential impact on accident source terms.
The topical report, " Extended Burnup Evaluation of Westinghouse fuel," (WCAP-10125-P-A) was used as the evaluation basis. This report, which was reviewed and approved by the NRC, demonstrated that there is very little effect on the source terms and that Page 5 of 9
m i
radiological consequences of accidents are not significantly affected.
lhe NRC, in their SER regarding this report, added discussion specifically in regard to the fuel handling accident.
There will be no significant increase in dose rate as a result of the use of VANTAGE-5 or OfA fuel with nominal enrichments of up to 5.0 w/o.
Evaluations have shown that no change to existing radiation controls is required.
Conformance of the proposed amendments to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 is shown in the following:
A.
IDIIcMcLDuel Enrichment for Rejttloe Core 1)
Operation of Joseph M. farley Units 1 & 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated because the applicable safety limits are within the bounds previously established.
Neither actuation of safety systems nor accident mitigating capabilities are adversely affected by operation of the plant in accordance with the proposed license amendment.
The analysis demonstrates that the proposed amendment does not pose a challenge to installed safety systems.
Therefore, no new performance requirements are bti,ig imposed on any system or component important to safety such that any design criteria will be exceeded.
The implementation of the criticality reanalysis is not an initiator for any of the postulated FSAR accidents analyzed.
This analysis does not impact accident analyses or plant accident scenarios.
The analysis does not impact the accidents as analyzed in the FSAR.
All accident acceptance criteria continue to be met.
The analysis demonstrates that the proposed amendment meets the acceptance criteria for criticality for spent fuel storage racks and new fuel storage racks. Operation of the plant in accordance with the proposed license amendment will hot impact accident analyses or plant accident scenarios as analyzed in the FSAR.
2)
The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated because no changes are being made to fuel which affect fuel handling methoos. No change to the plant other than that described for fuel is being made.
Thus, no new failure modes are l
being introduced.
Operation of the plant in accordance with the proposed license amendment will not create any initiators for accidents, including any accidents that may be different than already evaluated in the FSAR.
3)
The proposed license amendment does not involve a significant reduction in a margin of safety because increasing the fuel enrichment does not change the conclusions of the accident analyses or safety limits of the plant.
This analysis does not decrease the margin of safety as described in the bases to any Technical Specification.
The analysis does not adversely affect the operation of the fuel.
page 6 of 9 e rwwC WQQt'd(fi,
B.
jncreased Allowable Enrichment of New Fuel in the New Fuel Storaae Racks
- 1) Operation of Joseph M. Farley Units 1 & 2 in accordance with the proposed license amendment does not involve a significant increase
-in the probability or consequences of an accident previously evaluated with respect to new fuel in the new fuel storage racks because the applicable safety limits do not change and are within the bounds previously established. Neither actuation of safety systems nor accident mitigating capabilities are adversely affected by operation of the plant in accordance with the proposed license amendment. The analysis demonstrates that the proposed amendment does not pose a challenge to installed safety systems.
Therefore.
-no new performance requirements are being imposed on any system or component important to safety such that any design criteria will be exceeded.
The implementation of the criticality reanalysis is not an initiator for any of the postulated FSAR accidents analyzed.
This analysis does not impact accident analyses or plant accident scenarios.- The analysis does not ' impact the accidents as analyzed in the FSAR. All accident acceptance criteria continue to be met.
In' addition, the Keff design limits of 0.95 for the full water density condition and 0.98 for the optimum moderation condition are not exceeded.
Therefore, dose calculations are not affected by this reanalysis. The ability to mitigate the consequences of any accidents analyzed in the FSAR is not adversely affected by the implementation of the criticality reanalysis. As such, the conclusions presented 'n the FSAR remain valid such that no increase in radiological consequences will result.
l
- 2) The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated w;th respect to new fuel in the new fuel storage racks
'because no changes are being made to fuel which affect fuel l
handling methods. No change to the plant other than that described l
for fuel is being made. Thus, no new failure modes are being j
introduced. Operation of the plant in accordance with the proposed license amendment will not create any initiators for accidents, including any accidents that may be differ 9nt than already I
evaluated in the FSAR.
- 3) The proposed license amendment does not involve a significant reduction in a margir, of safety with respect to new fuel in the new fuel. storage racks because increasing the fuel enrichment does not change the conclusions of the accident analyses or safety limits of the plant.
The Keff design limits of 0.95 for the full water density condition and 0.98 for the optimum moderation condition continue to be met.
Page 7 of 9 a.
~.
.~..
l C.
Increased Allowable Enrichment of Fuel in the Spent Fuel StoraqLBicit
- 1) Operation of Joseph H. Farley Units 1 & 2 in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated with respect to fuel in the spent fuel storage racks because the applicable safety limits do not change and are within the bounds previously established.
Neither actuation of safety systems nor accident mitigating capabilities are adversely affected by operation of the plant in accordance with the proposed license amendment.
The analysis demonstrates that the proposed amendment does not pose a challenge to installed safety systems.
Therefore, no new performance requirements are being imposed on any system or component important to safety such that any design criteria will be exceeded. The implementation of the criticality reanalysis is not an initiator for any of the postulated FSAR accidents analyzed.
This analysis does not impact accident analyses or plant accident scenarios.
The analysis does not impact the accidents as analyzed in the FSAR. All accident acceptance criteria continue to be met.
In addition, the Keff design limit of 0.95 is not exceeded.
Therefore, dose calculations are not affected by this reanalysis.
The ability to mitigate the consequences of any accidents analyzed in the FSAR is not adversely affected by the implementation of the criticality reanalysis. As such, the conclusions presented in the FSAR remain valid such that no increase in radiological consequences will result.
- 2) The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evt.luated with respect to fuel in the spent fuel storage racks because no changes are being made to fuel which affect fuel handling methods. No change to the plant other than that described for fuel is being made. Thus, no new failure modes are being introduced. Operation of the plant in accordance with the proposed license amendment will not create any initiators for accidents, including any accidents that may be different than already evaluated in the FSAR.
- 3) The proposed license amendment does not involve a significant reduction in a margin of safety with respect to fuel in the spent fuel storage racks because increasing the fuel enrichment does not change the conclusions of the accident analyses or safety limits of the plant.
The Keff design limit of 0.95 continues to be met.
1 Page 8 of 9
t 1
CONCLUSION Based on the preceding analysis it is concluded that operation of Joseph H.
farley Units 1 & 2 in accordance with the proposed amendment does not result in the creation of a significant hazards consideration, a significant increase in the probability of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, nor involve a significant reduction in any margins to plant safety.
Therefore, the 11 tense amendment does not involve a Significant llazards Consideration as defined in 10 CFR 50.92.
Page 9 of 9
..