ML20045E711

Application for Amend to License NPF-3 Revising Provisions in TS to Allow Storage of New & Spent Fuel Assemblies W/Initial Nominal Enrichment of U-235 No Greater than 5.0 Weight Percent

ML20045E711
Person / Time
Site:	Davis Besse
Issue date:	06/23/1993
From:	Meyers T, Storz L CENTERIOR ENERGY
To:
Shared Package
ML20045E709	List: ML20045E708 ML20045E711 ML20045E713
References
2149, NUDOCS 9307020345
Download: ML20045E711 (13)

Text

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. Docket Number 50-346 License Number NPF-3 Serial Number 2149

. Enclosure Page 1 APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 Attached are requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Number NPF-3. Also included is the Safety Assessment and Significant Hazards Consideration, and the Environmental Assessment.

The proposed changes (submitted under cover letter Serial Number 2149) concern:

Appendix A, Technical Specification (TS) 3/4.9.13, Refueling Operations

- Spent Fuel Pool Fuel Assembly Storage, and TS 5.6, Design Features -

Fuel Storage.

For: L. F. Storz Vice President - Nuclear l

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By: / J-T. J. ye s Dire to - Technical Services Sworn and subscribed before me this 23rd day of June, 1993.

hin/t r f) -

Notary Pubkic, State of Ohio Al" EVEYNL DRESS NPUDUC.STATEOFCHO k W 4 4 1994 9307020345 930623 PDR ADOCK 05000346 p PDR

Docket Number 50-346 .l 1

License Number NPF-3 '

Serial Number 2149 Enclosure i Page 2  ;

The following information is provided to support issuance of the l requested changes to the Davis-Besse Nuclear Power: Station Unit Number i 1 Operating License Number NPF-3, Appendix A, Technical Specification ,

(TS) 3/4.9.13, Refueling Operations - Spent Fuel Pool Fuel Assembly -!

Storage, and TS 5.6, Design Features - Fuel ftorage. ,

A. Time Required to Implement: This change is to be implemented within 90 days after the NRC issuance of the License Amendment.  ;

i B. Reason for Change (License Amendment Request Number 90-0042):

i The proposed amendment would revise the provisions in the TS to '

allow storage of new and spent fuel assemblies'vith an initial nominal enrichment of uranium - 235 no greater than 5.0 veight  ;

percent. The proposed amendment would revise TS Figure 3.9-1 to  !

reflect the desired higher enrichment limitation. The proposed amendment would also revise TS 5.6.1.1 to add an appropriate cross-reference to the description of the spent fuel storage racks, ,

and would revise TS 5.6.1.2 to note the 5.0 veight percent {

limitation in the description of the new fuel storage racks. In  !

addition, the proposed amendment would revise the TS 3/4.9.13 .!

Surveillance Requirements to make them consistent with the nev  !

Standard Technical Specifications for Babcock and Vilcox type l plants (NUREG-1430). j In order to support specifying fuel enrichment for Cycle 10 (fuel f batch 12), these changes are requested by October 15, 1993. l I

C. Safety Assessment and Significant Hazards Consideration: See l Attachment 1. ,

D. Environmental Assessment: See Attachment 2. ,

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Docket Number 50-346 License Number WPF-3' >

Serial Number 2149

, Attachment 1 -

Page 1 SAFETY ASSESSMENi' AND SIGNIFICANT HAZARS$ CONSIDERATION i

TITLE-Revision to Technical Specifications (TS) 3.3.13,' Refueling Operations ,

- Spent Fuel Pool Fuel Assembly Storage,_5.6.1.1 and 5.6.1.2, Fuel i Storage Criticality  ;

P DESCRIPTION ,

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The purpose of this license amendment request is to propose. changes to j Technical Specifications 3.9.13, 5.6.1.1, and 5.6.1.2 that vill allow the receipt and storage of nuclear fuel at the Davis-Besse Nuclear .

Power Station having an initial enrichment of up to 5.0 veight percent (vt%) uranium-235 (U-235). In doing this, these Technical .

Specifications have been rewritten so that their format more closely-  !

follows that specified under the New Standard Technical Specification program (NUREG 1430) for Babcock & Vilcox type plants. j i

Increasing the maximum allowable fuel enrichment will permit longer fuel cycles with smaller fuel assembly feed batches, resulting in more -

efficient uranium utilization. Also, fever spent; fuel assemblies will .

be produced, thus reducing on-site spent fuel storage capacity  ;

requirements and reducing long term off-site spent fuel transportation t needs. The current enrichment limit is 3.8 vt% U-235.

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SYSTEMS, COMPONENTS, AND ACTIVITIES AFFECTED Reactor Core j Nuclear Fuel New Fuel Storage Racks (NFSR)

Spent Fuel Storage Racks (SFSR)

SAFETY FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS,_AND ACTIVITIES The nuclear fuel in the reactor core produces heat through the  ;

fissioning of uranium and plutonium. This heat is ultimately used to-produce steam which drives _the turbine to produce electricity. The safety functions performed by the reactor core and nuclear' fuel are to .

retain the fuel in an appropriate geometry for heat removal, Land to prevent the migration of radioactive fission products away from the fuel pellets by encapsulating the pellets in Zircaloy cladding.

The new fuel storage racks (NFSR) are designed to store new, non-irradiated nuclear fuel assemblies in a dry, vertical configuration. The safety functions of the NFSR are to prevent damage _ l to the fuel assemblies from seismic events, and to maintain the fuel assemblies in a non-critical configuration. The NFSR are Seismic Class- .

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Docket Number 50-346

- License Number NPF-3 Serial Number 2149 Attachrent 1 Page 2 l I structures with a nominal 21 inch center-to-center spacing between fuel assemblies. The Davis-Besse Updated Safety Analysis' Report (USAR)

Section 9.1.1 discusses new fuel storage. l The spent fuel storage racks (SFSR) are designed to store either new, non-irradiated fuel or depleted, irradiated fuel in a vertical configuration under water. The safety functions of the SFSR are to prevent damage to the fuel assemblies from seismic events, and to maintain the fuel assemblies in a non-critical configuration. The SFSR are Seismic Class I structures with a nominal 13-3/16 inch center-to-center spacing in one direction and a nominal 12-31/32 inch center-to-center spacing in the opposite direction. Each-fuel assembly is surrounded by a 9 inch square can composed of 1/8 inch stainless steel. The water gap _between the steel cans produces what is known as e

" flux trap", thus reducing rack reactivity. The Davis-Besse USAR Section 9.1.2 discusses spent fuel storage.

EFFECTS ON SAFETY Increasing the fuel enrichment does not affect the safety functions '

performed by the reactor core or the nuclear fuel. Since the actual-mass of the fuel assemblies vill not change, the capabilities of the ,

reactor core or nuclear fuel to maintain the fuel in a coolable geometry will not be impaired. Also, the ability of the cladding to prevent fission product escape vill not be degraded. Any changes to the nuclear properties of the reactor core vill be analyzed as part of the cycle-specific reload safety analysis. Changes to fission product inventories, and resulting potential changes in offsite dose consequences, as a result of the higher burnups achievable with higher enriched fuel, would also be analyzed as part of the cycle-specific reload safety analysis. Since actual fuel assembly-mass will not change, the seismic protection functions of the NFSR or SFSR vill not be impaired.  ;

The primary potential effect on safety catsed by increasing the ,

allowable fuel enrichment vill be on criticality in the NFSR and SFSR.

The criticality safety analysis tor both storage facilities has been '

reperformed assuming 5.0 vt% fuel and is discussed below.

Criticality Methodology The criticality analysis was performed by the B&W Fuel Company, using NRC-approved methods and codes, The primary tools for performing the criticality analysis were the KENO-IV. Monte Carlo code from Oak Ridge National Laboratory (ORNL) and the CASH 0-3 multigroup transport theory

• fuel assembly depletion code from Studsvik of America. Cross sections for KENO-IV came from the standard XSDRN 123-group library. The NITAVL code was used to generate resonance self-shielding corrections for the .

123-group library, and XSDRNPM was used to produce representative flux spectra for spatially weighting the 123-group library cross sections.

All of these codes have been used for previous criticality calculations ,

for the Davis-Besse NFSE and SFSR. The XSDRN, NITAVL, and XSDRNPH codes, as well as the 423-group cross section library, were all i developed by ORNL.

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Docket Number 50-346 License Number NPF-3 l Serial Number 2149' Attachment 1 Page 3 CASMO-3 was used to-calculate biases and penalties for off-nominal conditions (off-center assembly in rack, moderator temperature dependence, etc.). For modeling depleted fuel with KENO-IV, CASMO-3 was-used to generate fuel assembly isotopics as a function of assembly exposure that could be input to KENO-IV. CASH 0-3 is.used videly throughout the industry and has been extensively benchmarked'against critical experiments. By using CASMO-3 to determine moderator l temperature depend (nce effects, the KENO-IV temperature dependence errors discussed in NRC Information Notice 91-66 were avoided.

The NEMO three-dimensional advanced nodal code was used to determine limiting fuel assembly burnup shapes to account for non-uniform axial burnup. This was done by benchmarking NEMO against previous Davis-Besse operating history and comparing the assembly burnups and axial burnup shapes to measured data. In this vay, a conservative, limiting axial burnup shape was determined that was bounding for Davis-Besse fuel assemblies. Reactivity penalties were calculated using this limiting axial burnup shape and were applied to all. cases-involving irradiated fuel.

Assumptions >

For the NFSR, the following assumptions were used:

1. Manufacturing tolerances on the fuel assemblies and the NFSR vere applied in a conservative fashion (i.e., to maximize reactivity).

2. No burnable poisons, control rods, rack structural materials, fuel assembly spacer grids, upper and lover fuel assembly end fittings,

• or other structural neutron absorbers, with the exception of the concrete valls and steel top p? ate of the new fuel storage area, ,

were modeled in the criticality calculations. <

3. Calculations were performed both at flooded conditions (water density = 1.00 g/cm 2) and at optimum moderation or " mist" conditions. Sufficient cases were used to determine the " mist" density that produced optimum moderation (i.e., maximized reactivity).

4. The design of the NFSR-precludes inadvertent placement of fuel assemblies in the NFSR anywhere other than in designated fuel  ;

assembly storage locations. Thus, placement of fuel assembles in non-designated locations did not need to be considered.

For the SFSR, the following assumptions were used: i

1. Manufacturing tolerances on the fuel assemblies and the SFSR vere applied in a conservative fashion (i.e., to raximize reactivity).

2. With the exception of the 9 inch square stainless steel can surrounding each fuel assembly, no SFSR neutron absorbing structural materialc were modeled.

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Docket Number 50-346 License Number NPF-3 Serial Number 2149 l

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Attachment 1 Page 4 ,

3. For normal conditions, no credit was taken for soluble boron in the i vater in the spent fuel pool. By procedure, the spent fuel pool ,

soluble boron concentration is maintained'at a concentration of least 1800 ppm. Credit was taken for the 1800 ppm spent fuel pool soluble boron concentration for off-normal and accident conditions (fuel assembly misloading, dropped fuel assembly, etc.). 4

4. Fuel pellet density, diameter, and stack height were chosen so as to bound all fuel types used to date or projected to be used in the future at Davis-Besse. ,

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5. The SFSR vere considered to be infinite (i.e., reflective boundary conditions) in the horizontal (X-Y) directions and to.have a 12 ~

inch water reflector in the vertical (Z) direction.

6. No burnable poisons, control rods, assembly spacer grids,-or upper or lower assembly end fittings were modeled in the criticality:

calculations.

7. Irradiated fuel was assumed to contain no xenon-135. Further, decay ,

of promethium-149 to samarium-149 was assumed not to have occurred. e Existing Technical Specification 3.9.3 requires at least 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />s-of decay time following irradiation before fuel can be transferred from the reactor core to the SFSR. j i

8. Limiting (i.e., conservative) axial burnup distributions were- -

determined for irradiated fuel assemblies. Reactivity penalties were calculated using this limiting axial burnup shape and were ,

applied to all cases involving irradiated fuel. Also, fuel assembly burnups were conservatively adjusted to account for burnup ,

measurement uncertainty. ,

9. Conservative values were determined for burnable poison effect l (production of additional plutonium in irradiated fuel due to the  !

presence of a burnable poison) and were applied to all cases where irradiated fuel was present. However, as was' stated above, no burnable poison rods were actually present, or vere taken credit i for, in any of the criticality calculations.

10. The temperature of the water in the SFSR vas assumed to be that i vhich produced the most reactivity and vas still within the design basis for the spent fuel pool. Sufficient CASM0-3 cases vere run to determine this temperature. .

11. The design of the SFSR precludes inadvertent placement of fuel ,

assemblies in the SFSR anywhere other than in designated fuel assembly storage locations. Therefore, placement of fuel i assemblies in non-rack cell locations did not need to be  !

considered.

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- Docket Number 50-346 I License Number NPF-3:

Serial Number 2149 Attachment 1  :

Page 5-New Fuel Storage Rack Calculations The following biases and uncertainties were' applied to the NFSR  !

criticality calculations: l Statistical '

Description _ Bias (Ak) Uncertainty (6k)

KENO-IV Bias (flooded) 0.00000 0.00000.

KENO-IV Bias (misted) 0.01016 0.00385 Rack Cell Pitch (flooded) 0.00000 0.00000  ;

Rack Cell Pitch (misted) 0.00889 'O.00405 Vater Temperature 0.00000 -------

Mist Temperature 0.00731 -------

Assembly Homogenization -0.00312 0.00450~

Total Blas/ Uncertainty -0.00312 0.00450*

(flooded)

Total Bias / Uncertainty 0.02324 0.00717*

(misted)

• Square-root-of-the-sum-of-squares The KENO-IV bias was determined by comparing KENO-IV benchmarks against appropriate critical experiments for fuel storage geometries. Vith the large assembly pitch in the NFSR, this bias was determined to be zero ,

for flooded conditions. The rack cell pitch penalty assures _that the ,

NFSR pitch tolerances are conservatively applied (the penalty for this is statistically insignificant for flooded conditions). The temperature penalty assures that the water or mist temperature maximizes reactivity (for flooded conditions, it was determined using CASMO-3 that 90*F, the base temperature for all calculations, yielded maximum reactivity; hence, no penalty is required). The assembly homogenization bias provides for the fact that all final calculations performed using KENO-IV used an homogenized assembly model, which yields slightly_

higher values of reactivity than an explicit assembly model. Penalties shown above that do not have associated-statistical uncertainties were calculated using CASMO-3 (a deterministic method), as opposed to KENO-IV (a statistical method). The statistical uncertainties for the total biases vere calculated using the square-root-of-the-sum-of-squares method. The statistical uncertainties provide a one-sided 95/95 '

confidence level for the number of KENO-IV histories used.

The previous criticality analysis for the NFSR had determined that the full, 80 assembly capacity of the NFSR could not be used while still meeting the " mist" criticality criterion of keff < 0.98, as required byL Technical Specification 5.6.1.2b. The previous analysis required that columns 4 and 7 of the NFSR not contain fuel assemblies (i.e., the NFSR was limited to storing 64 fuel assemblies). Similarly, this. criticality analysis demonstrated that fuel assemblies with enrichments up to 5.0 vt% U-235 could not be stored in the NFSR unless rows "C" and "F" of the NFSR did not contain fuel assemblies. Thus, the storage capacity of the NFSR vould be limited to 60 fuel assemblies (see Figure 1).

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Docket Number 50-346

, License Number NPF-3 Serial Number 2149 Attachment 1 Page 6 Figure 1 Davis-Besse New Fuel Storage Rack Restrictions 1 2 3 4 5 6 7 8 9 10

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Docket Number 50-346 License Number NPF-3 Serial Number 2149 Attachment 1 Page 7 ,

A maximum k,ff of 0.93555, including all biases, uncertainties, and penalties, was calculated for flooded conditions, assuming 5.0 vt% fuel assemblies.in all eighty NFSR locations. This result was acceptable since it was less than the flooded criticality criterion of 0.95, as required by Technical Specification 5.6.1.2a (note that the~ calculation-was performed with the NFSR containing 20 more assemblies than vill actually be permitted). Under optimum moderation or " mist" conditions ,

with rows "C" and "F" empty and all other locations containing 5.0 vt%

fuel assemblies, a maximum k e f 0.96829, including'all biases, uncertainties, and penalties,ffwas determined. This result was .

acceptable since it was less than the optimum moderation criterion of 0.98. Therefore, 60 fuel assemblies with an enrichment of up to 5.0 vt%

U-235 can be safely stored in the NFSR, provided that none are placed in NFSR rows "C" and "F" (see Figure 1). ,

The following shows hov the final value of k eff was determined for the flooded NFSR:

k = 0.93318 (Base KENO-IV k

• N - 0.00312 (Totalbias(fl8Me)d))

+ (0.003148 (Base KENO-IV statistical uncertainty)

+ 0.004502)0.5 (Total bias (flooded) statistical uncertainity) i k = 0.93555 eH The following shows how the final value of keH vas determined for the ,

misted NFSR: ,

k = 0.93727 (Base KENO-IV k )

+ 0.02324 2

(Totalbias(mi$bd))

+ (0.00303 2 (Base KENO-IV statistical uncertainty)

+ 0.00717 )0.5 (Total bias (misted) statistical uncertainty) k = 0.96829 eH i Spent Fuel Storage Rack Calculations The following biases and uncertainties were applied to the SFSR critica ,

calculations:  !

Statistical Description Bias (6k) Uncertainty (6k)

I KENO-IV Bias 0.01453 0.00363 Off-center Placement 0.00679 ------- ,

Rack Cell Pitch 0.00279 -----

l Pool Temperature 0.00642 ------

Total Bias / Uncertainty 0.03053 0.00363*

• Square-root-of-the-sum-of-squares t

The KENO-IV bias was determined by comparing KENO-IV benchmarks against appropriate critical experiments for fuel storage geometries.-The off-center penalty assumes that the fuel assemblies are conservatively off-

Docket Number 50-346 License Number NPF-3 Serial Number'2149 Attachment 1 Page 8 center in their SFSR can. The rack cell pitch penalty assumes that the '

SFSR can pitch tolerances are' conservatively applied. The pool-temperature penalty assumes that the spent fuel pool is at the temperature that maximizes reactivity. For the Davis-Besse SFSR, it vas:

determined using CASMO-3 that reactivity increased monotonically with t pool temperature all the way to the design basis temperature of 206'F.

Penalties shown above that do not have associated statistical uncertainties were calculated using CASMO-3 (a deterministic method),

as opposed to KENO-IV (a statistical method). The statistical uncertainties for the total biases were calculated using the square-root-of-the-sum-of-squares method. The statistical '

uncertainties provide a one-sided 95/95 confidence level for the number of KENO-IV histories used.

Analysis of the limiting axial burnup shapes determined that, when depleted fuel was placed next to fresh fuel, the effects of the non-uniform axial burnup vere statistically insignificant. For cases where burned fuel assemblies were placed directly adjacent to other burned fuel assemblies, the following non-uniform axial burnup effect penalties were defined as a function of burnup:

Burnup Axial Effect Statistical (GVD/MTU) Penalty (6k) Uncertainty (6k) 20 0.01447 0.00317 30 0.03707 0.00229 45 0.06858 0.00206 l Since the only cases insolving burned fuel assemblies directly adjacent

• to other burned fuel assemblies were in the determination of the "B-C" region dividing line (see below), and since this line does not exceed 10.8 GVD/MTU at any point within the range of allowable initial enrichments, only the.20 GVD/MTU penalty defined above was actually used in any calculations, and this penalty was used for all cases.

containing burned fuel used in determining the "B-C" dividing line. ,

In addition to these penalties and biases, a 5 percent penalty was added to the burnup curves for Technical Specification Figure 3.9-1 l (curve burnups were increased by 5 percent) to account _for the measurement uncertainty in assembly-average burnup. As was stated earlier, all of the KENO-IV runs that included depleted fuel also assumed bounding burnable poison effects.

As was the case in the previous criticality analysis for the  :

Dtvis-Besse SFSR, a checkerboard pattern was used in which high enriched, low burnup assemblies were interspersed with lov enriched, high burnup assemblies. Three' categories of fuel assemblies were ,

determined by this analysis: Category'"A", Category "B", and Category "C". Category "A" fuel assemblies are of the lowest reactivity and can be stored in any location in the SFSR without restriction. Category "B" fuel assemblies are of intermediate reactivity and can be stored in any location in the SFSR except directly adjacent (i.e., side by side) to Category "C" fuel assemblies. Category "C" fuel assemblies are of the t

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Docket Number 50-346 License Number NPF-3 Serial Number'2149 Attachment 1 Page 9 highest reactivity and can only be stored directly adjacent to Category "A" fuel assemblies and/or locations not cont ining fuel assemblies.

Except for the renaming of Categories "B" and " fuel assemblies for administrative clarity, these rules are identit. . to those nov in use for the. Davis-Besse SFSR. Also, it should be noted that, at any given time, there should not be more than one batch of fuel that requires designation as Category "C", since, in general, less than one cycle of depletion is required for an assembly to transition from Category "C" _

to Category "B".

Technical Specification Figure 3.9-1 (attached) shows the combinations of fuel assembly burnup and initial enrichment that define these three different categories. Cases were run with KENO-IV that bounded all of '

the possible combinations of loading patterns permitted under the above rules. In all cases, the value of maximum k obtained was less than 0.95(burnupswereadjustedineachcaseto*fdtainatargetmaximum k of about 0.948), including all biases, uncertainties'and penalties (ikkludirgnon-uniformaxialburnup),asdefinedabove,andassumingno  ;.

soluble boron in the spent fuel pool vater. Therefore, using the above  ;

rules, fuel assemblies with an initial enrichments of up to 5.0 vt%

U-235 can be safely stored in the SFSR.

The following is a sample calculation that shows how the final value of 1 k vas determined for the case of depleted 5.0 vt% fuel with a burnup oib2GVD/MTU. This calculation was typical of all calculations performed in generating Technical Specification Figure 3.9-1:

0.88847 k"*

+ 0.03053 (Base KENO-IV"k ")

(Total bias)

> 0.01447 (Axial burnup effect)

+ (0.001622 (Base KENO-IV statistical uncertainty)

+ 0.003632

+ 0.003172)0*5(Axial (Totalburnup bias statistical uncertainty) effect statistical uncertainty) k = 0.93855 eH CONTROLS TO ENSURE PRESERVATION OF T.S. 3.9.13 BASIS A set of rigorous controls vill be used to ensure that the basis for Techn>. cal Specification 3.9.13 vill be preserved during all fuel

movements in the spent fuel pool. These controls include

1. Preparation and independent review of all fuel movement sheets for compliance with Technical Specification 3.9.13 by the Reactor Engineering Unit. l i

i 2. Reactor Engineering oversight of Operations during all fuel l movements.

3. Independent verification of refueling device (bridge, crane, etc.)

location prior to fuel assembly placement or retrieval in the SFSR.

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Docket Number 50-346 License Number NPF-3 Serial Number 2149 Attachment 1 Page 10

4. Visual verification that the SFSR loading pattern for those assemblies moved complies with Technical Specification 3.9.13 within 30 days of any fuel movement in the spent fuel pool.

5. Chemistry verification every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that the spent fuel pool boron concentration is at least 1800 ppm during fuel movements in the spent fuel pool and until the SFSR loading pattern verification mandated in item 4 is performed. (This frequency is consistent with the Technical Specification 3.9.1 requirement for verification >

of boron concentration during refueling activities.)

6. Verification that the SFSR map concurs with the actual SFSR loading pattern, verified in item 4, within 30 days of any fuel movement in the spent fuel pool.

7. Rows "C" and "F" of the NFSR will be physically blocked to prevent ,

inadvertent insertion of fuel assemblies.

Changes to the surveillance requirements for Technical Specification 3.9.13 reflect those specified under the New Standard Technical Specification program (NUREG 1430) for Babcock & Wilcox type plants.

These changes do not reduce the actual administrative controls, listed above, that will be applied to the SFSR. .

SIGNIFICANT HAZARDS CONSIDERATION The Nuclear Regulatory Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changes would: (1) not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) not create the possibility of a new or different kind of accident from any accident previously evaluated;-

or (3) not involve a significant reduction in a margin of safety.

Toledo Edison has reviewed the proposed change and determined that a significant hazards consideration does not exist because operation of Davis-Besse Nuclear Power Station, Unit No. 1, in accordance with these changes, vould:

la. Not involve a significant increase in the probability of an accident previously evaluated. It has been demonstrated, through rigorous criticality calculations performed for both the New Fuel ,

Storage Racks and the Spent Fuel Storage Racks, that an adequate margin of safety for a criticality accident would still exist,-

l and, with the proposed administrative controls, there is no l increase in the probability of such an accident.

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l lb. Not involve a significant increase in the consequences of an accident previously evaluated because increased fuel enrichment, alone, has no effect upon the consequences of any accident previously evaluated for the Davis-Besse Nuclear Power Station l

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Docket Number 50-346 License Number NPF-3 Serial Number 2149 Attachment 1 Page 11 (note that increased fission product inventories due to increased burnups achievable-vith higher enrichments vill be addressed in f cycle-specific reload safety analyses).

2a. Not create the possibility of a new kind of: accident from any accident previously evaluated because the only accident associated with an increase in fuel enrichment remains a criticality accident, and this accident has already been previously evaluated.

2b. Not create the possibility of a different kind of accident _from any accident previously evaluated because the only accident associated with an increase in fuel enrichment remains a criticality accident, and this accident has already been previously evaluated.

3. Not involve a significant reduction in the margin of safety since it has been demonstrated, through a rigorous criticality analysis, that an adequate margin of safety for a criticality accident would still exist.

CONCLUSIONS On the basis of the above,-Toledo Edison has determined that this License Amendment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical Specifications that must be reviewed by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.

l ATTACHMENTS Attached are the proposed marked-up changes to the Operating License.

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