ML20135D426

Criticality Analysis of Beaver Valley Unit 1 Fresh Fuel Racks

ML20135D426
Person / Time
Site:	Beaver Valley
Issue date:	01/31/1997
From:	Johansen B, Penkrot J, Robinson K WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20135D389	List: ML20135D386 ML20135D399 ML20135D426
References
CAC-97-013, CAC-97-13, NUDOCS 9703050234
Download: ML20135D426 (21)

Text

.

CAC-97-013 l

4 Criticality Analysis of the Beaver Valley Unit 1 Fresh Fuel Racks 1

January 1997 Prepared: i Jf bb K. R. Robinson Criticality Services Team Verified: M)

J. @Penkrot Critic lity S aces ea Approved s esto

. J. Johansen, Ma'ngd Core Analysis C ]

O Westinghouse Commerical Nuclear Fuel Division

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9703050234 970227 PDR ADOCK 0500G334 P POR

Table of Contents -

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1.0 Introduction...........................................................................................................1 1.1 Design Description... . .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1 1.2 Design Criteria ... . . ....... .. .. . . .. . . . ... . .. . . . . . . . . . . . . . . .1 )

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2.0 A n a l y ti ca l M e t h o d s ................................................................................................. 2  !

, 2.1 Criticality Calculation Methodology.. .. . . . . .. . .. . .2 3.0 Criticality A nalysis o f t he Fresh Fuel Racks ........................................................ 3 )

3.1 Full Density Moderation Analysis. . ... . ....... ..... . ..... . .. . . .. . .. ...4 l 3.2 Low Density Optimum Moderation Analysis... .. . . ..... ..... .. . . . . . . . .....5  !

4.0 Discussio n o f Postula ted A ccide nts........................................................................ 7 4.1 Fresh Fuel Storage Racks... . ...... .... ... . ... . . . . . . . . . . . . . . . . . . . . . . . . .7  ;

i 5.0 S u m m a ry o f C riticality Res ults ............................................................................. 8 l 1

B i b li o g ra p h y ....... ........ . .................. ................. .................. .... ........... . ....... ..... ........... 13 l 1

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List of Tables Table 1. Fuel Parameters Employed in the Criticality Analysis. . . . . . . . . . . . . . .9 1

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List of Figures i Figure 1. Beaver Valley Unit 1 Fresh Fuel Rack Array Layout... ......... . . . .. . ... .. .10 Figure 2. Beaver Valley Unit 1 Fresh Fuel Rack Cell Layout. ... . . . . .11 Figure 3. Sensitivity of Keff to Water Density in the Beaver Valley Unit 1 Fresh Fuel Racks... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Beaver Valley Unit 1 Fresh Fuel Racks i

1.0 Introduction This report presents the results of a criticaiity re-analysis of the Duquesne Light Beaver Valley ,

Unit I fresh fuel storage racks. The fresh fuel rack design considered herein is an existing array of I fuel racks, previously qualified for storage of 17xl 7 fuel assemblies having enrichments up to 4.5  ;

w/o U235 m, j 4

The Beaver Valley Unit I fresh fuel storage racks are being reanalyzed to allow storage of the Westinghouse 17x17 Standard (STD) and Optimized Fuel Assembly (OFA) fuel assembly types at higher enrichments. The following storage configuration and enrichment limits are considered

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in this analysis:

l Fresh Fuel Racks Storage of fresh fuel assemblies with nominal enrichments up to 5.0 w/o 4 U235 utilizing all available storage cells. No reactivity credit is taken for Integral Fuel Burnable Absorbers (IFBA)in this analysis.  ;

l The Beaver Valley Unit I fresh fuel rack analysis is based on maintaining Ke g 6 0.95 for storage of 17x17 fuel assemblies under full water density conditions and K,g 6 0.98 under low water density (optimum moderation) conditions. The optimum moderation condition applies only to .

fresh fuel racks since these racks are used to store fuel in a dry environment. I 1.1 Design Description The Beaver Valley Unit I fresh fuel rack layout is shown in Figure 1 on page 10. The Beaver Valley Unit I fresh fuel rack storage cell design is shown in Figure 2 on page 11. l The fuel parameters relevant to this analysis are given in Table 1 on page 9. The OFA fuel rod parameters will result in conservative K egvalues for all of the fuel types mentioned above and as i a result are used for all supporting dimensional calculations performed for this analysis. With the simplifying assumptions employed in this analysis (no grids, sleeves, axial blankets, etc.), the fuel parameters given in Table 1 on page 9 will also result in bounding K eg values for Westinghouse  !

17x17 VANTAGE-5H (V5H), Vantage +, and Performance + assembly types.

1.2 Design Criteria Criticality of fuel assemblies in a fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction. This is done by fixing the minimum separation between fuel assemblies.

The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a 95 percent probability at a 95 percent confidence level that the effective neutron multiplication factor, K,g, of the fuel assembly array will be less than 0.95 as recommended by ANSI 57.2-1983 t2 ), ANSI 57.3-1983(3) and NRC guidance W, and less than 0.98 under low water density (optimum moderation) conditions as recommended by NUREG-0800(5),

1 Beaver Valley Unit 1 Fresh Fuel Racks

2.0 Analytical Methods 2.1 Criticality Calculation Methodology The criticality calculation method and cross-section values are verified by comparison with  ;

1 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 uncertainty will apply to rack conditions which include strong neutron absorbers, large water gaps and low i moderator densities.

The design method which insures the criticality safety of fuel assemblies in the fuel storage rack is described in detail in the Westinghouse Spent Fuel Rack Criticality Analysis Methodology topical repoct(6). This report describes the computer codes, benchmarking, and methodology which are used to calculate the criticality safety limits presented in the report for the Beaver Valley Unit I fresh fuel storage racks.

As determined in the benchmarking in the topical report, the method bias using the described methodology of NITAWL-II, XSDRNPM S, and KENO-Va is 0.0077 AK with a 95 percent probability at a 95 percent confidence level uncertainty of 0.0030 AK. These values will be used throughout this report as needed.

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Beaver Valley Unit 1 Fresh Fuel Racks

e 3.0 Criticality Analysis of the Fresh Fuel Racks This section describes the analytical techniques and models employed to perform the criticality analysis for the storage of fresh fuel in the Beaver Valley Unit I fresh fuel storage racks. The complete criticality analysis of the fresh fuel racks is performed in two separate steps using the methodology outlined in Section 2 of this report. The two steps of this complete analysis are outlined below:

Step i The fresh fuel racks are analyzed for the full density water Hooding condition under nominal and " worst case" scenarios (highest enrichments of 5.0 w/o and 5.05 w/o U235 l respectively were considered). The KENO results for the " worst case" model are then used to develop the maximum Ke g which is compared to the criticality safety limit of 0.95. Details of this analysis are outlined in Section 3.1.

Step 2 The fresh fuel racks are analyzed for optimum moderation water Sooding conditions under the " worst case" scenario (5.05 w/o U235). The KENO results are then used to l develop the maximum K,g which is compared to the criticality safety limit of 0.98. .

I Details of this analysis are outlined in Section 3.2.

Since the fresh fuel racks are normally maintained in a dry condition, the criticality analysis will show that the rack K,g is less than 0.95 for the accidental full water density Hooding scenario and less than 0.98 for the accidental low water density (optimum moderation) Booding scenario.

The following assumptions were used to develop the KENO model for the storage of fresh fuel in l

. the Beaver Valley fresh fuel storage racks under full density and low density optimum moderation i conditions: l

1. The fuel assembly parameters relevant to the criticality analysis are based on the i Westinghouse 17x17 OFA design (see Table 1 on page 9 for fuel parameters). The OFA fuel

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rod parameters will result in conservative K,gvalues for all Westinghouse 17x17 fuel types.

. 2. The fuel assembly is modeled at its most reactive point in life.

] 3. All fuel rods contain uranium dioxide at an enrichment of 5.0 w/o (nominal) and 5.05 w/o

! (" worst case") over the entire length of each rod for the full density Hood scenario.

4. All fuel rods contain uranium dioxide at an enrichment of 5.05 w/o (" worst case") over the j entire length of each rod for the optimum moderation Hood scenario. i

5. The fuel pellets are modeled assuming a UO2 density which is 95 % of theoretical density l with a dishing fraction of 1.2110 % for nominal conditions, and a UO2 density which is 97 % l l of theoretical density with a dishing fraction of 0.0 % for " worst case" conditions.

6. No credit is taken for any natural or reduced enrichment axial blankets.

7. No credit is taken for any U 234 or U 236 in the fuel, nor is any credit taken for the buildup of fission product poison material.

8. No credit is taken for any spacer grids or spacer sleeves.

9. No credit is taken for any bumable absorber in the fuel rods.

10. For both the full density and optimum moderation cases, there is no boron in the water.

3 Beaver Valley Unit i Fresh Fuel Racks

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I1. Fuel rods are modeled with a fuel stack height which is 144 inches long for the full density scenario and the optimum moderation scenario.

3.1 Full Density Moderation Analysis ,

In the KENO model for the full density moderation analysis, the moderator is pure water at a 3

conservative density of 1.0 gm/cm (68 F). The fuel array is infinite in the lateral (x and y) extent and finite in the axial (vertical) extent which precludes any neutron leakage from the array in the x l and y directions. I The KENO calculation for the nominal case resulted in a K,g of 0.9096 with a 95 percent i probability /95 percent confidence level uncertainty of 0.0025 AK.

The maximum K,g for the full density flooding scenario arises from consideration of mechanical and material thickness tolerances resulting from the manufacturing process in addition to asymmetric positioning of fuel assemblies within the storage cells. Rack tolerances considered in this analysis were a +/- 1/16" tolerance on the 21" center to center spacing and a +/- 1/64" tolerance on the storage cell thickness of 0.125". The minimum values for each of these parameters was assumed to result in the most conservative K,g value. Furthermore, fuel parameters of 5.05 w/o U 235 ,97 % theoretical density and 0 % dishing fraction are assumed to conservatively account for l fuel parameter variability. A KENO model was set up using the above limiting fuel and rack  !

parameters which will be called the " worst case" full density flooding scenario.

The KENO calculation for the " worst case" full density flooding scenario resulted in a Ke g of 0.9183 with a 95 percent probability /95 percent confidence level uncertainty of 0.0024 AK. 4 The following equation is used to develop the maximum Ke g for the Beaver Valley Unit I fresh fuel storage racks under full density water conditions:

K,g = K ;;p + B,,,sog + ,jkspig 2+ ks m,,sog 2 where:

K3ff = maximum KENO K,g with full density water B method = method bias determined from benchmark critical comparisons kspff = 95/95 uncertainty in the full density water KENO Ke g ks method = 95/95 uncertainty in the method bias Substituting calculated values in the order listed above, the result is:

K,g = 0.9183 + 0.0077 + J0.00242 + 0.00302 = 0.9298 Beaver Valley Unit i Fresh Fuel Racks 4

Since K,g is less than 0.95 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met for the Beaver Valley Unit i fresh fuel storage racks under full density water flooding conditions for storage of Westinghouse 17x17 fuel assemblies with nominal enrichments up to 5.0 w/o U 235 3.2 Low Density Optimum Moderation Analysis For the low density optimum moderation analysis, the fuel array is finite in all directions. " Worst

case" assumptions on UO2 material properties and spent fuel rack dimensions are used in modeling the entire fresh fuel rack array. All available fresh fuel storage cells are utilized as depicted in Figure 1 on page 10. Fuel rods are modeled with a nominal active fuel length of 144.

inches. The fresh fuel storage rack is assumed to be surrounded by concrete in all directions.

i Under low water density conditions, the presence of concrete is conservative because neutrons are reflected back into the fuel array more efficiently than they would be with just low density water.

Minimum distances are assumed between the concrete wall and the fresh fuel racks to maximize the neutron reflection back into the fresh fuel array.

Analysis of the Beaver Valley fresh fuel racks shows that the maximum rack K,g under low 3

density moderation conditions occurs at 0.07 gm/cm water density. The Ke g of the fresh fuel racks at 0.07 gm/cm3water density is 0.9271 with a 95 percent probability /95 percent confidence level uncertainty of 0.0022 AK. Figure 3 on page 12 shows the fresh fuel rack reactivity as a function of water density in the range where the optimum moderation 3 peak occurs. Over the remainder of the density range between 0.10 gm/cm and 1.00 gm/cm (fully flooded), the rack i reactivity will be less than the values calculated for the optimum moderation peak or for the full j density condition. The fresh fuel rack K,g value obtained at the optimum moderation condition

(0.07 gm/cm3 water density) will bound any K,g value which results due to the presence of aqueous foam in the fresh fuel racks.

The following equation is used to develop the maximum Ke g for the Beaver Valley Unit I fresh fuel storage racks under low water density optimum moderation conditions:

K,g = K op ,,,,, + B,,,s,g + ,]ks optimum + ks,,,s,g 2 where:

i K,y,,,,, = maximum KENO K,g with low density optimum moderation

. B ,, hod = method bias determined from benchmark critical i comparisons ks,y,,,,, = 95/95 uncertainty in the low density optimum moderation KENO K,g

. ks,,,3,j = 95/95 uncertainty in the method bias 5

Beaver Valley Unit 1 Fresh Fuel Racks

1 Substituting calculated values in the order listed above, the result is: l K,g = 0.9271 + 0.0077 + J0.00222 + 0.00302 = 0.9385 Since K eg is less than 0.98 including uncertainties at a 95/95 probability / confidence level, the acceptance criteria for criticality is met.

6 Beaver Valley Unit i Fresh Fuel Racks

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4.0 Discussion of Postulated Accidents l l

4.1 Fresh Fuel Storage Racks Under normal conditions, the fresh fuel racks are maintained in a dry environment. The I introduction of water into the fresh fuel rack area is the worst case accident scenario. The water flooding cases analyzed in this report are bounding accident situations which result in the most l

conservative fuel rack Keg.

Other accidents can be postulated which could cause a reactivity increase in the fresh fuel racks and these are a fuel assembly drop on top of the rack and a fuel assembly drop between the rack and the wall. The fuel assembly drop between the rack and the wall is not possible for the Beaver i Valley Unit I fresh fuel racks due to the presence of a stainless steel sheet on top of the racks l which precludes the drop of an assembly into any position other than a storage cell. For the fuel l assembly drop on top of the rack, the double contingency principleO is applied. This states that i one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for the case of the fuel assembly drop on top of the rack, the j absence of a moderator in the fresh fuel storage racks can be assumed as a realistic initial condition since assuming its presence would be a second unlikely event.

Experience has shown that the maximum reactivity increase associated with a fuel assembly drop l on top of the rack is less than 10 percent AK.

Therefore, since the normal, dry fresh fuel rack reactivity for Beaver Valley Unit 1 is less than 0.65, and the maximum reactivity increase for the fuel assembly drop on top of the rack is less than 10 percent AK, the maximum rack K,g for the fuel assembly drop on top of the rack will be less than 0.95.

7 Beaver Valley Unit 1 Fresh Fuel Racks

t 5.0 Summary of Criticality Results The acceptance criteria for criticahty requires the effective neutron multiplication factor, Keg, in

' the fresh fuel storage racks to be less than or equal to 0.95, including uncertainties, under flooded conditions and less than or equal to 0.98, including uncertaimies, under optimum moderation conditions.

This report shows that the acceptance criteria for criticality is met for the Beaver Valley Unit I fresh fuel storage racks for the storage of Westinghouse 17x17 STD, OFA, V5H, Vantage +, and i Performance + fuel assemblies with the following configuration and enrichment limits:

i Fresh Fuel Racks Storage of fresh fuel assemblies with nominal enrichments up to 5.00 w/o U 235 utilizing all available storage ce.ls.

The analytical methods employed herein conform with ANSI N18.2-1973(8), " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," Section 5.7, Fuel Handling System: ANSI 57.3-1983(3) " Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants"; and ANSI N16.9-1975 l9 ), " Validation of Calculational Methods for Nuclear Criticality Safety".

Beaver Valley Unit 1 Fresh Fuel Racks 8

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Table 1. Fuel Parameters Employed in the Criticality Analysis 4

Parameter W 17x17 OFA j Number of Fuel Rods per Assembly 264 Rod Zirc-4 Clad O.D. (inch) 0.3600

' 1 1

Clad Thickness (inch) 0.0225 j Fuel Pellet 0.D.(inch) 0.3088 Fuel Pellet Density (% of Theoretical) 95 Fuel Pellet Dishing Factor (%) 1.2110 Rod Pitch (inch) 0.496

Number of Zirc-4 Guide Tubes 24 0.474 Guide Tube O.D. (inch) i Guide Tube Thickness (inch) 0.016 Number ofInstrument Tubes 1 Instrument Tube O.D. (inch) 0.474 l Instrument Tube Thickness (inch) 0.016  ;

1 Beaver Valley Unit i Fresh Fuel Racks 9

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Figure 1. Beaver Valley Unit 1 Fresh Fuel Rack Array Layout 10 Beaver Wiley Unit i Fresh Fuel Racks

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Figure 2. Beaver Valley Unit 1 Fresh Fuel Rack Cell Layout 11 Beaver Wiley Unit 1 Fresh Fuel Racks

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Figure 3. Sensitivity of Kgr to Water Density in the Beaver Valley Unit i Fresh Fuel Racks 12 Beaver Wiley Unit i Fresh Fuel Racks

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d Bibliography i

1

l. W. A. Boyd, M. W. Fecteau, D. E. Mueller, Criticalin' Analysis ofBeaver Vallev 1 Fresh and Spent FuelRacks. August 1986.

I

2. l American Nuclear Society, American National Standard Design Requirementsfor Light 3 Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants ANSilANS-57.2-1983, October 7,1983, l 3. American Nuclear Society, American National Standard Design Requirementsfor New Fuel Storage Facilities at Light Water Reactor Plants, ANSIIANS-57.3-l983,1983. l

4. Nuclear Regulatory Commission, Letter to All Po ver Reactor Licensees from B. K. Grimes,

]

1

OTPositionfor Review and Acceptance ofSpent l'uel Storage and Handling Applications, April 14,1978.

i

5. U. S. Nuclear R egulatory Commission, Standard Review Plan, NUREG-0800, July,1981.

i 6. W. D. Newmyer, Westinghouse Spent Fuel Rack Cridcality Analysis Methodology, WCA?-

14416-NP-A, November 1996.

i 7. American Nuclear Society, American National Standardfor Nuclear Criticality Safety in l Operations with Fissionable Materials Outside Reactors, ANSIIANS R.1-l983,0ctober 7,

[:

1983.

i~ 8. American Nuclear Society, Nuclear Safety Critteriafor the Design ofStationary Pressur-i ired Water Reactor Plants, Section 5.7, ANSI /ANS-N18.2-1973,1973.

9. Americar Nuclear Society, Validation ofCalculational Methodsfor Nuclear Criticalin' 2

Safeo', ANSI /ANS-N16.9-1975,1975.

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!1 j

l i

Beaver Valley Unit i Fresh Fuel Racks 13

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j ATTACHMENT D

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Beaver Valley Power Station, Unit No. 1 i

Proposed Technical Specification Change No. 241 i

1 Technical Specification Change Nos. 234 and 107 Attachment B l

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+ 1 1 I l

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ATTACHMENT B Beaver Valley Power Station, Unit Nos. 1 and 2 Proposed Technical Specification Change Nos. 234 and 107 INCORPORATE ISTS DESIGN FEATURES A. DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would modify the technical specification ,

(TS) design feature section to reflect those provided in the l Improved Standard Technical Specifications (ISTS) of NUREG-1431 Revision 1. l l

B. BACKGROUND l

. The current design fee.ture section contains information based on Revision 0 of NUREG-0452. The proposed amendment updates this section to contain information based on the latest version of the ISTS. A previous submittal for Unit 2 dated April 29, 1996, proposes specific changes to Section 5.3.1, which if approved, will be integrated into these proposed changes.

C. JUSTIFICATION 10 CFR 50.36 sets forth those categories of information required to be included in the TS. The design features is one of those categories and the types of information to be included are those such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in other sections of the TS. The ISTS were developed to specifically comply with these requirements and to provide a model for plants to use when developing or modifying the TS design feature section. Much of the information contained in the current design feature section is redundant to information provided in various sections of the UFSAR and is not needed to comply with the requirements of 10 CFR 50.36. The NRC issued a Final Policy Stateusni in 1993 which identified four criteria to be used in determinir.g those requirements to be included in TS.

The information being removed from the TS design feature section does not fall within these four criteria. Therefore, the design features can be modified to reflect the ISTS requirements without any loss of the current information or affecting the safety of the plant.

D. SAFETY ANALYSIS The TS design feature section has been modified by incorporating the ISTS wording in our format and removing the information not specified in the ISTS. Section 5.1 has been re-titled to Site Location and includes a text description of the site in place of reference to BV-1 Figures 5.1-1 through 5.1-6 and BV-2 Figures 5.1-1 through 5.1-4. The site boundary, provided in Figure 5.1-1 for both units, is maintained in the offsite dose calculation manual (ODCM) Figure 5-1. Changes to the ODCM are implemented in accordance with the provisions indicated in the administrative

a;.'.j, ATTACHMENT B, continusd Propossd Tschnical Spscification Changs Nos. 234 and 107 Page 2 control section of the . technical specifications. Reference to Unit 1 Figure 5.1-2 in Section 5.1.2 has been deleted because this figure was previously deleted by Amendment No. 121 since Figure 5.1-1 provides the site boundary for both gaseous and liquid effluents. The exclusion area boundary, provided in Figure 5.1-3 for both units, is described in Unit 1 UFSAR Section 2.1.1 including UFSAR Figure 2.1-7 for Unit 1, and in Unit 2 UFSAR Section 2.1.2.1 including Figure 2.1-2 for Unit 2. The low population zone, provided in Figure 5.1-4 for both units, is described in Unit 1 UFSAR Section 2.1.2 for Unit 1 and in Unit 2 UFSAR Section 2.1.3.5 fcr Unit 2. The gaseous and liquid release points provided in Unit 1 Figures 5.1-5 and 5.1-6 and Unit 2 Figure 5.1-2 are maintained in ODCM Figure 2.4-2 for the gaseous release points and Figure 1.4-1 for Unit 1 and Figure 1. 4-2 for Unit 2 for 'the liquid. release points. The flood control information provided in Section 5.1.5 is addressed in Unit 1 UFSAR Section 2.3.3 and Unit 2 UFSAR Section 3.4.1. Section 5.2, Containment, has been removed since this information is'provided in Unit 1 UFSAR Sections 5.2.1, 5.2.2, 5.2.4 and Table 14.3.3a and Unit 2 UFSAR Sections 3.8.1, 6.2.1, 6.2.4 and Table 6.2-3.

Section 5.3, Reactor Core, has been renumbered to 5.2. The text for Section 5.2.1, Fuel Assemblies, and 5.2.2, Control Rod Assemblies, remains unchanged for both units. Section 5.4, Reactor Coolant System, has been removed since this information is provided in Unit 1 UFSAR Sections 4.1 and 4.2 and Unit 2 UFSAR Sections 5.2, 5.3 and.5.4. Section 5.5, Emergency Core Cooling Systems, has been removed; this information is provided in UFSAR Section 6.3 for both units. Section 5.6, Fuel Storage, has been renumbered to 5.3. The text ior renumbered Section 5.3.1, Criticality, has been replaced with Section S.3.1.1 for the spent fuel storage racks and Section 5.3.1.2 for the new fuel storage racks like the ISTS. Section 5.3.1.1 provides the spent fuel storage rack design criteria where item "a" limits fuel assembly maximum enrichment to ' that specified in Specification 3.9.14.

This is different from the-ISTS which lists a specific enrichment number in weight percent; however, this is intended to reduce the possibility of error since the enrichment limit is specified in only one location. Item "b" limits the K.gf to S 0.95 when the storage racks are flooded with unborated water and includes reference to the applicable UFSAR section where the uncertainties are described. Item "c" specifies the spent fuel storage rack cell center to center distance separation; this is consistent with the current criteria and ensures a change to the rack-design cannot be made without prior NRC approval. Item "d" refers the reader to Specification 3.9.14 since this specification provides the spent fuel pool fuel assembly storage requirements. Section 5.3.1.2 provider the new fuel storage rack design criteria, where item "a" limits fuel assembly maximum enrichment to 4.5 weight percent for Unit 1 and 4.85 weight percent for Unit 2. Item "b" limits the Kegg to S 0.95 including applicable uncertainties when the new fuel storage racks are flooded with unborated water, this is consistent with Unit 1 UFSAR Section 9.12 and Unit 2 UFSAR Section 9.1. Item "c" limits the Kegg to S 0.9S including l

B-2 W "

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ATTACHMENT B, continund '

Propo cd T chnical Sp cificction Chnnga No3. 234 and 107 i Page 3 applicable uncertainties when the new fuel storage racks are moderated by aqueous foam; this is consistent with Unit 1 UFSAR  ;

Section 9.12 and Unit 2 UFSAR Section 9.1 which refer to low l density optimum moderation conditions. Item "d" specifies the 21 l inch new fuel storage cell nominal center to center distance l botveen fuel assemblies. This is consistent with that specified l in Unit 1 UFSAR Section 9.12 and Unit 2 UFSAR Section 9.1.

Section 5.6.2, Drainage, and Section 5.6.3, Capacity, have been {

renumbered to 5.3.2 and 5.3.3, respectively; the text remains '

unchanged. Section 5.7, Seismic Classification, has been removed since this information is provided in Unit 1 UFSAR Appendix B and I

Unit 2 UFSAR Section 3.7. Section 5.8, Meteorological Towei Location, has been removed since this information is provided in ODCM Figure 5-1.

The proposed change removes information that is redundant to that provided in the UFSAR, is not in the ISTS, and is not required to be listed in the TS design features section in accordance with 10 CFR 50.36. As a result of reducing the information presented, the TS index has been updated to reflect the change in section and page numbers. Format and editorial changes along with additional information describing the new fuel storage racks has been added for consistency with the ISTS. These changes do not affect any of the UFSAR accident analyses and are consistent with the design of the plant; therefore, these changes have been determined to be safe and will not reduce the safety of the i plant.

E. NO SIGNIFICANT HAZARDS EVALUATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three l standards set forth in 10 CFR 50.92(c) as quoted below:

l The Comnission may make a final determination, pursuant to  ;

the procedures in paragraph 50.91, that a proposed amendment i to an operating license for a facility licensed under i paragraph 50.21(b) or paragraph 50.22 or for a testing  !

facility involvos no significant hazards consideration, if l operation of the facility in accordance with the proposed amendment would not: I i

(1) Involve a significant increase in the probability or  !

consequences of an accident previously evaluated; or l (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or  !

(3) Involve a significant reduction in a margin of safety.

The following evaluation is provided for the no significant hazards consideration standards.

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ATTACHMENT B, continusd Propossd Tschnical Spacification Change Nos. 234 and 107 Page 4

! 1. Does the change involve a significant increase in the

probability or consequences. of an accident previously

evaluated?

4 The proposed change reduces the content of the technical 1 specification (TS) design feature section consistent with the Improved Standard . Technical Specifications (ISTS) of NUREG-  ;

1431. The information that has been removed is 'also ;

contained in the UFSAR or offsite dose calculation manual i

(ODCM); therefore, duplication of the information is  !

j eliminated to improve the use of the TS. Because the i 2

information removed from the TS is maintained in the UFSAR or  !

! ODCM where changes are controlled in accordance with I regulatory requirements, there is no reduction in commitment  !

and adequate control is provided. Elimination of information I i

from the design feature section of the TS which duplicates 1 information in the UFSAR enhances the usability of the TS l

! without reducing commitments. These changes clarify and i i

improve the understanding and readability of the TS. 'Since l t

the requirements remain the same, these changes only affect

the method of presentation and would not affect possible l l initiating events-for accidents previously evaluated or any '

system functional requirement. Therefore, the proposed 1 i changes would not involve a significant increase in the l

probability or consequences of an accident previously  ;

! evaluated.

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2. Does the change create the possibility of a new or different l l kind of accident from any accident previously evaluated? l The relocation of existing requirements, the elimination of l requirements which duplicate existing.information, and making administrative improvements are all changes that are administrative in nature. The proposed changes will not affect any plant system or structure, nor will they affect any system functional or operability requirements.

Consequently, no new failure modes are introduced as a result of the proposed changes. The proposed changes are consistent with the ISTS, for the most part, as plant-specific information is included in this section. Therefore, the proposed change will not create the possibility of a new or different type of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of safety?

The proposed changes are administrative in nature in that no change to the design features of the facility are being made.

The .Sesign features section is being reformatted to be j consistent, for the most part, with the ISTS. The proposed i changes do not affect the UFSAR design bases, accident j assumptions, or technical specification bases. In addition, '

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' ATTACHMENT B, continu:d Proposcd T;chnien1 Spacification Changa Nos. 234 and 107 Page 5 the proposed changes do not affect release limits, monitoring i equipment or practices. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

4 F. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION

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Based on the considerations expressed above, it is concluded that the activities associated with this license amendment request i satisfy the no significant hazards consideration standards of 10 CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified.

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