ML20005E751

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Application for Amend to License NPF-30,revising Tech Specs 5.3 & 5.6.1.1 & Bases 3/4.9.3 Re Spent Fuel Pool Storage & Bases for Decay Time Prior to Fuel Movement
ML20005E751
Person / Time
Site: Callaway Ameren icon.png
Issue date: 12/28/1989
From: Schnell D
UNION ELECTRIC CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20005E752 List:
References
ULNRC-2130, NUDOCS 9001100245
Download: ML20005E751 (32)
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< i 1901 Gratiot Street -

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December 28, 1989 Amidmm#

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NS U.S.cNuclear Regulatory Commission i ATTN: Document Control Desk y

Mail Station P1-137

Washington, D . C .' 20555 ,

Gentlemen: 'ULNRC-2130 ..

DOCKET NUMBER 50-483 i CALLAWAY PLANT UNIT 1

~ REVISION TO TECHNICAL SPECIFICATION SECTIONS 5.3, 5.6'.1.1, AND BASES 3/4.9.3 .

CONCERNING SPENT FUEL POOL STORAGE AND BASES FOR DECAY TIME PRIOR TO FUEL MOVEMENT

?  :

h Union Electric Company herewith transmits an application for Amendment to

' Facility Operating License No. NPF-30 for the Callaway. Plant.

This amendment request is in support-of Callaway Cycle 5. In Cycle 5 fuel enrichments will increase to 4.40 w/o U235 and projected core region average burnup will ,

increase to 52~000 MWD /MTU.

, Specifice.tions 5.3 and 5.6.1.1 are revised to allow storage in Region 1_of the spent fuel pool of Vantage 5

(VS) fuel with integral fuel burnable absorbers (IFBAs) and maximum initial enrichments of 4.45

[f w/o U235. Specification 5.6.1.1 is further revised to reference a requirement that the reference kcw in unborated water be less than or equal to 1.455 at 68 F. This requirement lyields a Keff.of less than or equal to 0.95 in l

Region 1 of the spent fuel pool. Specification 5.6.1.1 is also-revised to reference the Callaway FSAR, Chapter 9 lA, for a description of the uncertainties considered in the criticality analysis of the Region 1 spent fuel storage racks. Finally, Bases Section 3/4.9.3 is revised to include an additional description of the basis for the decay time required after shutdown and prior to fuel movement. Analyses verified that the decay time did not change as a result of the changes requested in this l amendment.

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9001100245 891228 bOM PDR ADOCK 05000483 P PDC i g a . +

"This amendment request is complete for storage of V5 fuel with IFBAs and maximum initial' enrichments of 4.45 w/o U235 in Region

  • l'of the spent-fuel pool . _ Storage in Region 2

-of the spent fuel pool will not be required  ;

until Callaway Refueling 6 (approximately the Fall _of 1993)., Union Electric will submit an amendment request concerning storage in Region 2_of the. spent fuel pool. . Union Electric will not store VS-fuel with maximum initial- ,

zenrichments greater than 4.25 w/o'U235 per  ;

Technical Specification Figure 3.9-1,_in Region 2 of'the spent fuel pool, until the,NRC has reviewed and approved the amendment request.

Supplemental criticality analyses have been performed to verify the storage of V5' fuel with maximum initial enrichments of 5.00 >

w/o U235 in the new fuel storage racks. The results confirm that these fuel assemblies may be safely stored in the new fuel storage racks without exceeding criticality safety limits.

The results of these analyses have been reviewed under the provisions of 10 CFR 50.59 and.are not-included in this amendment request since the Technical Specifications do not address storage in the new fuel racks.

The Cal'laway Plant On-Site Review Committee and the Nuclear Safety Review Board have reviewed and approved this amendment request. Attachments.1, 2, and 3 provide the Safety Evaluation,-the Significant Hazard Evaluation, and the proposed Technical Specification changes. Attachment 1 Appendix C, presents the changes to radiological consequences from those previously reported.

Union Electric expects to begin receipt of fuel on June 1, 1990 for Callaway Cycle 5. NRC's review and approval of this amendment request in support of this date is requested.

Very truly yours,

/7

.&w Donald F. Schnell DJW/kea

s Attachments:

1. Safety Evaluation '

Appendix-A - Criticality Data Appendix.B - Environmental Evaluation' Appendix C - Radiological Consequences for Accident. Analyses

2. Significant Hazard Evaluation
3. Proposed Technical Specification changes !

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o STATE OF MISSOURI )-

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Alan.C. Passwater, of lawful age, being-first duly sworn upon oath says,that he is Manager, Licensing and Fuels (Nuclear) for Union Electric Company; that he has read the foregoing document and

  • knows'the content thereof; that he has executed the same for and on behalf:of:said company with' full' power and authority to do so;-and that-the' facts therein stated are true and correct to the best of his knowledge, information and' belief.

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i By *I MN Alan C. Passwater .-i Manager, Licensing and Fuels-Nuclear SUBSCRIBED and sworn to before me this cd W day of' 2T W , 1989 -

BARBARA J. PFAF NOTARY PUBLIC, STATE OF MISSOURI MY COMMISSION EXPlRES april 22, 1993 ST. LOUIS COUNTY .,.g

l cc: Gerald Charnoff, Esq.

Shaw, Pittman, Potts & Trowbridge 2300=N. Street, N.W.

Washington, D.C. 20037 Dr. J.-O. Cermak CFA, Inc.

4 professional Drive (Suite 110).

Gaithersburg, MD 20879 R. C. Knop Chief,; Reactor Project Branch 1 U.S. Nuclear. Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Office U.S.-Nuclear. Regulatory Commission RR# 1. <

Steedman, Missouri 65077 Tom Alexion (2)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission <

1 White Flint, North, Mail Stop 13E21 l 11555 Rockville Pike Rockville, MD 20852 Manager, Electric Department Missouri Public Service Commission -!

P.O. Box 360 Jefferson City, MO 65102 Ron Kucera Department of Natural Resources l P.O. Box 176 Jefferson City, MO 65102 i

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i .bcc: D..Shafer/A160.761 f:? /QA Record-(CA-758)

Nuclear-Date-E210.01 DFS/ Chrono

-D..F..-Schnell ,

J. E.-Birk ,

J. V. Laux M.-A. Stiller G. .L. Randolph R. J. Irwin H. Wuertenbaecher W. R. Campbell A. C. Passwater R. P. Wendling D. E. Shafer D. J. Walker O. Maynard (WCNOC)

N. P. Goel (Bechtel)

T. P. .Sharkey NSRB (Sandra Auston)

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'Atttchm:nt 1:

ULNRC-2130 Page 1 of 6 SAFETY EVALUATION

This amendment request is in support of.Callaway Cycle 5. In Cycle ~5 fuel enrichments will increase to'4.40 '

w/o U235 and projected core region average burnup will increase'to,52,000 MWD /MTU. This amendment requests: that '

Technical Specification sections 5.3.1 and 5.6.1.1 be revised-to reflect a maximum enrichment of 4.45 w/o U-235

-for fuel storage in Region I; that Technical-Specification l; section 5.6.1.1.a be revised to remove the referenced reactivity uncertainty and refer to FSAR section 9.1A; that '

Technical Specification 5.6.1.1.c be added to include an additional requirement that the reference k<>6 for fuel assemblies be less than or equal to 1.455 in unborated water

-at 68'F; and that Bases Section 3/4.9.3 include an additional description of the basis for the required decay time after shutdown and prior to fuel movement.

Callaway's second reload core (Cycle 3) introduced the Westinghouse Vantage 5 Fuel (VS) option as a mix with the Westinghouse Standard Fuel Assemblies (SFA) and Optimized Fuel Assemblies (OFA). Beginning with Cycle 5, Callaway plans to utilize only the V5 fuel design. In order to '

achieve Union Electric's economic goals, fuel strategies were developed for Cycle 5 which required utilizing fuel enrichments of 4.40 w/o U-235, which exceeds the current enrichment limit of 4.25 w/o.U-235 for stored fuel.in Region I. Supplemental criticality analyses were performed.to support storage of 4.45 w/o U-235 fuel (allowance for manufacturing tolerances on enrichment), and additional t assessments were made to determine the impact of using the higher enrichment fuel on spent fuel' pool design criteria.

The referenced generic Westinghouse reactivity l uncertainty in Technical Specification section 5.6.1.1.a is deleted and a reference to FSAR, Section 9.1A is provided for a more complete description of the uncertainties and tolerances involved in the criticality analyses. Moving the uncertainty from the Technical Specifications to the FSAR is administrative in nature and does not increase the probabilities or consequences of any accidents.

Bases Section 3/4.9.3 is revised to include a statement that further justifies the required decay time after shutdown and prior to fuel movement. This statement adds the fact that the decay time assures the validity of the thermal hydraulic analyses and the assumptions used in the fuel handling accident radiological consequences. This change expands the bases section only. Analyses show that the required decay time in Specifications 3/4.9.3 is unchanged as a result of the increased fuel enrichments and extended burnups.

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Attechmsnt 1 ULNRC-2130 Page 2 of 6 1" ,

' The analyses and evaluations performed to support -the storing of higher enrichment fuel in Region 1 concluded thatt (1)L spent fuel criticality limits are maintained:when

-storing fuel to a maximum initial enrichment of 5.00 w/o U-235, provided fuel with enrichments greater than 3.85 w/o U-235 have sufficient Integral-Fuel Burnable Absorbers

-(IFBA): to maintain' a . reference . fuel assembly koo less than j or equal-to 1.455 at 68*F; (2) the thermal-hydraulic, and =!

design bases of fuel rack / fuel _ pool / cooling systems are met; andl(3) environmental and radiological aspects are within .

the bounds of the licensing ~ basis for the plant.

A-reanalysis of the thermal-hydraulic behavior of the spent-fuel-pool and radiological and environmental considerations-(including the postulated dropped bundle accident)~was performed to support the increase in fuel '

enrichment and subsequent increase in analyzed burnup in future cycles, and therefore fission product inventory increase,, ensuring that the higher enriched fuel can be safely discharged. The criticality analysis to support j storage of the higher enriched fuel in Region 2 will be j submitted in.a separate _ amendment.  !

1 Description of the Callaway Spent Fuel __ Pool j The Callaway spent fuel pool utilizes the maximum j .

density rack (MDR) design concept. Under this concept, the 1 spent fuel pool is divided into two_ separate and distinct 'l regions which, for-the purpose of criticality

  • onsiderations, may be considered as separate pools.

Suitability of this design assumption regarding pool separability is assured through appropriate design restrictions at the boundaries between Region 1 and Region 2. Region 1 of the pool is designed on-the basis of j conservative unirradiated fuel assemblies and a full core i off-loading if that should prove necessary. Region 2 is j designed to. safely store irradiated fuel assemblies in large j numbers. The only c?ange in criteria between Region 1 and  !'

Region 2 is the recognition of actual fuel and fission product inventory accompanied by a system for verifying fuel burnup prior to moving any fuel assembly from Region 1 to Region 2. In both Regions 1 and 2, subcriticality (Keff < 0.95) is maintained during all normal, abnorn al, or accident conditions.

The spent fuel pool ie a reinforced concrete structure with a stainless steel liner. Fuel storage rack modules are constructed with square boxes which form a honeycomb structure. The rack modules are freestanding on the floor liner plate of the pool. The pool is filled with borated 3 water with a boron concentration of at least 2000 ppm. The fuel pool cooling and cleanup system functions to limit the pool temperature with one train operating during normal plant conditions; removes impurities for visual clarity; and limits the radiation dose to operating personnel during normal and refueling operations.

AttCchm:nt 1 I ULNRC-2130 Page 3 of 6 Description of the Callaway Plant Fuel Designs The physical characteristics of OFA, SFA, and V5 fuel  ;

assemblies are similar. MDue designs employ 17 x 17 fuel rod  !

arrays and the fuel rods are zircaloy clad. The OFA and V5 l

designs, however, utilize a smaller fuel rod diameter with j' chamfered pellets and employ zircaloy rather than inconel ,

mixing vane spacer grids. The V5 fuel utilizes intermediate L

flow mixer grids which are nonstructural zircaloy grids installed between the three uppermost zircaloy grids. Thus

[ the V5 fuel is conservatively represented by the OFA fuel i

design which does not contain the intermediate flow mixing grida (alse neutron absorbing members). The V5 design also t incorporates IFBA's which consist of a thin boron coating on the outside of the fuel pellet. Ae a result, the IFBA is a i non-removable and thus integral part of the fuel assembly once it has been manufactured. The IFBA absorber material is in the form of a zirconium dibcride (ZrB 9 ) coating on the fuel pellet. Each IFBA rod has a nominal p51 son material ,

loading of 0.0015 grams BlO per inch. For the analysis a coated length of 108 inches was reduced by 5% to yield a mass of 0.1539 grams B10 per IFBA rod. With respect to all other components in the active fuel region, the OFA and V5 fuel types contain approximately the same fuel weight (UO2 )*

The V5 weight will be slightly different due to incorporation of sxial blankets; however, the analysis assumed full length fuel for conservatism. Enrichments used for cycle 5 exceed the 4.25 w/o U-235 used in the previous criticality analyses. For this reason supplemental criticality analyses were performed to confirm the extension  !

to 4.45 w/o U-235.

priticality ADalysis Extensive analyses have been previously performed to support the storage of both SFA and OFA fuel assemblies under both nor al and postulated accident conditions and to store the fuel tp to a maximum initial enrichment of 4.25 w/o U-235. To increase the maximum allowable enrichment for the Callaway Region 1 storage, the concept of reactivity equivalencing is used. The concept of reactivity equivalencing is predicated upon the reactivity decrease associated with the addition of IFBA fuel rods. A series of reactivity calculations are performed to generate a set of IFBA rod number versus enrichment ordered pairs which all yield the equivalent Keff when the fuel is stored in the spent fuel racks.

The data points for the reactivity equivalence curve are calculated with a transport theory computer code, PHOENIX. PHOENIX is a depletable, two-dimensional, multigroup, discrete oruinates, transport theory code. A 25 energy group nuclear data library based on a modified version of the British WIMS library is used with PHOENIX.

The PHOENIX code has been validated by comparisons with experiments where isotopic fuel composition has been

Atttchm:nt 1 g

ULNRC-2130 Page 4 of 6 examined following discharge from a reactor. In addition, r

6 an extensive set of benchmark critical experiments has been analyzed with PHOENIX. Results of these experiments are given in Table 1 of Appendix A. Figure 1 and Table 2 of ,

[ Appendix A show the constant Keff contour for the Callaway l

" Region 1 spent fuel racks. Note in the figure the endpoint L at 0 IFBA rods where the enrichment is 3.85 w/o U-235 and at L

80 IFBA rods where the enrichment is 5.0 w/o U-235. The '

interpretation of the data is as follows: the reactivity of

[

the Region 1 spent fuel racks containing fuel with 80 IFBA rods which has an initial enrichment of 5.0 w/o U-235 is i equivalent to the reactivity of the spent fuel racks i containing fresh fuel having an initial enrichment of_3.85 L w/o U-235 and containing 0 IFBA rods. Figure 1 in Appendix L A will be incorporated and maintained in the Callaway FSAR.

t The equivalent Keff for the storage of spent fuel in the fuel racks is determined by modeling two out of four storage locations using only the OFA/V5 fuel assembly design. Studies have shown that an OFA/VS assembly yields a ';

larger Keff than does a SFA assembly of the same enrichment.

The KENO-IV computer code was used to calculate the maximum fuel enrichment with no IFBA's to encure the storage rack >

multiplication factor was less than or equal to 0.95. This fuel enrichment was determined to be 3.85 w/o. The KENO calculation for the nominal case resulted in a Keff of  ;

0.9205 with a 95 percent probability /95 percent confidence level uncertainty of 1 0.0066. The maximum Keff under normal conditions was determined with a " worst case" KENO model which included mechanical and material tolerances in ,

addition to asymmetric positioning of fuel assemblies within ,

the storage cells. The maximum Keff for the Callaway Region 1 spent fuel storage racks was 0.9476 including method biases and uncertainties at a 95/95 probability / confidence level.  ;

To simplify verification of acceptability for storage of fuel in the spent fuel racks, an infinite multiplication factor for a fresh 3.85 w/o U-235 fuel assembly was determined. The infinite multiplication factor, or Keo, is used as a reference criticality reactivity point which eliminates the need to specify an acceptable enrichment versus number of IFBA rods correlation. The fuel assembly Koo depletion calculations were performed using the Westinghouse licensed core design codes. The codes include TURTLE and PHOENIX-P. Calculation of the infinite multiplication f actor resulted in a reference kooof 1.455 at 68 F. This value is consistent for the varying enrichment /IFBA combinations.

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Att2 chm:nt 1 ULNRC-2130 Page 5 of 6 L_

Thermal-Hydraulic and Fuel Buildino Ventilation Analysis The thermal-hydraulic analyses were reanalyzed to assure that depleted V5 assemblies can be atored in the spent fuel pool with adequate cooling and without adverse impact on fuel building HVAC performance. A reanalysis of the spent fuel pool (SFP) cooling system was performed to assure that discharged spent fuel assemblies would be adequately cooled and free from boiling for both normal and off-normal. conditions. This reanalysis was performed to assess the impact on the SFP system due to the increases in fuel enrichment, cycle lengths and capacity factors which will eventually result in burnups which exceed those previously analyzed. The change from SFA to OFA was previously submitted in ULNRC-1192 dated October 15, 1985 and ULNRC-1207, dated November 15, 1985 and the change to V5 L was submitted in ULNRC-1422 dated December 19, 1986 and ULNRC-1470 dated March 31, 1987.

With increased fuel enrichments and extended burnups, the heat load to the fuel pool and to the fuel building environment is increased. The results of the reanalysis show that the calculated peak bulk pool temperatures for normal and off-normal conditions are limited to less than 140 degrees for normal operation, and less than 160 degrees for accident conditions. The limit of 140'F is an increase over the current Callaway FSAR value of 135'F, but remains consistent with the temperature requirements established in the Standard Review Plan. The FSAR will be revised to reflect the value of 140'F. In addition, the calculated peak clad temperatures are below the saturation temperature at the peak clad temperature location. The SFP cooling system will provide adequate cooling of discharged spent fuel assemblies to limit peak bulk pool temperatures to below these limits and to assure that the spent fuel assemblies are free from boiling.

Plant operation at increased fuel enrichraent and extended burnup was also reviewed to ensure the proper operation of the fuel building HVAC. Previous analyses  ;

demonstrated that the fuel building HVAC is adequate to maintain an environment consistent with personnel comfort and safety. The ability of the fuel building HVAC to limit the accidental release of radioisotopes to below applicable limits is not impacted. The analyses verified adequate margin exists in the fuel building HVAC system to maintain its required performance at the increased heat loads.

An increase in fuel enrichment of the V5 fuel does not ,

alter the normal performance of the fuel pool cleanup systems, fuel building ventilation or radiological control systems.

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Attcchm:nt 1 ULNRC-2130 Page 6 of 6

_ Environmental and Radiolocical Evaluations An evaluation of the environmental impact was performed

.and is included as Appendix C to this safety evaluation.

! The increased fuel enrichment and subsequent extended fuel burnups do not significantly provide an adverse environmental effect.

A review of postulated accidents, including fuel  ;

handling accidents, was performed. The radiological l consequences were updated and are included as Appendix C to this safety evaluation. The updated consequences yield an <

increase in dose for several accidents, but the results are clearly-within the limits of 10CFR100.

Conclusions Based on the above discussions and attached Appendices, this amendment request does not adversely affect or endanger the health or safety of the public and does not involve an unreviewed safety question, nor an unreviewed environmental ,

question.

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' APPENDIX.A

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CRITICALITY DATA.

$~.v ' INCREASED ENRICHMENT TO J

2fi 4 ','45'W/o s ,

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WESTINGHOUSE 17lX 17 VANTAGE 5 FUEL s

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b TABLE 1 Benchmark Critical Experiments PHOENIX Comparison

  • I~ Description of Number of PHOENIX Keff Using Experiments Experiments Experiment Bucklings UO 2

Al clad 14 0.9947 SS clad 19 0.9944 Borated H 2 O 7 0.9940 Subtotal 40 0.9944 Al clad 41 1.0012 TOTAL 81 0.9978

  • These critical experiments benchmark the Westinghouse methodology and use of the PHOENIX Computer Code.

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TABLE 2 35 L Callaway Fuel Assembly Minimum IFBA rods vs. Initial U

!- Enrichment for Region 1 Spent Fuel 235

  • Initial U IFBA Rods Enrichment In Assembly 3.85 0 4.0 10 4.2 24 4,4 38 46 52 4.8 66 5.0 80 Each IFBA rod has a nominal poison material loading of 0.1539 grams BIO.

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ENVIRONMENTAL EVALUATION  ;

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. INCREASED ENRICHMENT AND EXTENDED FUEL BURNUP l

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APPENDIX B ULNRC-2130 Page 1 of 3 Environmental Evaluation This amendment request is in support of Callaway Cycle 5. ~In Cycle 5 fuel enrichments will increase to 4.40 w/o U235 and projected core region average burnup will increase to 52,000 MWD /MTU. This amendment requests: that !

Technical Specification sections 5.3.1 and 5.6.1.1 be revised to reflect a maximum enrichment of 4.45 w/o U-235 for fuel storage in Region It that Technical Specification section 5.6.1.1.a be revised to remove the referenced i

reactivity uncertainty and change the referenced FSAR'

- section 9.1A; that Technical Specification 5.6.1.1.c be -

added to include an additional requirement that the reference.k es for fuel assemblies in Region 1 be less than or equal to 1.455 at 60*F; and that the Bases Section 3/4.9.3 include an additional description of the basis for the required decay time after shutdown and prior to fuel movement.

A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration:

(2)' result in a significant change in the types or significant increase in the amount of any effluents that may be released offsite: and (3) result in an increase in individual or cumulative occupational radiation exposure.

Union Electric Company has reviewed this request and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR I

51.22(C)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be ,

prepared in connection with the issuance of the amendment.

The.following provides the basis for this determination.

Proposed Chance Included in the proposed Technical Specification changes is a revision to the limit of maximum fuel enrichment.

Specifically, the change will:

' change Technical Specification 5.3.1, which currently requires that reload fuel have a maximum enrichment of 4.25 weight percent U-235. Extended fuel burnups project a core region average burnup increase to 52,000 MWD /MTU.

add to Technical Specification 5.6.1, concerning design requirements of the Spent Fuel Storage Racks, an additional requirement to require that a maximum core geometry Koo for PWR fuel assemblies in Region 1 be less than or equal to 1.455 at 68'F.

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b' APPENDIX D ULNRC-2130 Page 2 of 3 Basis-L 'The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

1. As demonstrated in Attachment 2 to this letter,  ;

the proposed amendment does not involve a significant hazards consideration.

t I 2. The proposed amendment does not result in a '

significant change in the types or significant increase in the amounts of any effluents that may i:

be released offsite. The proposed amendment L

allows an increase in fuel enrichment from 4.25 weight percent U-235 to 4.45 weight percent which L will allow an increase in the batch-average burnup level to 52,000 MWD /MTU (60,000 MWD /MTU peak rod burnup). Based on extensive studies  !

conducted for the NRC by Pacific Northwest i Laboratories (NUREG/CR-5009, " Assessment of the Use of Extended'Burnup Fuels - In Light Water Power Reactors"), the NRC has concluded that there are no significant adv9rse radiological or nonradiological impacts associated with the use of-extended burnups up to 60,000 MWD /MTU and fuel enrichments up to 5.0 weight percent U-235. This i conclusion was documented by the NRC in a public ,

notice, " Extended Burnup Fuel Use in Commercial '

LWRs; Environmental Assessment and Findings of No Significant Impact," dated February 23, 1988. l Based on the above the proposed amendment does not result in a significant change in the types or -

significant increase in the amounts of any effluents that may be released offsite.

3. The proposed amendment does not result in a significant increase in individual or cumulative occupational radiation exposure. This conclusion is supported by the NRC as documented in a public notice, " Extended Burnup Fuel Use in Commercial LWRs; Environmental Assessment and Finding of No Significant Impact," dated February 23, 1988.

Conclusions In conclusion and based on NUREG/CR-5009, there are no significant adverse effects from the increased fuel enrichment. A loss of fuel integrity is not likely as power levels for the fuel rods remain normal. The activity inventory may increase for long lived radionuclides of j l

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APPENDIX B ULNRC-2130 Page 3 of 3

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concern, however, for short-lived fission products, the inventories will essentially remain the same. If leakage >

' from a fuel element occurs during operation, the activity will be removed by the plant cooling-water cleanup system.

Cooling-water activity is limited by Technical Specifications. Extending fuel burrup results in environmental consequences which are either less than or virtually the same as those evaluated in the Final =

Environmental Statement (NUREG-0813) by the NRC. This evaluation remains applicable to the nuclear-fuel cycle required to support extended fuel burnups to 60,000 MWD /MTU..

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7 RADIOLOGICAL" CONSEQUENCES j o

'FORL.' ACCIDENT ANALYSES

.it CALLAWAY' PLANT 7

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WESTINGHOUSE <1 -7 X 17-VANTAGE 5 FUEL .

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ULNRC-2130

! Page 1 of 8 I.

SUMMARY

AND CONCLUSIONS Union Electric has reviewed the accidents analyzed in the FSAR with respect to radiological source terms and radiological consequences.

Radiological consequences are evaluated with source terms based on the 3636 MWt core rating (3565 MWt plus 2% postulated calorimetric error) and Callaway specific meteorology based on three years of combined meteorological data.

The updated consequences yield an increase in dose for several accidents, o but the results are clearly within the limits of 10CFR100. For some cases-margin has been reallocated, resulting in a decrease or increase in doses as ,

shown in the attached Table.

II. RADIOLOGICAL CONSEQUENCES FOR ANALYZED ACCIDENTS

\

FISSION PRODUCT INVENTORIES The calculation of the core fission product inventory employs the ORIGEN II computer code modelling a three region enveloping cycle core with a core power level of 3636 MWt. Of the 96 assemblies in core Region 1, 32 have-

+ operated at a specific power of 50.7 MW/HTU for 474 days and 64 have operated at a specific power of 57,0 MW/MTU for 474 days. Of the 88 assemblies in core Region 2, 24 have operated at a specific power of 48.5 MW/MTU for 474 days and at 44.3 MW/MTU for.474 days and 64 have operated at a specific power of 52.7 MW/MTU for 474 days and at 33.8 MW/MTU for 474 days. The 9 assemblies in core Region 3 have operated at a specific power of 50.6 MW/MTU for 474 days, at 38.0 MW/MTU for 474 days and at 21.1 MW/tfrU for 474 days. The average burnup in the regions at the end of a cycle (MWD /NTU) is 26,000, 42,000 and 52,000 respectively. The isotopic yicids utilize data for fissioning of U-235, U-238, and Pu-239 and accounts for the depletion of U-235.

STEAM SYSTEM PIPING FAILURE Radiological Consequences Since no core damage is postulated to occur during a Main Steam Line Break Accident, the doses remain unchanged.

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q APPENDIX,C g- .

ULNRC-2130 Page.2fof 8  ;

1 Doses (rem)

REGULATORY.

e FSAR REANALYSIS LIMIT SOURCE I Case 1, 1.0p Ci/gm I-131' equivalent W/I spike p, ' Exclusion area boundary l~ (0-2 hr.) '

U < Thyroid. 2.5 No Change' 30 Rem NUREG-0800 15.1.5 -

> Whole Body 6.23E-3 No Change 2.5 Rem NUREG-0800' 15.1.5 s;,

. Low-population zone.

( ,

(duration)

. Thyroid 3.45 No Change 30 Rem NUREG-0800 l

{ 15.1.5 Whole Body 5.46E-3 No Change 2.5 Rem NUREG-0800' 15.1.5 y

( '

Case 2, 60p Ci/gm I-131 ID equivalent (Callaway i Technical Specification

. Limit) .

. Exclusion' area boundary (0-2 hr.) .;

Thyroid 3.47 No Change 30 Rem NUREG-0800 6 15.1.5 ,

Whole Body 4.30E-3 No Change 2.5 Rom NUREG-0800. l 15.1.5 t F

Low-population zone ,r (duration)-

Thyroid 1.42 No Change 30 Rom NUREG-0800 15.1.5

LOSS OF NON-EMERGENCY AC POWER TO THE PLANT AUXILIARIES i i

Radiological Consequences r i

Since no core damage is postulated to occur during a loss of Non-Emergency AC Power event, the doses remain unchanged.

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3

) ;.

! APPENDIX C  ;

ir ULNRC-2130 Page 3 of 8

[, s l Doses (rem) ,

REGULATORY l c FSAR' REANALYSIS LIMIT SOURCE r

'^

Exclusion. area' boundary-(0-2. hr. )'

6.75E-2 No Change Thyroid 300 Rem 10CFR100 Section 11 Whole Body 1.94E-4 No Change 25 Rem 10CFR100 Section-11

[.

J Low-population zone

-(duration)

Thyroid 3.45E-2 No Change 300 Rem 10CFR100 Section 11 Whole Body 7.11E No Change 25 Rem 10CFR100 Section 11 fI)

REACTOR COOLANT PUMP SHAIT SE1ZURE (LOCKED ROTORl

^ Radiolonical Consequences v

. . Utilizing.the new source terms, the radiological consequences were

!- - recalculated as follows:

Doses (rem) t.

j :. REGULATORY 4

FSAR REANALYSIS LIMIT SOURCE L

L Exclusion area boundary

["

(0-2 hr.)

Thyroid 23.9 23.7 30 Rom NUREG-0800 Whole Body 4.02E-1 3.89E-1 2.5 Rom NUREbb80b 15.3.3.II.C L-Low-population zone (duration)

Thyroid 9.39 9.53 30 Rom NUREG-0800 15.3.3.II.C Whole Body 9.12E-2 8.61E-2 2.5 Rem NUREG-0800 15.3.3.II.C SPECTRUM OF ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENTS (I)

Radiological Consequences Utilizing the new source terms, the radiological consequences were recalculated as follows:

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APPENDIX C UrIJ4RC-2130

'c Page 4 of 8 Doses (rem)-  ;

REGULATORY  :

FSAR. REANALYSIS LIMIT SOURCE  :

Exclusion area boundary

-(0-2 hr.)  ;

,.. Thyroid 17.6 20.6 '75 Rem- NUREG-0800 l 15.4.8, App. A  ;

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1.37E-1 1.62E-1 6 Rem NUREG-0800 Whole Body- '

  1. 15.4.8, App. A k' l

' Low-population zone-

? (duration) ,

Thyroid 16.0 18.8 75 Rem NUREG-0800 ,

15.4.8, App. A ~

Whole Body 3.52E-2 4.16E-2 6 Rom NUREG-0800  ;

4 15.4.8, App. A

-BREAK IN INSTRUMENT LINE OR OTHERLLINES FROM REACTOR COOLANT .[

~ PRESSURE BOUNDARY THAT PENETRATE CONTAINMENT Radiolonical Consequences Since no core damage is postulated to' occur during this event,.tlas doses. .

remain unchanged.

Doses (rem)

REGULATORY FSAR REANALYSIS 1.IMIT SOURCE Exclusion area boundary  ;

(0-2 hr.) '

Thyroid- 1.15E-1 No Change 30 Rem NUREG-0800 15.6.2.11 Whole Body 3.84E-3 No Change 2.5 Rem NUREG-0800 15.6.2.11 Low population zone  :

(duration)

Thyroid 1.50E-2 No Change 30 Rem NUREG-0800 15.6.2.11 Whole Body 5.02E-4 No Change 2.5 Rem NUREG-0800 15.6.2.11

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APPENDIX C l ULNRC-2130 i Page 5.of 8 i STEAM GENERATOR TUBE FAILURE Radiolonical Consecuences )

Since no core damage is postulated to occur during this event, the doses remain as referenced in the letters SLNRC 86-01 (1/8/86) and SLNRC 86-03 (2/11/86). The regulatory limit for these radiological consequences is given in NUREG-0800, Section 15.6.3.II as 2.5 Rem for whole body dose and 30 Rem for
the thyroid dose.. ,

4

[" LOSS-OF-COOLANT ACCIDENTS RESULTING FROM A SPECTRUM OF POSTUIATED '

L PIPING BREAKS WITHIN THE REACTOR C001 ANT PRESSURE BOUNDARY h l Radiolonical Cons _equences  ;

Utilizing the new source terms, the radiological consequences were recalculated as follows: .

Doses (rem)

REGULATORY  ;

L LIMIT FSAR REANAI.YSIS SOURCE Exclusion area boundary (0-2 hr.) ,

a. Containment leakage r (0-2 hr.)

Thyroid 83.2 97.8 300 Rem NUREG-0800' 15.6.5.11 Whole Body 2.84 4.65 25 Rom NUREG-0800 15.6.5.11

b. ECCS recire. leakage *

(0.47-2 hr.)

Thyroid 26.1 30.6 300 Rem NUREG-0800 15.6.5.II Whole Body 7.8E-2 1E-1 25 Rem NUREG-0800 -

15.6.5.II f-Low population zone (0-30 days)

a. Containment leakage ,

(0-30 days)

Thyroid 54.9 64.3 300 Rem NUREG-0800 15.6.5.11 Whole Body 9.8E-1 1.21 25 Rem NUREG-0800 15.6.5.11

b. ECCS recire. leakage (0.47 hr. - 30 days)

Thyroid 58.6 68.5 300 Rom NUREG-0800 15,6.5.II Whole Body 5.6E-2 6.6E-2 25 Rem NUREG-0800 15.6.5.II

g APPENDIX C U ULNRC-2130  ;

Page 6 of 8 Doses (rem)

REGULATORY  !

FSAR REANALYSIS LIMIT SOURCE .q

-)

!. Control Room (0-30 days).

a. Containment leakage Thyroid 15.8 21.7 30 Rem NUREG-0800-6.4.11 i

Whole Body 3.92E-1 4.53E-1 5 Rem NUREG-0800 6.4.11 f Skin 7.24 7.49 30 Rem NUREG-0800

- 6.4.11

b. .ECCS recire. leakage IE Thyroid 2.68 3.85 50 Rem NUREG-0800 6.4.11 Whole Body 7.99E-5 1.17E-4 5 Rem .NUREG 0800 6.4.11 Skin 6.91E-4 9.98E-4 30 Rem NUREG-0800 6.4.11 RADIOACTIVE WASTE GAS DECAY TANK FAILURE Radioloalcal Conseauences y Since no core damage'is' postulated to occur during this event, the doses F remain unchanged.

Doses (rem)

REGULATURY ESAR REANALYSIS LIMIT SOURCE Exclusion area boundary (0-2 hr.).

Thyroid 8.85E-2 No Change 300 Rem 10CFR100 Section 11 Whole Body 3.29E-2 No Change 25 Rem 10CFR100 Section 11

' Low-population zone (duration)

Thyroid 1.16E-2 No Change 300 Rem 10CFR100 Section 11 Whole Body 4.28E-3 No Change 25 Rem 10CFR100 Section 11 RADIOACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE Radiolog! cal Conscauences Since no core damage is postulated to occur during this event, the doses remain unchanged.

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APPEND 1X C

, -ULNRC-2130 l Page 7 of 8 Doses (rem) p ,

REGULATORY j.' FSAR REANALYSIS LIMIT SOURCE Boron Recycle Tank Exclusion area boundary.

4' .

(0-2 hr.)

Thyroid 4.25E-2 No Change. 300 s v 10CFR100 Section 11 Whole Body. 5.10E-3 No Change 25 Rem 10CFR100 Section 11 (i;,

i :- Low-population zone (duraticu)

Thyroid 5.56E-3 No Change- 300 Rem 10CFR100 Section 11 Whole Body 6.65E-4 No Change 25 Rem 10CFR100 Section 11 Primary Evaporator Bottoms Tank Exclusion area boundary-(0-2 hr.)

Thyroid -2.63E-1 No Change 300 Rom 10CFR100 Section 11'

~

Whole Body- 6.11E-5 No Change 25 Rem 10CFR100 Section 11 Low-population zone :

(duration)

Thyroid 3.47E-2 No Change 300 Rom 10CFR100 Section 11 Whole Body 8.09E-6 No Change 25 Rom 10CFR100 Section 11 FUEL HANDLING ACCIDENTS __

Radiological Consequences The maximum assembly activity was recalculated for the new fuel design at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown at a core power level of 3636 Mwt. These activity levels were adjusted by a radial peaking factor of 1.65 in keeping with the

-guidance of Regulatory Guide 1.25. Utilizing these values and the new source terms,1the radiological consequences were recalculated as follows:

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APPENDIX C t

p V- ..

" ULNRC-2130 i

< Page 8 of 8 .

Doses (rem)  !

REGULATORY  ;

FSAR REANALYSIS LIMIT SOURCE In Fuel Buildina,

[

l 1 Exclusion area boundary' ,

, (0-2 hr.) i lJ Thyrold' 6.61 7.'1 300 Rem NUREG-0800 p . .

15.7.4.II l F 'Whole Body; 2.63E-1 2.82E-1 -25 Rem NUREG-0800  :

f 15.7.4.11 ,

a;  ;

k Low population zone i f , (duration) 6.60E-1 7.71E-1 300 Rem NUREG-0800

~

< Thyroid:

'E 15.7.4.11  !'

Whole Body 2.60E-2 2.82E-2 25 Rem NUREG-0800 15.7.4.11

'- .In Reactor Building  ;

Exclusion area bo'2ndary'

.(0-2 hr.): .

" . Thyroid' '27.6 32.2 300 Rem NUREG-0800 15.7.4.11 Whole Body'. 1.34E-1 1.47E-1 25 Rem NUREG-0800'

y, 15.7.4.11 P
Low-population zone .

I (duration);

. Thyroid 2.75 3.22 300 Rem NUREG-0800 15.7.4.11  !

Whole. Body 1.34E-2 1.47E-2 25 Rom NUREG-0800 15.7.4.II  ;

i (1) Note:- The locked rotor and rod ejection accidents have doses that include the effects of steam generator tube uncovery as discussed in ULNRC-1808, dated July 15, 1988.

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L Attcchm:nt 2 i ULNRC-2130  :

Page 1 of 3  !

SIGNIFICANT HAZARDS EVALUATION j i

This amendment request is in support of Callaway Cycle 5. i In Cycle 5 fuel enrichments will increase to 4.40 w/o U235 l and projected core region average burnup will increase to ,

52,000 MWD /MTU. This amendment requests: that Technical 1 Specification sections 5.3.1 and 5.6.1.1 be revised to reflect a maximum enrichment of 4.45 w/o U-235 for fuel 4 storage in Region I; that Technical Specification section , 5.6.1.1.a be revised to remove the referenced reactivity >

c uncertainty and refer to FSAR section 9.1A; that Technical Specification 5.6.1.1.c be added to include an additional E requirement that a maximum reference k for fuel assemblies in stegion i be less than or equal to 1.455 at 68*F; and that

  • Besen section 3/4.9.3 include an additional description of the basis for the required decay time after shutdown and prior to fuel movement.

The Safety Evaluation supporting this amendment request 4

-provides the bases for concluding that the proposed changes '

are consistent with the licensing bases of the spent fuel pool and verify that the proposed changes do not alter safe operation of the spent fuel pool systems nor violate pool  :

criticality safety limits. The reevaluations further demonstrate that an increase in maximum initial enrichment . .

for storage can be up to 4.45 w/o, with sufficient IFBAs, in Region 1 and that a reload size can be extended to nominally include up to 88 assemblies and thrice burned fuel can

-accumulate average burnups up to 52,000 MWD /MTU without a significant reduction in a margin of safety. Since the criticality. safety analysis and the thermal-hydraulic and structural analysis confirm that the original criteria are met, the possibility of a new or different kind of accident or condition over previous evaluations is not credible.

Physically all three fuel types are similar. OFA and V5 fuel are geometrically compatible with SFA. The fuel assembly dimensional envelope, skeletal structure, and internal grid locations are essentially the same. The structural differences for OFA/V5 fuel are a smaller fuel rod outer diameter and zircaloy spacer grids rather than inconel. Neutronic differences between the two fuel designs have been analyzed and determined to not alter spent fuel pool criticality safety limits. The Technical Specification changes incorporate an increase in maximum enrichment limit to 4.45 w/o U-235 which is required for a core reload.

WCAP-10444 describes the V5 fuel design. It has been reviewed and received generic approval by the NRC. V5 fuel was approved for use at Callaway in Amendment 28 dated October 9, 1987.

The referenced uncertainty in Technical Specification section 5.6.1.1.a is a generic Wertinghouse value. This value is deleted and a reference to FSAR Section 9.lA is provided for the description of uncertainties and tolerances

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Atttchm:nt 2  :

ULNRC-2130 i Page 2 of 3 considered'in_the criticality analyses. Moving the l uncertainty from the Technical Specifications to the FSAR is 1 administrative in nature and does not increase the J probabilities or consequences of any accidents.

INCREASE IN MAXIMUM ENRICHMENT TO 4.45 W/O U-235 FOR FOLL STORAGE IN THE SPENT FUEL POOL Extensive analyses were previously performed to support storage of V5 fuel to maximum enrichmenta of 4.25 w/o U-235.

The results of these analyses were submitted in amendment requests via ULNRC-1422 dated December 19, 1986 and i

ULNRC-1470 dated March 31, 1987. Supplemental analyses were

! performed to support the increase to 4.45 w,h U-235 and  :

L increased core region average burnup tc.52,000 MWD /MTU, ^

inc?.uding critictlity analyses, thermal-hydravite analyses p' and rtdiological analyses. Increasing the reaximan l enrichaent limit to 4.45 w/o U-235, including IFBA rods, for-allowed storage in the spent fuel pool does not represent a  ;

significant hazard in that:

1. An increase to a maximum enrichment of 4.45 w/o -

U-235 does not in/olve a significant increase in the probability or consequences of an accident or other adverse condition-over previous evaluations. t Because of the conservative techniques and assumptions used to evaluate the maximum possible neutron multiplication factor, there is more than reasonable assurance that no significant hazard based on criticality safety, is involved in storing fuel assemblies of up to and including '

5.00 w/o U-235, with sufficient IFBAs, in the -

Region 1 spent fuel storage racks under both normal and postulated accident conditions. For example, ignoring the 2000 ppm soluble boron in the spent fuel pool calculations results in conservative values of the multiplication factor.

Storing fresh fuel in the Region 1 configuration at an enrichment of 4.45 w/o U-235, with greater than 42 IFBA rods, would result in a maximum multiplication factor of 0.9476 including all uncertainties.

2. An increase to a maximum enrichment level of 4.45 w/o U-235 does not create the possibility of a new or different kind of accident or condition over previous evaluations. An increase to the enrichment level of 4.45 w/o U-235 involved performing extensive evaluations to envelope the corresponding changes in reactivity. Use of the reactivity equivalencing procedures ensures that the spent fuel pool Region I criticality limits are not exceeded.

= ,

[ Attcchm:nt 2 i ULNRC-2130 Page 3 of 3

3. An increase in the maximum enrichment level to 4.45 w/o does not involve a significant reduction in a margin of safety. As discussed above, in all cases the multiplication factors for worst case j assumptions fall below the regulatory limit and do not represent significant reductions in margin.  ;

, An increase to the enrichment level of 4.45 w/o *

l. does not adversely impact operation of the various plant systems, ie HVAC, spent fuel pool cooling, or radiological control systems.
i. Based on the above discussions and those presented in A Attachment 1, it has been determined that the requested

, Technical Specification revisions do not involve a (t he significant increase in the probability or consequences of an accident .Or other adverse condition over previous 1

evaluationet.cr create the poscibility of a new cr different .

kind of accident or condition over previcos evaluations; or '

involva a eignificant reduction in a margin of safety.

Therefore, the requested license amendment does not involve '

a significant hazards consideration.

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