ML20196J903

Proposed Tech Specs Modifying OL NPF-1,removing 3.5% U-235 Limit on Reactor Fuel Assemblies & Increase Enrichment Limit for New Fuel Storage Racks to 4.5%.Certificate of Svc Encl

ML20196J903
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	03/01/1988
From:	PORTLAND GENERAL ELECTRIC CO.
To:
Shared Package
ML20196J865	List: ML20196J861 ML20196J886 ML20196J903
References
NUDOCS 8803150093
Download: ML20196J903 (13)

Text

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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for a maximum internal pressure of 60 psig and a temperature of 286*F.

PENETRATIONS 5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the original design provisions contained ir. Section 6.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements. f J

5.3 REACTOR CORE t

FUEL ASSEMBLIES The reactor core shall contain 193 fuel assemblies with each fuel 5.3.1 i assembly containing 264 fuel rods clad with Zircaloy-4. Each fuel rod shall l have a nominal active fuel length of 144 inches. The initial core loading

. shall have a maximum enrichment of 3.15 weight percent U-235. Reload fuel ,

1 shall be similar in physical design to the initial core loading and of low ,

enrichment.

CONTROL R0D ASSEMBLIES s 5.3.2 The reactor core shall contain 53 full length control rod assemblies.  !

The full length control rod assemblies shall contain a nominal 142 inches of r absorber material. The nominal values of absorber material shall be .

80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. Eight part length control rod assemblies originally installed in the core contained a nominal 36 inches ,

j of absorber material at their lower ends. The part length control rod  ;

assemblies have been removed and are stored in the spent fuel pool. (

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

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TROJAN-UNIT 1 5-4 Amendment No. 70, TI6  ;

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) 880315/3093 880301 PDR 4

P ADOCK 05000344 DCD [

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DESIGN FEATURES

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant '

to the applicable Surveillance Requirements,

, b. For a pressure of 2485 psig, and

c. For a temperature of 650*F, except for the pressurizer which "4 is 680*F. ,

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,900 100 cubic feet at a nominal Tavg of 584.7'F.

5.5 EMERGENCY CORE COOLING SYSTEMS ,

i 5.5.1 The emergency core cooling systems are designed and shall be 'l maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

5.6 F05_L ST0kAff(

CRITICALITY 5.6.1.1 The fuel storage racks for new fuel are designed and shall be maintained with a nominal 21-inch center-to-center distance between fuel assemblies placed in tho storage racks to ensure a Keff of 50.95 with the storage pit f~llied with unborated water, and with new fuel containing not more than 57.0 grams of U-235 per axial centimeter of active fuel  ;

assembly (4.5 wt% enrichment).

5.6.1.2 The' storage racks for spent fuel are designed and shall be main- l tained with a nominal 10.5-inch center-to-center distance between f uel assemblies placed in the storage racks to eneure a Keff of 10.95 with the storage pool filled with unborated water and with new fuel containing

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not more than 57.0 grams of U-235 per axial centimeter of active fuel l

assembly (4.5 wt% enrichment). The criticality analysis includes a con-servative allowance for uncertainties as described in Section 3.1 of PGE-1037.

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9 TROJAN-UNIT 1 5-5 Amendment No. 25, 3#, 88, ITS  ;

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NUCLEAR CRITICALITY ANALYSIS OF 4.'. Wl'IGHT PERCENT URf,NIUM '

I ENRICHMENT IN THE NEW FUEL STORWE RAC.S OF THE d TROJAN NUCLEAR PLANT i

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' Attechment l Page 2 of 10 TABLE OF CONTENTS reEtion Title Page 1 Introduction . . . . . . . . . . . . . . . . . . . . 1

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>4 2 New Fuel Storage Storage Rack and

'j Fuel Descriptions. . . . . . . . . . . . . . . . . . 1 3 Postulated Accidents . . . . . . . . . . . . . . . . 1 14 1

4 Description of Methods Used for Criticality Analysis . . . . . . . . . . . . . . . . 1 Determination of Calculational Uncertainty . 2 (f -

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, 6 Criticality Analysis Results . . . . . . . . . . . . 3 7 conclusion . . . . . . . . . . . . . . . . . . . . . 4

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8 References . . . . . , , . . . . . . . . . . . . . . 4

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1 Table 1 Fuel Data. . . . . . . . . . . . . . . . . . . . . . 5

,Yigure 1 PWR,New Fuel Rack Geometry . . . . . . . . . . . . 6

, Figure 2 PWR New Fuel Rack Cell Geometry. . . . . . . . . . . 7 Fir,ure 3 Moderator Density. . . . . . . . . . . . . . . . . . 8 i

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LCA 165 Attachment Page 3 of 10

1. INTRODUCTION A nuclear criticality safety analysis has been performed for the Trojan Nuclear Plant new fuel storage racks assuming storage of Westinghouse standard 17 x 17 fuel of 4.5 weight percent (wt%) enrichment. The design basis assumed in this analysis is that, including uncertainties, there must be 95 percent probability at a 95 percent confidence level that the effective neutron multiplication factor (k gg) e of the fuel assembly array will be 1095 flooded (l). In the unlikely event of the introduc-tion of water into the storage racks, criticality is prevented by the design of the rack which restricts the minimum ceparation between fuel assemblies to take advantage of neutron absorption in the water and stainless steel between assemblies, and limits the size of the fuel array to cause high neutron leakage at low water densities.

2. NEW FUEL STORACE RACK AND FUEL DESCRIPTIONS The new fuel storage rack for the Trojan Nuclear Plant consists of four rows of fuel cells with a nominal spacing of 21 inches center-to-center between cells. The fuel cells consist of 2 inch wide and 0.25 inch thick

stainless steel angles in the four corners around the assemblies. Tho internal clearance between these angles is 9 inches. Figures 1 and 2 j

i illustrate the rack geometry.

j The analysis assumed fresh 4.5 weight percent enriched Westinghouse standard 17 x 17 fuel with no burnable poison or control rods. The assembly is conservatively modeled with no grids and no U-236 in the fuel. The fuel assembly input data is presented in Table 1.

3. POSTULATED ACCIDENTS The absence of moderation under normal conditions guarantees sub-criticality for enrichments less than 5 weight percent (2). Any acci- l dents (such as a dropped assembly) other than the introduction of a i moderator in the storage area will meet the ek gg 01 98 limit. Because l the most limiting accident is the introduction of moderation (water). '

i this accident was considered in determining the maximum ke gg for the storage racks.

The criticality analysis included calculations of the k egg of the racks with various water densities from 1 percent to 100 percent to determine the optimum achievable moderation.

! 4. DESCRIPTION OF METHODS USED FOR CRITICALITY ANALYSIS The criticalit analysis at the various modarator densities employed the CASMO-3 code (3 for generating cross-sections which were used in the l j PDQ-7 code to model the storage racks. The CASMO-3 code is a multigroup, j l two dimensional transport theory code capable of performing calculations j j on light water reactor assemblies in typical fuel storage rack geome-  !

1 tries. The nuclear data library is based on ENDF/B-4 with some data from ENDF/B-5 and other sources. The production library, which was used to 4

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LCA 165 Attachment

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develop the curve in Figure 3, contains microscopic cross sections in r 40 energy groups. At the 100 percent and 7 percent water density condi-tions, additional calculations were made with microscopic cross sections in 70 energy groups. Seven energy groups were used in the two dimen-sional transport theory calculation. Macroscopic cross sections in four energy groups were generated by CASMO-3 for use in the PDQ-7 code.

The widely used PDQ-7 code (4) solves the neutron diffusion equations in one, two, or three dimensions. For this analysis, two dimensional cal-culations in the xy plane and one dimensional calculations in the axial direction were made. Calculations made in the xy plane modeled an infinitely long array of four rows of fuel with a 21 inch spacing between the outer rows and a 36 inch spacing between the inner rows, the concrete on the side of the new fuel storage area, and the stainless steel angle iron on the four corners of each assembly.

Neutron leakage factors in the axial direction were calculated by tuodeling the storage racks in the axial direction in a one dimensional calculation. These leakage factors were applied to the multiplication  :

factor from the calculations in the xy plane to determine a ke gg for the new fuel storage racks which neglected the slight neutron leakage out the ends of the racks.

Since use of the PDQ-7 diffusion theory code for this analysis may lead J to some inaccuracies in the results due to the approximations of diffu-sion theory, at two water densities (7 percent and 100 percent), trans-j port theory to diffusion theory adjustment factors were used to adjust the kegg., calculated for the new fuel storage rack. These adjustment factors were determined by finding the difference between the multiplica-  ;

tion factor calculated by the CASMO-3 code and the multiplication factor calculated by the PDQ-7 code for identical storage cell simulations.

This resulted in a downward adjustment of the ke gg at / percent water ,

density and an upward adjustment of the ke gg at 100 percent water density.

j The maximum possible increase in the calculated kegg at the optimum i water density due to less than nominal center-to-center spacing of the

fuel cells and asymmetric assembly positions within fuel cells was deter-mined by evaluating a worst-case condition. l S. DETERMINATION OF CALCULATIONAL UNCERTAINTY The CASMO-3 code (with the DIXY diffusion theory code) has been bench- i l marked against a wide range of experiments (5), including pin cell 'l

lattices, light water reactor lattices, and fuel storage rack geometries.  ;

The particular parameters which have been examined include:

• k egg versus temperature (290*Kolvin to 520* Kelvin).

+ kegg versus enrichment (1.3 weight percent to 3.0 weight percent).

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. k egg versus water to fuel ratio. l I

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. k egg versus leakage.

• k egg versus water gap between assemblies (0 to 6.544 contimeters).

The benchmarking studies used the 70 group microscopic cross sections described above with seven groups for the two dimensional transport theory calculations. Cross sections generatod by CASMO-3 for fuel assemblies as fuel pins were used in DIXY calculations of the experimental array. The i results of the benchmarking analysis of 37 experiments are an average kegg of 1.00039 with a standard deviation of 10.00107. This benchmarking shows no trend of kegg with any of the parameters examined.

To ensure that this benchmarking work is applicable to this analysis, which used PDQ-7 for solving the diffusion equations rather than the DIXY code, a critical experiment of a fuel storage rack geometry was modeled with CASMO-3 and PDQ-7. As expected, the result did not differ significantly from the result reported in the referenced benchmark study. Thus, the referenced benchmark study was used to comply with the requirements of ANSI N16.9-1975(6) for validation of methods for nuclear criticality safety.

Based on a 95/95 probability /confidencu level criteria, the uncertainty in ,

the bias is 2.17 x 0.00107 = 0.0023 (where 2.17 is the one-sided tolerance limit for 37 data points and a 95/95 probability / confidence level) over the range of experimental conditions benchmarked. Although the new fuel storage rack analysis lies outside the range of experimental conditions benchmarked for enrichment, size of water gap between assemblies, and (for the low water density cases) water density, the lack of any observable l trend in the benchmarking study indicates that no additional allowance for uncertainty in the bias need be made.

6. CRITICALITY ANALYSIS RESULTS l

The results of the CASMO-3/PDQ-7 calculations (7,8) are illustrated in l Figure 3. The results of these calculations indicate that optimum moderation occurs when the racks are flooded with 100 percent dense water.

The maximume k gg of the racks with a 95 percent probability and a 95 percent I confidence level is determined in the following manner:

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] kegg = knominal + akblas + akuncertainty + dkmechanical 1 in bias where knominal = kegg calculated for the optimum I moderation case kblas = the bias in the calculated methods akuncertainty = uncertainty in the bias in the in bias calculational methods e

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LCA 165 Attachment Page 6 of 10 okmechanical = maximum possible increase 1 gg due to asymmetric loading of assemba..J in the storage racks and less than nominal spacing between storage locations.

Substituting calculated values in the order listed above, the result is:

kegg = 0.9378 - 0.0004 + 0.0023 + 0.0006 = 0.9403'

7. CONCLUSIONS The conclusion of this analysis of the Trojan new fuel storage racks is that the current design is adequate to ensure that the k gg will be

<0.95 at a 95/95 probability / confidence level under normal storage condi-tions and under postulated accident conditions when loaded with standard 17 x 17 fuel with enrichments up to 4.5 weight percent. This satisfies the acceptance criteria of Reference 1. ,

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8. REFERENCES

1. NUREG-0800, "Standard Review Plan, Section 9.1.1, New Fuel Storage".

Revision 2 (1981).

2. Thomas, J. T., "Nuclear Safety Guide", NUREG/CR-0095, Revision 2, (June 1978).

3. Edenius, M., et al, "CASMO-3 Users Manual" (November 1986).

4. Cadwell, W. R., "PDQ-7 Reference Manual" (January 1967).

5. Jernberg, Per, "CASMO-3 Benchmark Against Critical Experiments" (August 1986).

6. ANSI N16.9-1975, "Validation of Calculational Methods of Nuclear Criticality Safety".

7. PGE calculation, "New Fuel Storage Rack Criticality Analysis",

TRO-87-01 (March 1987).

8. PGE calculation, "New Fuel Storage Rack Criticality Analysis Suppinment ", TRO-87-02 (Apell 1987).

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LCA 165 Attcchment Page 7 of 10 TABLE 1 FUEL DATA 1.0 Fuel Assembly Type 17 x 17 Westinghouse 2.0 Pellet outer diameter 0.3225 inches 3.0 Clad Data 3.1 Outer diametcr 0.374 inches 3.2 Thickness 0 0225 inches 3.3 Material Zlt-4 4.0 Fuel Rod Pitch 0.496 inches 5.0 U235 Enrichment 4.5 weight percent 6.0 UO2 Density 95 percent theoretical 7.0 Stack Density 10.4 grams per cubic centimeter 8.0 Water Hole Data 8.1 Thimble Material 2R-4 8.2 Thimble Dimensions Instrument Guide Tube (24)

Tube (1) Upper Lower outer diameter 0.482 inches 0.482 inches 0.429 inches inner diameter 0.450 inches 0.450 inches 0.397 inches 9.0 Active Fuel Length 144 inches l

10.0 Plenum Length 6.3 inches 1

11.0 Number of Grid Spacers 8 (not modeled)  !

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of )

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PORTLAND CENERAL ELECTRIC COMPANY, ) Docket 50-344 THE CITY OF EUGENE, OREGON, AND ) Operating License NPF-1 PACIFIC POWER & LIGHT COMPANY )

)

(TROJAN NUCLEAR PLANT) )

CERTIFICATE OF SERVICE l 1 hereby certify that copies of License Change Application 165, to the l

Operating License for the Trojan Nuclear Plant, dated March 1,1988, have been served on the following by hand delivery or by deposit in the United States mail, first class, this 1st day of March 1988:

Mr. William Dixon State of Oregon Department of Energy 625 Marion St NE Salem OR 97310 Mr. Michael J. Sykes Chairman of County Commissioners Columbia County Courthouse St. Helens OR 97051 s . .:

. R. Ankrum, Acting Manager Nuclear Regulation Branch Nuclear Safety & Regulation Subscribed and sworn to before me this 1st day of March 1988.

O-dk Notary Public of Oregon D

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