Text

Proposed Tech Specs Allowing Use of Higher Enrichment (Up to 4.0 W/O U-235) Fuel for Future Core Reloads on Facilities
	ML20238F426
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Site:	Catawba
Issue date:	09/08/1987
From:	; DUKE POWER CO.
To:	;
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ML20238F422	List: ML20238F421; ML20238F426;
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ATTACHMENT 1 PROPOSED CATAWBA UNITS 1 AND 2 TECHNICAL SPECIFICATION Cl!ANGES l

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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE' 5.2.2. The reactor containment vessel is designed and shall be maintained for a maximum internal pressure of 15 psig and a temperature of 328'F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The core'sha11'contain 193 fuel assemblies with each fuel assembly containing 264' fuel rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 144 inches. The initial core loading shall have a maximum enrichment of 3.5 weight percent U-235. Reload fuel shall be similarJn physical d to the initial corploading and shall have a maxis {enrif

**1'nt of 34 eight 9.0 percent U- 35fcwilh a taaAmvn enschrned leierane CONTROL RODTSSEMBLIE ofiO.oS wei ht percent V-235

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5.3.2 The core shall contain 53 full-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber .

material of which 102 inches shall be 100% boron carbide and remaining 40-inch tip shall be 80% silver, 15% indium, and 5% cadmium. All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure a 2485 psig, and

c. For a temperature of 650*F, except for the pressurizer which is 680*F.

j VOLUME i

5.4.2 The total water and steam volume of the Reactor Coolant System is {

12,040

100 cubic feet at a nominal T,yg of 525*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown in Figure 5.1-1.

CATAWBA - UNITS 1 & 2 5-6 Amendment No. ~(Unit 1) {

.(Unit 2)

Amendment No.

ATTACHMENT 2 JUSTIFICATION AND SAFETY ANALYSIS l

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ATTACHMENT 2 JUSTIFICATION AND SAFETY ANALYSIS

. Based on_ analyses which indicate that fuel. cycles longer than 12 months will be more economical for Duke Power (longer cycles can facilitate higher nuclear-l capacity factors and lower Duke's overall generation costs), longer fuel cycles

!' -(330+ EFPD).are planned for Catawba Nuclear Station which require high' enrichments.

'Since future reload fuel regions will have_ enrichments exceeding the current

,. reFctor. core technical' specification reload fuel' maximum enrichment limit of 3.5

!' weight percent U-235 (enrichments between 3.5 w/o and 4/0 w/o will be required to realize the economic advantages of these longer-cycles), the proposed amendments seek to revise the Technical Specification 5.3.1 maximum allowable fuel enrichment to 4.0 w/o for both units.

The 3.5 weight percent reload fuel limit in Technical Specification Section 5.3.1 is based on the criticality analyses for the new and spent-fuel storage facilities i: Jat Catawba Nuclear Station (reference FSAR Sections 9.1.1 and 9.1.2). The spent fuel racks are described in FSAR Section 9.1.2, with the storage capabilities governed by Technical Specification section 5.6. The storage capabilities of the new fuel storage vault are also governed by Technical Specification Section 5.6, and are further discussed in FSAR Section 9.1.1 (which indicates that the {

criticality evaluation performed for the new fuel storage vault assumes assemblies ;

to be 3.5 w/o enrichment U-235 and unirradiated). Attachment 2A demonstrates a 4.0 w/o storage capability with a maximum enrichment tolerance of 0.05 w/o U-235 for normal storage _and accident conditions including a conservative margin for enrichment variance for both the new and spent fuel storage facilities.

The reactor core is similarly capable of handling 4.0 (and higher) weight percent reload fuel. This capability will be demonstrated as necessary in the cycle-specific reload safety evaluations (RSE) which are performed prior to fuel

-loading (the RSE's consider the standard reload design methods described in WCAP-9272 and 9273, " Westinghouse Reload Safety Evaluation Methodology", and/or-other appropriate criteria to demonstrate that the core reload'will not adversely affect the safety of the plant). Criticality accidents during refueling operations are precluded by stringent administrative procedures.

Consequently, based-on the above evaluation it is concluded that use of reload- j fuel enriched up to 4.0 w/o U-235 with a maximum enrichment tolerance of 0.05 1 weight percent U-235 at Catawba Nuclear Station is acceptable, and it is requested that Technical Specification 5.3.1 should be revised accordingly. No ;

other Technical Specification changes are required for this enrichment upgrade. 1 l

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j ATTACHMENT 2A CRITICALITY ANALYSES FOR THE CATAWBA ]

FUEL STORAGE FACILITIES 1

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n .4 CRITICALITY ANALYSES FOR THE. CATAWBA FUEL STORAGE FACILITIES

l. - 1.0 CAIAWBA NEW FUEL VAULT ANALY515 1.1' CRITICALITY DESIGN CRITERIA

' Criticality.of fuel stored in the new fuel vault is prevented by the design'of the ; storage racks which maintain a minimum separation between stored fuel assemblies.

The design of the new fuel storage facility is based on the criteria presented in . Catawba ~ FSAR Section 9.1.1.1. The design basis for preventing criticality in the new fuel vault is taken from ANSI N18.2-1973, Section 5.7.4.1, which 1 states: I "The design of spent. fuel storage racks and transfer-equipment shall be such that the effective multiplication factor will not' exceed 0.95 with new fuel of the highest i anticipated enrichment in place assuming flooding with !

pure water. The desigri of normally dry new fuel storage racks shall be such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place assuming optimum moderation."

The following accidents are considered in the criticality design of the new fuel storage vault:

(A) Flooding: complete immersion of the entire array in pure, unborated l full density water.

(B) Envelopment of the entire array in a uniform density aqueous foam of optimum moderation density (that density' which maximizes the reacti-vity of the array).

Accidents resulting in an increase in k-eff because of geometrical changes of

'the racks are not considered credible due to the following design bases:

.(A) The new fuel racks are designed to withstad normal operating loads )

as well as SSE seismic loads meeting ANS Safety Class 3 and AISC '

requirements.

(B) The new fuel storage racks are located in the New Fuel Storage Buildings which protect the racks from weather conditions and external forces such as those resulting from tornado or wind loads.

The nominal . center-to-center spacing of 21 inches is sufficient to meet the (

criticality criteria of ANSI N18.2-1973 under the postulated conditions of '

complete flooding with unborated water or an optimum moderator. The racks are designed to prevent the insertion of fuel between the storage positions.

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I 1.2 FACILITY DESCRIPTION Each unit of the Catawba Station has an independent new fuel storage system. The New Fuel Storage Buildings are Seismic Category I, reinforced concrete structures. The structures are designed to withstand static and SSE seismic l loads as well as tornado generated missile impact loads. The New Fuel Storage Racks are contained entirely within the structure and further segregated from other equipment and operations by a reinforced concrete wall.

New fuel for each unit may be stored in the Spent Fuel Pool or stored dry in racks which are bolted to the floor of the New Fuel Storage Buildings. The new fuel racks are designed to accommodate 98 fuel assemblies for each unit at a .,

nominal center-to-center spacing of 21 inches. Only 92 of these racks may actually be used because six (6) are inaccessible with the overhead crane.

Storage cells are formed by 1/8 inch nominal thickness, minimum cell wall thickness 0.12 inches, type 304 stainless steel that completely encloses the fuel on four sides, whereas the supporting racks are fabricated from painted carbon steel conforming to AISC tolerances and specifications.

The nominal fuel cell interior dimension is nine inches square with all interior edges finished to a minimum 1/16 inch radius of chamfer. If chamferred, all f intersecting edges are blended. All interior surfaces which may come in contact with the fuel assemblies are smooth and clean of all weld spatter dirt, and grease. Design conditions that could cause hang-up during insertion or withdrawal of fuel assemblies have been avoided.

General arrangement diagrams of the Unit 1 New Fuel Storage Buildings and racks are provided in Catawba FSAR Figures 9.1.1-1 and 9.1.1-2 (Attachments 1 and 2, respectively). Unit 2 facilities are a mirror image of Unit 1 facilities. .

l 1.3 CRITICALITY ANALYSIS METHOD The analysis method which ensures the criticality safety of fuel assemblies in the new fuel storage racks uses the Criticality Analysis Sequence No. 2 (CSAS2) and the 123GROUPGMTH naster cross-section library included in the SCALE-3 system of codes (Reference 1). CSAS2 consists of two cross-section processing codes (NITAWL and BONAMI), a 1-D transport code for cell-weighting cross-section data (XSDRNPM), and a 3-D monte-carlo code (KENO-IV) for calculating the effective multiplication factor for a system. ,

CSAS2 and 123GROUPGNTH cross-sections are operational on the Duke Fairview IBM computer system. A set of 40 critical experiments have been analyzed using the CSAS2/123GROUPMTH reactivity calculation method to demonstrate its applicability to criticality analysis and to establish method bias and variability. The experiments analyzed represent a diverse group of water moderated, oxide fuel arrays separated by various materials (stainless steel, Boral, water, etc.) that are representative of LWR shipping and storage conditions, including the Catawba new and spent fuel storage racks. The set of 40 critical experiments analyzed using the above method are summarized in Table 1.

The average K-eff of the benchmarks is 0.9951 with a standard deviation of 0.0059 l

AK. The 95/95 one-sided tolerance limit factor for 40 values is 2.13.

Therefore, there is a 95 percent probability at at 95 percent confidence level that the uncertainty in reactivity, due to the method, is not greater than 0.012 AK.

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1.4 CRITICALITY EVALUATION

S A number of criticality analyses considering a full loading of either Westinghouse 17x17 Standard (STD) or Optimized (0FA) fuel assemblies are performed using aqueous moderator densities ranging from 0.05 to 1.0 gm/cc.

Evaluating Lneutron multiplication over this range of moderator densities is necessary to demonstrate compliance with the design basis lfor preventing criticality in the new fuel vault as ' provided by ANSI N18.2-1973, Section 5.7.4.1.

The following assumptions were used in the criticality evaluation:

1) Fuel assembly parameters modeled in, each case are summarized in Table 2. ?

2) Credit is taken for the inherent neutron absorption in full length structural materials as allowed by ANSI N18.2-1973.

3) No ' burnable poisons, control rods, th supplements'l neutron poisons are assuned to be present.

4) Effects

of r3flectors other than water are included if their neglect would have been nonconservative. This includes the storage vault's concrete walls, ceiling, and floor.

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5) All assemblies are assumed to be 4.1 w/o U-235 enriched and unirradi-ated. This worst-case enrichment assumption allows for a specified maximum nominal enrichment of 4.0 w/o U-235 with an enrichment tolerance of <_ 0.1 w/o U-235.

6) The new fuel storage vault is conservatively modeled as an infinite series of 2 infinite rows of 12 foot high fuel assemblies in minimal thickness SS304 cell enclosures.

7) Each fuel assembly is treated as a heterogeneous system with the fuel pins, control rod guide tubes, and instrumentation thimble ,

guide tube modeled explicitly. !

8) Mechanical uncertainties and biases due to construction tolerances are considered by using worst-case conditions. Uncertainties consi-dered include cell I.D. , center-to-center spacing, and cell enclosure i thickness.

In order to address / accidents involving flooding of the new fbel storage l facilities with full density and optimum density unborated water, eighteen ~(18) criticality analyses are performed for each fuel type considered (i.e.,

Westinghouse 17x17 STD and 0FA fuel types.) The 18 analyses performed for each fuel type case use aqueous moderator densities ranging from 0.02 to 1.0 gm/cc. ,

I The K-eff for a full loading of new fuel in the new fuel storage racks are ]

calculated in each fuel type / moderator density case as follow:

+ s

S Keff = KN+Bmech + Bmeth + [(Ksn ) mech) meth)

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where:

K.N

= n minal case Keff calculated by KENO-IV = Table 3.

B mech

= bias to account for mechanical tolerances which can increase K above the nominal case values.

eff

= 0.0 AK (mechanical tolerances are considered by using j worst-case conditions in the nominal case KENO-IV models).

B meth

= method bias as determined by benchmark critical experiments.

= 0.0049 AK Ks n

= 95 percent probability, 95 percent confidence level uncertainty in the nominal case K,ff. l

= Table 3 Standard Deviations x 2.0.

Ks mech

= 95 percent probability, 95 percent confidence level uncertainty in mechanical tolerance bias.

= 0.0 AK (mechanical tolerances are considered by using worst-case conditions in the nominal case KENO-IV models).

Ks meth

= 95 percent probability, 95 percent confidence level uncertainty in the method bias.

= 0.012 AK Substituting calculated values results in final rack K-eff values which include all biases and uncertainties as required by ANSI /ANS-57.2-1983, Sec-tion 6.4.2.2.1 for demonstrating compliance with criticality design criteria.

Criticality analysis results are summarized graphically in Figures 1 and 2 for Westinghouse 17x17 STD and 0FA fuel types, respectively. Figures 1 and 2 show nominal K-eff values calculated in each fuel type / moderator density case as well as error bars which include the calculational biases and 95/95 uncertain-ties discussed above.

1. 5 NEW FUEL VAULT ANALYSIS RESULTS AND CONCLUSIONS The criticality analysis results illustrated by Figures 1 cnd 2 demonstrate that 4.1 w/o U235 enriched Westinghouse STD and 0FA fuel can be safely stored in the Catawba new fuel storage vaults in accordance with ANSI N18.2-1973, Section 5.7.4.1 criteria.

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9 2.0; CAIAWSA SPENT FUEL STORAGE RACK AfMLYSJS yi )

2.1 4 CRITICALITY D #IGN CRITERIA-f' 'b

' Criticality of fuel, stored in the: spent fuel pool is prevented by the design of thel storage . racks : which maintain a minimum separation between stored fuel

' assemblies. '

. The design of the spent fuel storage . facility. 'is based on~ criteria presented l in Catawba FSAM Section 9.1.2.1. The design basis for preventing criticality ;y

, in the spent L fuel pool is taken from ANS( N210-1976 and ANSI N18.2-197a r _y Section 5.7.4.1, which states: c5 g y .

s "The design of spent ~ fuel storage racks and transfer ,,

equipment shall be such that the effective multiplication I

'b factor will not exceed 0.95 with new fuel of the highest anticipated enrichment in place assuming flooding with pure water. The design of normally dry new fuel storage 5 racks shall be such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place assuming optimum moderation.' y Postulated accident conditious - do not resu)t in an increase in K-eff valuet .

-beyond those - calculated for [the infinite" array in the normal design basis"./ Y ;

analyses. -

Accidents considered include: 1). loss cf spent fuel pool cooling, 2) the 1 sliding of free standing ract modules ,such that peripheral cells of two rack' j modules have C-C spacings belw those/ assumed in normal design basis analyses,

.and 3) the dropping of fuel assemblirts on top of a rack module. or lowering of: a fuel assembly by the side of abrack module in a non-sLarage location.

Cask. drop accidents are not analyzed /fot criticality consequences since the dropping of a cask into jthe fuel cto; age areas at Catawba is precluded !by

' design features and cask,';,andling procedures.

2. 2 FACILITY DESCRIPTION Each unit of the Catawba Station has an independent spent fuel storage system.

The Fuel Handling System associated with the pool is discussed in Catawba

-FSAR Section 9.1.4, and Spent Fuel Cooling System is presented in Sec- 1

' tion 9.1.3. Radiation shielding and monitoring are presented in Catawba i FSAR Sections 12.1 and 11.4, respectively. There are sufficient fuel storage {

racks ' to accommodate the number of fuel assemblies discharged f rom approxi- l mately 19 normal Catawba refueling cycles plus one complete Catawba core.

Provisions are also made to store control rods and burnable poison rods. The 1 dimensions .and location of the fuel pool are . included on Catawba FSAR l Figures 9.1.2-2 and 9.1.2" (Attachments 3 and 4, respectively). For location I of. the fuel pool in the station complex, see Catawba FSAR Figures 1.2.2-3 and 1.2.2-4. Major components, piping, valves Ed instrumentation in contact with the feel pool wate / are stainless steel. The fuel pools, transfer canals, and

5 cask pits are lined with stainless steel plate. This fuel pool liner plate is designed, fabricated, and installed as a nuclear safety related, QA Condition 1 system.

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ls The spent fuel a:;semblies are held in a vertical position by the spent fuel pool storage racks. The fuel assemblies are supported within the fuel storage !

racks by a stainless steel plate located six inches above the fuel pool floor.

Openings are provided that allow coolant water to flow through the rack and up around. the fuel assemt:ly. A lead-in assembly is provided at the top of each rack to guide fuel into its proper storage location, j l The fuel racks are designed as free standing, self-supporting, independent J There are spaces available for modules which stand on the fuel pool floor.

the potential storage of 1418 fuel assemblies per unit. Spent fuel storage i cell general arrangement is provided in Catawba FSAR Figure 9.1.2-7 (Attach- !

! ment 5). Spent fuel storage rack plans and details are provided by Catawba .

FSAR Figures 9.1.2-1 and . 9.1.2 (Attachments 6 and 7). Unit 2 facilities are 4 a mirror image of Unit 1 facilities. )i 2.3 CRITICALITY ANALYSIS METHOD J

! The analysis method which ensures the criticality safety of fuel assemblies in the spent fuel storage racks is the same as used in the Catawba New Fuel Storage Rack Analysis described previously in Section 1.3 of this summary.

2.4 CRITICALITY EVALUATION

i 2.4.1 NORMAL STORAGE This section presents the analyses which demonstrate the acceptability of j storing up to 4.0 w/o enriched Westinghouse STD or 0FA fuel in the Catawba I spent' fuel storage racks under normal conditions. A nominal case model for each fuel type. is described, and a neutron multiplication factor, K-eff, for l each nominal 'model in presented. Construction tolerances and assembly posi-tioning effects -are addressed and uncertainties which are to be applied to the nominal calculated' K-eff values are presented. The final K-eff values produced represent maximums with a 95 percent probability at a 95 percent confidence (level as required by ANSI /ANS-57.2-1983 to demonstrate criticality safety.

The following assumptions were used in the criticality evaluation:

1) Fuel assembly parameters modeled in each case are summarized in Table 2,

2) Credit is taken for the inherent neutron absorption in full length structural materials as allowed by ANSI N18.2-1973.

3) No burnable poisons, control rods, or supplemental neutron poisons are assumed to be present.

4) All assemblies are assumed to be unirradiated 4.05 w/o U-235 enriched Westinghouse STD or 0FA type. This worst case assumption allows for a specified maximum nominal enrichment of 4.0 w/o U-235 with an en-richment tolerance of 5 10.05 w/o U-235.

5) The spent fuel storage array is conservatively modeled as infinite in lateral and axial extent.

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6) Geometrical ' and material ' uncertainties due. to mechanical tolerances y" are treated by either using worst case configuration or by performing sensitivity calculations and obtaining appropriate uncertainty values. -The uncertainties considered include:

Stainless steel cell wall thickness Center-to-center spacing Cell ID i --

' Cell blowing Assembly positioning

7) Each fuel. assembly is treated as a heterogeneous, system with the -

fuel pins.. control rod guide tubes, and instrument guide tube modeled explicitly.

8) .The moderator is. pure, unborated full density water.

Table 4 provides the. nominal dimensions and tolerances of the unit cell model illustrated in Figure 3. Figure 4 illustrates the unit cell model complete with the KENO-IV box type orientations used to model the fuel assembly in the unit cell.

2.4.1.1 Westinghouse Standard Type (STD) Fuel j This section describes the nominal model 'and addresses -the uncertainties and biases for an infinite array of Westinghouse 17x17 STD fuel ~ assemblies com-

. pletely enclosed in 1/4" thick,- full length stainless. steel cells representa-tive of.the Catawba spent fuel storege racks under normal conditions.

NOMINAL CASE Referring to Figures 3 and 4, and Tables 1 and 4, the following zones are modeled explicitly in the nominal case CSAS2 analysis:

(a) The 0.25" thick cell enclosure

.(b) . The 0.3225" 0.0. fuel pellets (264 rods)

(c) The 0.374" 0.D. fuel rod clad (264 rods)

(d) The void gap between the fuel' pellets and clad

-(e) The 0.474" 0.D. guide tubes (24 tubes)

(f) The 0.48" 0.D. instrument tube The nominal case CSAS2 calculation in which 30,476 neutron histories were E followed, resulted in a K-eff of 0.90386 with a 95 percent probability /

95 percent confidence level uncertainty of +/- 0.00798.

BIASES AND UNCERTAINTIES A method bias and uncertainty has been established as discussed in Section 1.3 of this summary. A statistical bias of 1.0049 and a 95/95 uncertainty of 0.012ak is associated with the CSAS2 method used. The 95/95 uncertainty in the !

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nominal case analysis is 0.00798 A K. In addition to ' these uncertainties, there are other considerations which may effect the final K-eff value assigned to the array. These considerations are treated as either worst case in the {

nominal run or a sensitivity run is performed. to determine the AK associated j with a variable parameter (e.g., enclosure wall thickness.). i The worst-case conditions incorporated in the nominal CSA52 run are as follows: ;

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(a)- All assemblies are assumed to be unirradiated 4.05 w/o U-235 enriched.

(b) Fuel . assemblies are modeled in the center of the storage cells.

Sensitivity calculations indicate that this is a conservative assumption and may contribute as much as 0.002 AK to the nominal K-eff value depending on how may cells are modeled with the fuel assemblies off-center.

(c) A water density of 1.0 gram /ml is modeled in the nominal case run.

Any credible water density decrease from full density will result in a decrease in the calculated K-eff.

(d) Fuel storage cells are modeled uniformly at a center-to-center (C-C) spacing of 13.5". Although a construction tolerance of +/-0.125" on this parameter (not to accumulate) allows groups of four assemblies to be at a reduced C-C spacing of 13.375", sensitivity calculations indicate that the net effect of reducing the spacing on two sides of the unit cell is more than canceled out by the increased spacing resulting on the other two sides. This conservative assumption may contribute as much as 0.0005 AK, depending on how may _ cells are modeled at this reduced groups-of-four spacing. .

l The following are treated as uncorrelated tolerance uncertainties 'n the calculation of the final rack K-eff:

(a) The _ sheet metal from which the storage enclosures are fabricated has a tolerance of +/-0.05". The effect of a reduced wall thickness is determined in conjunction with the effect cell I.D. variability has on the storage array K-eff. The construction tolerance for cell I.D. i s +/-0. 0625". Sensitivity calculations indicate that a thin wall cell combined with a large cell I.D. assumption can add as much as 0.022 AK to the nominal case K-eff value.

(b) Individual cell bowing is assumed to be as much as 0.25" from verti-cal. Cell bowing is assumed to occur in conjunction with the reduced C-C spacing tolerance. This assumption yields an array of a

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groups-of-four assemblies at a C-C spacing of 13.25". Sensitivity calculations indicate that groups-of-four assemblies at this reduced C-C spacing can add as much as 0.012 AK to the nominal case K-eff value.

WORST-CASE MAXIMUM RACK K-eff The worst case maximum rack K-eff is determined by combining the nominal case results with the uncertainties and biases developed from the method benchmark 4

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h-calculations c and the sensitivity studies performed- for the Catawba- spent fuel storage' racks. The ; resulting maximum final K-eff value is accurate with a 95 percent probability at a 95 percent confidence. level.

l The following equation 'is used .to develop the final K-eff result for the Catawba spent fuel storage racks:

'K,7f. = K-nominal + 8-method +

[(ks-nominal)2 + (ks-method)2 + (ks-mechanical)2]l/2 Where:

K-noiainal = Nominal case K-eff = 0.90386 B-method =. Method bias = 0.0049 ks-nominal = 95/95 uncertainty in the nominal case K-eff value

= 0.00798 AK ks-method = 95/95 uncertainty in the method bias = 0.012 AK-ks-mechanical = 95/95 uncertainty resulting from material and construction tolerances = Q 251 aK (or, [(0.022)2 + (0.012)2] )

Substituting the appropriate values in order listed:

K-eff = 0.90386 + 0.0049 + [(0.00798)2 + (0.012)2 + (0.0251)2]1/2 K-eff = 0.9377 2.4.1.2 Westinghouse Optimized Type (0FA) Fuel This section describes the nominal. model and addresses the uncertainties and biases for an infinite array of Westinghouse 17x17 0FA fuel assemblies com-pletely enclosed in 1/4" thick, full length stainless ' steel cells representa-

-tive of the Catawba spent fuel storage racks under normal conditions.

NOMINAL CASE Referring- to Figures 3 and 4, and Tables 1 and 4, the following zones are modeled explicitly in the nominal case CSAS2 analysis:

(a) The 0.25" thick cell enclosure (b) The 0.3088" 0.D. fuel pellets (264 rods)

(c) The 0.36" 0.D. fuel rod clad (264 rods)

(

(a) The void gap between the fuel pellets and clad (e) The 0.474" 0.D. guide tubes (24 tubes) ,

(f) The 0.476" 0.D. instrument tube 1

l-l _ _ . _ - - _ - . ---_--_______-_________.__._L

The nominal case CSAS2 calculation in which 30,476 neutron histories were followed, resulted in a K-eff of 0.90676 with a 95 percent probability /

95 percent confidence level uncertainty of +/-0.00894.

BIASES AND UNCERTAINTIES A method bias and uncertainty has been established as discussed in Section

1. 3 of this s ummary. A statistical bias of +.0049 and a 95/95 uncertainty of 0.012 AK is associated with the CSAS2 method used. The 95/95 uncertainty in -

the nominal case analysis is 0.00894 AK. In addition to these uncertainties, there are other considerations which may effect the final K-eff value assigned to the array. These considerations are treated as either worst-case in the nominal run or a sensitivity run is performed to determine the AK associated with a variable parameter (e.g., enclosure wall thickness). ..

The worst-case conditions incorporated in the nominal CSAS2 run are as follows: ,

(a) All assemblies are assumed to be unirradiated 4.05 w/o U-235 enriched.

(b) A water density of 1.0 gram /mi is modeled in the nominal case run.

Any credible water density decrease from full density will result in a decrease in the calculated K-eff.

(c) Fuel storage cell C-C spacing is nominally 13.5". Fuel storage cells are modeled in groups-of-four at a reduced C-C spacing of 13.375" to account for the construction tolerance of +/-0.125" on I this parameter (non-accumulating). Sensitivity calculations indicate I

that the net effect of reducing the spacing on two sides of the unit cell and the corresponding increased spacing resulting on the other two sides may contribute as much as 0.001 AK to the nominal K-eff value, depending on how many cells are modeled at this reduced groups-of-four C-C spacing.

The following are treated as uncorrelated tolerance uncertainties in the calculation of the final rack K eff: '

(a) The sheet metal from which the storage enclosures are fabricated has a tolerance of +/-0.05". The effect of a reduced wall thickness is determined in conjunction with the effect cell I.D. variability has on the storage array K-eff. The construction tolerance for cell I.D. i s +/-0. 0625" . Sensitivity calculations indicate that a thin wall cell combined with a reduced cell I.D. assumption can add as much as 0.020 AK to the nominal case K-eff value.

(b) Individual cell bowing is assumed to be as much as 0.25" from verti-cal. This assumption yields an array of groups-of-four assemblies at a C-C spacing of 13.25". Sensitivity calculations indicate that this reduced C-C spacing can add as much as 0.011 AK to the nominal case K-eff value.

(c) Eccentric assembly positioning in the cells can result in groups-of-four assemblies being closer in C-C spacing than in the nominal case.

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. Sensitivity calculations' indicate that eccentric positioning of fuel' l assemblies in the cells can add as much as 0.020 AK to the nominal case K-eff value.

, WORST-CASE MAXIMUM RACK K-eff The worst-case maximum rack K-eff is determined by combining the nominal-case results with the uncertainties and biases developed from the method benchmark calculations and the sensitivity studies performed for the Catawba spent fuel stor age . racks. The ' resulting maximum final K-eff value is- accurate with a 95 percent prob' ability at a 95 percent confidence level.

The. following equation is used to develop the final K-eff result for the Catawba spent fuel storage racks:

Keff = K-n minal + B-method +

[(ks-nominal)2 + (ks-method)2 + (ks-mechanical)2]l/2-Where:

K-nominal = Nominal case K-eff = 0.90676 B-method = Method bias = 0.0049 ks-nominal = 95/95 uncertainty in the nominal case K-eff value =

0.00894 AK ks-method = 95/95 uncertainty in the method bias = 0,012 AK ks-mechanical = 95/95 uncertainty resulting from material and construction tolerances = 0.0303 AK (or, [(0.020)2 + (0,011)2 + (0.020)2]1 2)

Substituting the appropriate values in the order listed:

K-eff = 0.90676 + 0.0049 + [(0.00894)2 + (0.012)2 + (0.0303)2]1/2 K-eff = 0.9455 2.4.2 ACCIDENT CONDITIONS Postulated accident conditions will not result in an increase in K-eff values beyond those presented for the infinite arrays in Section 2.4.1 of this summary.

i Accidents considered include: 1) loss of spent fuel pool cooling, 2) the sliding of free standing rack modules such that peripheral cells of two racks

modules have C-C spacings below those assumed in normal storage analyses, and )

3) the dropping of fuel assemblies on top of a rack module or lowering of a j L fuel assembly by the side of a rack module in a non-storage location. Cask drop accidents are not analyzed for criticality consequences since the dropping of _ a cask into the fuel storage areas at Catawt,a is precluded by design fea-tures and cask handling procedures.

1 l

LOSS OF C00LI'NG The loss of spent fuel pool cooling with a decay heat load present in the pool l would result in lower moderator densities than assumed in the normal storage analyses. This would result in a lowering of storage array reactivity.

Optimum moderation is not a concern in the spent fuel storage pool since the moderator densities required for this effect (<0.4. grams /ml) are lower than would be credible in the pool. Since the fuel pool is supplied with redundant, safety related coolant makeup systems, the fuel will always be covered by saturated water at > or = to 1 ATM.

. SLIDING OF RACK MODULES Since the storage rack modules are free standing, it is possible two racks may slide due to seismic forces resulting in reduced C-C spacing for fuel cells stored in peripheral locations. However, the Catawba pools will contain ap-proximately 2000 ppm boron and the double contingency principle ANSI-N16.1-1975 may be' applied for this accident. This principle states that two unlikely, in-dependent, concurrent events need' not be considered to' ensure protection against a criticality accident (i.e. , rack sliding need not be assumed concur-rent with loss of pool boration).

Calculations indicate that 2000 ppm boron present .in the fuel pool coolant would reduce the nominal case K-eff values presented in Section 2.4.1 by approximately 0.23 AK. Therefore, the rack K-eff is easily held below 0.95 since any reactivity increase resulting from reduced peripheral cell C-C spacings would be much less than the negative worth of the dissolved boron.

DROPPED FUEL ASSEMBLIES

.opped fuel assembly resting on top of a storage rack will not result in an iNrease in the K-eff values presented in Section 2.4.1 since 1) the normal condition. K-eff calculations assume infinite axial fuel length, 2) the rack structure would provide some amount of water separation between the active' fuel regions of the dropped and stored assemblies, and 3) the upper end fit-tings of stored fuel assemblies are not modeled in normal condition K-eff calculations and would contribute negative reactivity to this postulated confi-guration relative to the normal case. In any case, a dropped fuel assembly resting on top of a full storage rack will not result in a K-eff value greater than the 0.95 criteria since the double contingency principle could be applied to take credit for dissolved boron for this accident condition.

Similarly, -dropping or lowering a fuel assembly in a non-storage location beside a rack module will not result in a K-eff value greater than the 0.95 criterion. As in the postulated sliding rack module accident case, the double contingency principle . could be applied to' take credit for dissolved boron in this case. The approximately 0.23 AK negative reactivity provided by 2000 ppm boron would more than compensate for the additional reactivity added by a fuel assembly being present at some reduced C-C spacing on the periphery of a rack module.

l'

B 2.5 RESULTS AND CONCLUSIONS i: . .

The calculated worst-case K-eff values for the 'two fuel types analyzed in this:

calculation are as.follows:

WESTINGHOUSE:17x17 STD FUEL: K-ef f .= 0. 9377 WESTINGHOUSE 17x17 0FA FUEL: K-eff = 0.9455-These calculated maximum K-eff. values are valid for the .specified fuel types for up to' 4.05 t w/o enrichment. These calculated values include geometrical and material uncertainties and biases at a 95 percent probability, 95 percent confidence level, as required by ANSI /ANS-57.2-1983, to demonstrate criticality safety. The uncertainties considered include:

- Stainless steel cell wall thickness

- Center-to-center spacing

- Cell ID

- Cel'1 bowing.

- Assembly' positioning As specified by ANSI N18.2-1973 and ANSI N210-1976, the maximum K-eff value in the spent fuel pool shall ,be less than 0.95, including all uncertainties, under all conditions. The analyses presented in this summary demonstrates that this criterion is met.

T - . _ _ . - _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _

4 t

+-

3.0 REFERENCES

- 1) Oak Ridge National Laboratories, " SCALE-3: A Modular Code System for Performing Standardized Computer ' Analysis for Licensing Evaluation,"

NUREG-0200 - Vols. 1-2-3-Bk. 4, Revision 3, December, 1984. l

2) M. N. Baldwin, et. al. , " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel," The Babcock and Wilcox Company, BAW-1484-4, November, 1978.

3) G. S. Hoovler, et. al., " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel," Nuclear Technology, Vol. 51, l

December, 1980.

4) S. R. Bierman, etc. al., " Critical Experiments Measuring the Reactivity Worths of Materials Commonly Encountered as Fixed Neutron Poisons," i Battelle Pacific Northwest Laboratories, BNWL-2129, October,1976. i 1

5) S. R. Bierman, etc. al., " Critical Separaticn Between Subcritical Clusters of 2.35 w/o U-235 Enriched UO2 Rods in Water With Fixed Neutron 1 Poisons," Battelle Pacific Northwest Laboratories, PNL-2438, October, 1977.

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TABLE 2 FUEL ASSEMBLY PARAMETERS (INCH /CM)

FUEL ASSEMBLY PARAMETER W 17x17 STANDARD W 17x17 OPTIMIZED FUEL RODS / ASSEMBLY 264 264 ZIRC-4 CLAD 0.D. 0.374/0.94996 0.36/0,9144 CLAD THICKNESS 0.0225/0.05715 0.0225/0.05715 FUEL PELLET 0.D. 0.3225/0.81915 0.3088/0.784352 FUEL PELLET DENSITY (% T.D.) 95.0 95.0 R0D PITCH 0.496/1.25984 0.496/1.2Fo84 NUMBER ZIRC-4 GUIDE TUBES 24 24 GUIDE TUBE 0.D. 0.474/1.20396 0.474/1.20396 GUIDE TUBE THICKNESS 0.016/0.04064 0.016/0.04064 NUMBER ZIRC-4 INSTRUMENT TUBES 1 1 INSTRUMENT TUBE 0.D. 0.48/1.2192 0.476'1.20904 INSTRUMENT TUBE THICKNESS 0.016/0.04064 0.016/0.04064 ACTIVE FUEL HEIGHT 144/365.76 144/365.76

~ "

m v.: 4...

q TABLE 3 CATAWBA FUEL VAULT CRITICALITY EVALUATION: 3 i

KENO-IV CALCULATED K-eff RESULTS FOR 4.1 W/0 ENRICHED WESTINGHOUSE 17x17 STANDARD AND OPTIMIZED FUEL i

l 1

MODERATOR DENSITY CALCULATED RESULTS (am/cc) W 17x17 STD W 17x17 0FA

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0.02 .83252 i .00405 .82555 .00420 0.03 .89945 i .00399 .89255 .00387 I

0.04 .91655 .00394 .91022 .00430 0.05' .90690 .00445 .89970 .00429 l 0.06 .88229 i .00487 .87432 .00437 0.07 .85506 i .00399 .84998 .00445 0.08 .82753 i .00469 .82235 .00404 0.09 .80791 .00433 .79571 .00436 0.10 .77312 .00461 .76538 .00433 0.20 .61111 .00421 .60772 .00400 0.30 .55903 .00405 .55537 .00388 0.40 .57934 .00487 .57970 .00494 0.50 .62112 i .00408 .62432 .00376 0.60 .68024 .00473 .68649 .00452 0.70 .71368 .00477 .73502 .00498 0.80 .79018 .00457 .78987 .00423 0.90 .83609 .00469 .82016 .00479 1.00 .87432 .00492 .87394 .00530

'l& U110L,

_ _ _ _ _ _ _ _ _ .__ .- . - - )

TABLE 4 - SPENT FUEL STORAGE CELL GE0 METRY PARAMETERS l

NCMINAL DI::ENSICN TOLEEANCE GEOMETRY CESC2:PT!ON (INCH /CM) (+/-INCH /CM)

CENTER-TO-CENTER SPACING 13.5/34.29 0.125/0.3175' CELL WALL THIC:CIESS 0.25/0.635 0.05/0.127 CELL I.D. 9.0/22.86 0.0625/0.15875 CELL-TO-CELL GAP 4.0/10.16 0.4125/1.04775~~

CELL HEIGHT 144.0/365.76 ---

i

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l ATTAC&fENT 3 ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION i A-________
r u. I ATTACHMENT 3 , ANALYSIS OF SIGNIFICANT HAZARDS CONSIDERATION As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed amendments involve significant hazards considerations, as defined by 10 CFR 50.92. Standards for determination that a proposed amendment involves to significant hazards considerations are if operation of the facility in accordance with the proposed amendment would not: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety. The proposed amendments would allow use of higher enrichment (up to 4.0 w/o U-235) fuel for future Catawba Units 1 and 2 core reloads. The current technical specification limit of 3.5 w/o is based on the criticality analyses for the new and spend fuel storage facilities at the Catawba Nuclear Station. The Attachment 2A analysis damoustrates a 4.0 w/o storage capability with a maximum enrichment tolerance of 0.05 w/o U-235 under normal storage and accident conditions. Use of higher enrichment fuel in the reactor core will be demonstrated to be acceptable via the cycle-specific reload safety evaluations which are performed prior to fuel loading and criticality accidents during refueling operations are precluded by stringent administrative procedures. The above referenced criticality evaluation demonstrates that use of the higher enrichment fuel does not involve a significant increase in the consequences of an ' accident previously evaluated or involve a significant reduction in a margin of safety (as the reactor core is capable of handling 4.0 (and higher) weight percent reload fuel similar conclusions are anticipated to be documented in the cycle-specific RSE's). Since only the fuel enrichment is being changed no accident causal mechanisms are affected or created and consequently the probability of an accident previously evaluated is unaffected and no new or different kind of 1 accidents from any accident previously evaluated can be created. The commission has provided examples of amendments #1ikely to involve no significant hazards considerations (48 FR 14870). One example of this type is (iii) "For a Nuclear Power Reactor, a change resulting from a nuclear reactor core reloading, if no fuel assemblies significantly different from those found previously acceptable to the NRC for a previous core at the facility in question are invalved. This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable." Since the proposed reload fuel assemblies differ from previously approved reload core , assemblies only in a slight increase in enrichment levels, the acceptance criteria ' for the technical specifications is met (e.g. compliance with Techn5 cal l Specification 5.6 criteria was demonstrated by the criticality evaluation with the exception of the proposed amendment itself, and the analytical methods used to . demonstrate conformance of the fuel storage facilities with the technical { specifications and regulations (as well as the methods anticipated to be used in j the cycle-specific RSE's) are those previously used/ acceptable to the'NRC, the j above cited example can be applied to this amendment. In addition, another example l l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ____a
4 i , f Q ATTACIDIENT 3 i Page 2 i
't i
of actions not likely to involve a significant hazards consideration is (vi), "A change which either may result ~ht some increase to the probability or consequences of a previously analyzed eccident or may reduce in some way a safety margin,'but'" where results of the change are clearly within all acceptable criteria with respbet I to the system or component specified in the standard review plan: for example, a! i change resulting from the application.of a small refinement of a previously used' calculational model or design method". Because the evaluations previously _, referenced for the new and spent fuel storage facilities show that all acceptable criteria are met, and similar conclusions are expected from the cycle spec 1ff6
,RSE's, this example also applies.
Sased upon the preceding analyses, Duke Power Company concludes that the proposed amendments do not involve a significant hazards consideration. 4
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ML20238F426

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1.4 CRITICALITY EVALUATION

2.4 CRITICALITY EVALUATION

3.0 REFERENCES

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ML20238F426

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1.4 CRITICALITY EVALUATION

2.4 CRITICALITY EVALUATION

3.0 REFERENCES

= H O 4 2O W f E0 O E

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