Text

New Fuel Storage Vault Criticality Analysis
	ML20211L381
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Site:	Maine Yankee
Issue date:	12/10/1986
From:	Cacciapouti R, Digiovine A, Napolitano D; Maine Yankee
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W MAINE YANKEE NEW FliEL STORAGE VAULT I CRITICALITY ANALYSIS by I D. G. Napolitano A.

and S. DiGiovine I

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I Prepared bY D. G.

.u- Mr 24 Napolitdno, P' clear Engineer Reactor Physics G up

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Prepared by '7 C N 8 /

A. S. DiGiovine, Nuclegr7 ngineer (Date)

Reactor Physics Group Approved by M / / Pfo R J. Caccf outi, Manager ('Dat'e) actor PK ics Group Approved by 2 /2//d I B. C. Slifer,'M ager (Date)

Nuclear Enginee ng Department Yankee Atomic Electric Company I Nuclear Services Division 1671 Worcester Road Framingham, Massachusetts 01701 fDR612160005 p

861208 ADOCK 05000309 1

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DISCLAIMER OF RESPONSIBILITY This c'ocument was prepared by Yankee Atomic Electric I Company (" Yankee"). The use of information contained in this document by anyone other than Yankee, or the Organization for which this document was prepared und contract, is not g authorized and, with respect to any unauthorized use, neither E Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make I

any warranty or representation as to the accuracy or completeness of the material contained in this document.

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ABSTRACT r

L This report presents the criticality analysis for the Maine Yankee new fuel storage vault. The criticality of the

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vault is studied as a function of moderator density. This included a fully loaded vault in: a flooded condition, the optimum moderation condition and a dry condition. A repetitive unit of the vault was modelled using the NITAWL-KENO Monte Carlo methodology. With fresh unshimmed L

Maine Yankee fuel of 3.5 Wt% U235, the vault is below the .95 NRC limit in a flooded condition and is below the .98 NRC

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limit in the optimum moderation condition.

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I TABLE OF CONTENTS I PAGE DISCLAIMER OF RESPONSIBILITY............................ 11 A STRAcT................................................ m LIST OF FIGURES......................................... v LIST OF TABLES.......................................... vi

1.0 INTRODUCTION

....................................... 1 2.0 NEW FUEL VAULT GEOMETRY............................ 3 3.0 METHOD OF ANALYSIS................................. 8 4.0 RESULTS............................................ 11 d

REFER'NCES......................................... 14 I APF.NDIX A - Validation of the NITAWL-KENO Methcdology in Modelling New Fuel Storage Critica.'.ity........................................ A-1 AFPENDIX B - Uncertainty Calculations.............. B-1 I

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LIST OF FIGURES Number Title Page i 2.1 Maine Yankee Fuel Building Arrangement -

Plane A-A 4 2.2 Maine Yankee Fuel Building Arrangement -

Plane B-B 5 2.3 Maine Yankee Fuel Building Arrangement -

Plane C-C 6 3.1 Maine Yankee New Fuel Rack Array 9 )

3.2 Maine Yankee New Fuel Rack - 3D Model 10 I 4.1 New Fuel Storage Rack K-effective vs.

Moderator Void 12 I

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I LIST OF TABLES Number Title Page 2.1 Mechanical Dimensions of Maine Yankee Fuel Assemblies 7 4.1 New Fuel Storage Rack K-effective vs.

Moderator Void 13 A.1 Effective Multiplication Factors A-4 A.2 Two-Group Capture and Fission Rates for Fuel A-5 A.3 Two-Group Capture Rates for Hydrogen A-6 A.4 Two-Group Capture Rates for Oxygen A-7 A.S Fuel Assembly Normalized Monte Carlo Flux A-8 I

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1.0 INTRODUCTION

The new fuel vault at Maine Yankee Power Station is a temporary storage area for fresh unirradiated fuel.

Criticality control in the new fuel vault is accomplished by the dry, wide spacing between arrays of fuel assemblies.

However, since the intrusion of water by flooding or fire fighting sprays cannot be totally precluded, the criticality of the new fuel storage area is studied as a function of moderator density with particular emphasis on conditions of low density (.40 to .05 g/cc of water).

If moderator is introduced into the new fuel vault and assemblies are present, two types of moderation occur:

1. moderation between the assembly pin cells and

2. moderation in the space between assemblies.

The first type of moderation dominates the criticality of the array in high density situations (1.0 to .40 g/cc of water).

The second type of moderation deninates in low density situations (.40 to .05 g/cc of water). This second type of mode 2ation can produce large increases in reactivity, and the moderator density at which the peak occurs is called cptimum moderation. Optimum moderator density is usually in the range of .10 to .05 g/cc of water depending on the vault array spacing.

It is important to model leakage effects under conditions of very low density (i.e. .10 to .05 g/cc of water). Otherwise, erroneously high values of K-effective are produced at optimum moderation which are not I

I realistically achievable. Neutron leakage from the array under very low density conditions suppress the criticality of the vault.

The NRC position on the limits for new fuel rack l criticality indicates that the K-effective of the racks should be below .95 when the racks are fully loaded and flooded with nonborated water and that the K-effective of the i racks should be below .98 under conditions of optimum 1

moderation (l).

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I 2.0 NEW FUEL VAULT GEOMETRY The Maine Yankee fuel building arrangement is shown in Figures 2.1, 2.2 and 2.3. The new fuel vault is adjacent to the spent fuel pool wall on one side and has l' thick concrete walls on three sides. The vault dimensions are 28' by 47' and 14' in height. Fresh assemblies can be arranged in an 8 by 20 array with a minimum spacing of 20" center-to-center. Metallic grids at the top and bottom of the vault hold the assemblins in the proper spacing.

Normally, 72 assemblies are placed in the new fuel vault for refuelling the Maine Yankee core each cycle. The mechanical dimensions of the Maine Yankee fuel assembly are given in Table 2.1. The Maine Yankee assembly is a 14 x 14 array (C.E. type) with 5 large water holes for CEA and instrument thimble insertion.

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FIGURE 2.1 MAINE YANKEE FUEL BUILDING ARRANGEMENT PLANE A M

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TABLE 2.1 I

MECHANICAL DIMENSIONS OF MAINE YANKEE FUEL ASSEMBLIES ITEM DIMENSION (in.) )

Maximum Outer Envelope

8.115 x 8.115 i Pin Pitch (14 x 14 Array) 0.580 Zr-4 Clading I OD 0.440 ID 0.378 U02 Pellet OD 0.370 Zr-4 Guide Tube OD 1.115 ID 1.055 l

. MATERIAL DENSITY (g/cc)

U02 Stack Density 10.1994

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Set by Grids 6.55 !

, 8 g

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B 3.0 METHOD OF ANALYSIS

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I The method of criticality analysis is the NITAWL-KENO methodology with the 123 group XSDRN library (2,3,4). The application of this methodology to the criticality analysis of new fuel storage under flooded and low density conditions of moderator has been validated by comparison to continuous energy Monte Carlo, see Appendix A.

The basic repetitive unit of analysis modelled with RENO is shown in Figure 3.1 in relation to the rest of the vault.

In this basic repetitive unit of the rack array, half a fuel assembly and the two major inter-assembly spacings are modelled. Reflecting boundary conditions are imposed on the sides and axial leakage is allowed from the top and bottom ends. The active fuel length and twenty centimeters of axial reflector at the top and at the bottom are modelled. The final 3-D KENO model is shown in Figure 3.2. It is assumed in the analysis that moderator is introduced uniformly throughout assembly and inter-assembly gap. The analysis is performed with a fresh unshimmed assembly at 3.5 Wt% U235.

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FIGURE 3.1 ,

I MAINE YANKEE NEW FUEL RACK ARRAY I CONCRETE BASIC UNIT OF WALL ANALYSIS I

": 50 " r b 20 " ': 52 " : !c 20 "t26 "+

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M M M M M M M M M M W m M M M M M M M FIGURE 3.2 MAINE YANKEE NEW FUEL RACK 3D MODEL

= 8.12" c M.Y. ASSEMBLY

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0000000 00

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4.06" o

INTERASSEMBLY GAP 10"

- 10" : 26" ;

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136.7" 152.45" o

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4.0 RESULTS The results of the criticality analysis are tabulated in Table 4.1 and ploted in Figure 4.2. The K-effective of the model is shown vs. % void. 0% void is a fully flooded condition with water at 68 F, 1.0 g/cc, and 100% void is the fully dry condition. In the fully flooded conditon, the K-effective of the racks is .84421 0058, and this is well below the .95 NRC limit. The K-effective of the racks

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decrease with increasing void and reaches a minimum at 60%

void of .60691 0045. The K-effective of the racks then increases rapidly and reaches a peak of .87751 0046 at 95%

void. This peak is well below the .98 NRC limit at optimum moderation. The K-effective of the racks then drops rapidly to .61031 0048 at 98% void and .17661 0016 at 100% void.

The final K-effective values in flooded and optimum

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moderator conditions with 95/95 uncertainty added are given in Appendix B.

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FIGURE 4.1 MRINE YANKEE NEW FUEL STORA9E RROKS l 3.5 W/0 U-235 ENRICHMENT K-EFFECTIVE VERSES PERDENT VOID l

I 0.9 -

0.8 -

I 0. 7 -

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0.3 -

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10 20 30 40 50 00 70 80 90 100 l PERCENT VOID I

TABLE 4.1 NEW FUEL STORAGE RACK K-EFFECTIVE VS. MODERATOR VOID WATER DENSITY L % VOID (g/cc) K-EFFECTIVE + 10 0 1.0 .8442 i .005H -

20 0.8 .7731 .0049 40 0.6 .6817 1 0047 g 60 0.4 .6069 1 0049 80 0.2 .6697 1 0045 90 0.1 .8166 i .0057 93 0.07 .8585 i .0047 -

95 0.05 .8775 i .0046 '

98 0.02 .6103 i .0048 100 0.0 .1766 i .0016 b

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REFERENCES

1. NUREG-0800, USNRC Standard Review Plan, Section 9.1.1, "New Fuel Storage", page 9.1.1.1-4.

2. ORNL/TM-3706, "AMPX: A Modular Code System for Generating Coupled Multi-Group Neutron Gamma Libraries From ENDF/B", N.

M. Greene, et. al., 1976.

3. ORNL-494F, " KENO-IV - An Improved Monte Carlo Criticality Program", L. M. Petrie and N. F. Cross, 1975.

4. DLC-16, "123-Group Neutron Cross-Section Data Generated From ENDF/B-II Data for Use in the XSDRN Discrete Ordinates Spectral Averaging Code", W. R. Cable, 1971.

5.BAW-1484-7, " Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel", M. N.

Baldwin, et. al., July 1976.

6." Safety calculations and Benchmarking of Babcock and Wilcox Designed Closed Spaced Fuel Storage Racks", W. D. Bromley and I J. S. Olzewski, Nuclear Technology, Volume 41, December 1978.

7." Evaluation of the AMPX-KENO Benchmark Calculations for High Density Spent Fuel Storage Racks", S. E. Turner and M.

I K. Gurley, NS&E, 80, 1982.

8. Personal Communication with D. R. Harris of RPI, November I 1982.

9.YAEC-1224, " Yankee Criticality Calculation Methodology", J.

A. Handschuh and K. J. Morrissey, January 1981.

10.EPRI CCM-8, "The SAM /CE Monte Carlo System for Radiation Transport and Criticality Calculations in Complex I Configurations (Revision 7.60)", H. Lichtenstein, et. al.,

1980.

11.BNL-NCS-17541, "ENDF/B Summary Documentation",

I Nuclear (ENDF-201), 3rd Edition (ENDF/B-V), R. Kinsey, Editor, July 1979.

I 12.YAEC-1343, " Criticality Analysis of Seabrook Station's New and Spent Fuel Storage Racks", D. G. Napolitano, May 1983.

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I 1 APPENDIX A validation fo the NITAWL-KENO Methodology in Modelling New Fuel Storage Criticality I

The MITAWL-KENO method with the 123-group library has been firmly validated against experiment and detailed Monte Carlo calculation for spent fuel rack geometries with I poison (5,6,7). However, NITAWL-KENO criticality calculations '

for new fuel storage with fire fighting foams or mist have not been benchmarked against experiment or detailed calculations. Because of this, it was suggested that an independent calculation be performed with continuous energy Monte Carlo and modern cross section data (8).

At YAEC, NITAWL-KENO with the 123-group library was validated for spent fuel rack analysis by comparison to benchmark criticals (9). We decided to validate this method for use in new fuel rack criticality analysis under conditions of flooding and optimum moderation by comparison to a Brookhaven National Laboratory (BNL) calculation using the SAM /CE Monte Carlo code (10) and modern ENDF/B-V cross section data (11). This calculation represents the highest level of computer benchmarking short of an actual experiment.

The BNL benchmarking was performed in support of the Seabrook Station new fuel storage vault criticality analysis (12). Two models of the Seabrook new fuel vault were set up, the 3-D and the realistic 3-D. The 3-D model contained a single Westinghouse 17 x 17 assembly surrounded by inter-assembly spacing at the minimum pitch, 21" center-to-center. The active fuel length and 20 centimeters of axial reflector at the top and bottom were modelled. The A-1 I

I realistic 3-D model, in addition, contained the concrete wall surrounding the vault and three partial assemblies cut along their centerlines. This model realistically accounted for radial as well as axial leakage.

Table A.1 shows K-effective vs. void for the 3-D model and a single calculation at 90% void, optimum moderation, for the realistic 3-D model. KENO-IV and SAM /CE agree with statistics except in the 97.5% void case which shows a 5%

discrepancy in K-effective with KENO overpredicting. The peak values at 0% and 90% void show excellent agreement. One would expect the 123-group KENO-IV calculation to be adequate except for certain features. In particular, the neutron absorption in the sharp U238 resonances may not be adequately modelled by the Nordheim calculation in the NITAWL code.

Continuous energy Monte Carlo, however, describes the resonances in detail. This detail may be important in the pin cells at the edges of an assembly in the highly voided cases.

Comparison of the two-group reaction rates are s!1own in Tables A.2 though A.S. Capture rates for U235 and U238 show some significant differences in the fast group. However, there seems to be rough compensation between KENO's I overprediction of U235 capture and underprediction of U238 capture. The same appears to be true for fast fission in U235 and U238. Thermal group reaction rates appear to be in statistical agreement between SAM /CE and KENO-IV. The two-group reaction rates for hydrogen also appear to show good agreement between SAM /CE ands KENO-IV. Some significant differences again appear for oxygen fast group reaction rates A-2 I 1

I in fuel, assembly and interassembly oxygen, but thermal group rates appear to be in statistical agreement.

Twelve-group fuel assembly fluxes are shown in Table A.S. Agreement is within statistical error for most groups and there does not seem to be any strong disagreement in spectrum as a function of void.

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I TABLE A.1 EFFECTIVE MULTIPLICATION FACTORS Void Histories Kegg Histories Model %

K,gg SAM /CE SAM /CE KENO-IV KENO-IV 3-D 0 43000 0.8546+0.0044 15000 0.8587+0.0060 3-D 60 68000 0.6411+0.0041 20000 0.6296+0.0043 3-D 90 48000 1.0997+0.0038 15000 1.0929+0.0043 3-D 95 39000 1.0795+0.0043 15000 1.0930+0.0044 3-D 97.5 I

33000 0.8090+0.0048 15000 0.8414+0.0051 Real 3-D 90 37000 0.8738+0.0043 15000 0.8711+0.0056 I

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I TABLE A.2 TWO-GROUP CAPTURE AND FISSION RATES FOR FUEL I Group % Void U Capture U U Fission U

0 1.2550-2 7.9767-2 2.7873-2 1.5890-2 1

1.4328-2 7.7519-2 2.8978-2 1.4524-2 14.2 -2.81 3.96 -8.60 60 1.2293-2 7.6600-2 2.8205-2 1.9853-2 1.4035-2 7.3277-2 2.8915-2 1.7788-2 14.2 -4.34 2.52 -10.40 90 1.8978-2 1.0751-1 4.5143-2 3.0304-2 2.2498-2 9.4562-2 4.7713-2 2.6680-2 18.5 -12.0 5.69 -11.96 95 2.8380-2 1.5781-1 6.6998-2 3.8053-2 3.4426-2 1.3736-1 7.1998-2 3.3657-2 I 21.3 -13.0 2.0146-1 7.46 8.60159-2

-11.55 4.6380-2 97.5 3.6070-2 I 4.4328-2 22.9 1.7933

-11.0 9.2854-2 7.95 3.9339-2

-15.18 I 2 0 5.2393-2 5.3879-2 4.6762-2 4.7113-2 3.0386-1 3.0402-1 0.05 2.84 0.75 3.2822-2 2.11967-1 KEY: SA./CE M

60 3.6607-2 g 'd!'-' ': "- :5 T-"

90 6.3970-2 5.7701-2 3.70534-1 6.5384-2 5.7701-2 3.6879-1 I 95 2.21 5.7749-2 0.0 5.2371-2 0.47 3.3189-1 I 5.9206-2 2.52 5.2694-2 0.62 3.3278-1 0.27 3.3715-2 3.0854-2 1.9120-1 I 97.5 3.5981-2 6.72 3.2256-2 4.54 2.0073-1 4.98 I

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L TABLE A.3 Tk'O-CROUP CAPTURE RATES FOR HYDROGEN e

L Assembly Reflector r Group % Void Hydrogen Hydrogen L

1 0 2.4801-3 3.3021-3*

2.3315-3 3.1913-3

-6.00 -3.36

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60 9.0967-4 4.8169-3*

8.2998-4 4.7591-3

-8.76 1.20 90 3.7444-4 4.6849-3 I 3.6370-4 4.5067-3

-2.86 -3.80 F 95 2.8025-4 3.5432-3 L 2.8684-4 3.4836-3 2.35 -1.68 -

97.5 1.6818-4 2.0428-3 1.6699-4 2.0580-3

-0.74 0.74 2 0 4.0877-2 4.0000-1 4.0212-2 4.0082-1

-1,63 0.21 60 1.0561-2 5.4495-1 KEY: SAM /CE 1.0019-2 5.6784-1 KENO-IV

-5.132 4.2 %*

90 4.4567-3 2.1532-1 4.4181-3 2.2817-1 0.87 5.97

[ 95 2.0224-3 2.0289-3 7.2011-2 7.7678-2 0.32 7.87

, 97.5 5.7643-4 1.6299-2 6.2298-4 1,8419-2 8.08 13.01

Radial values, the rest is radial plus axial values.

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I TABLE A. 4 TWO-GROUP CAPTURE RATES FOR OXYGEN I Group % Void Fuel (UO2 )

Oxygen Assembly (H 2O)

Oxygen Interassembly (H 2O)

Oxygen 9.5580-4~ 1.1533-3 1.1103-3 I 1 0 1.0306-3 1.2066-3 1.4388-3 7.83 4.62 29.60 I 60 1.1231-3 1.3026-3 16.0 5.8358-4 6.5223-4 11.8 1.5575-3 1.8946-3 21.6 90 1.7494-3 2.3210-4 1.3968-3 1.9286-3 2.6034-4 1.5934-3 10.2 12.2 14.1 I 95 2.0929-3 2.3174-3 1.3809-4 1.5994-4 1.0250-3 1.1058-3 10.7 15.8 7.9 97.5 2.5142-3 8.0686-5 6.6732-4 2.5243-3 8.5810-5 7.0383-4 I 0.40 6.35 1.1000-5 5.5 1.0657-4 2 0 6.3357-6 6.3868-6 1.0780-5 1.0745-4 0.61 -2.00 0.83 60 4.4439-6 2.8429-6 1.4655-4 FEY: SAM /CE 4.3516-6 2.6859-6 1.5223-4 KENO-IV 2.08 -5.52 3.88 %f I 90 7.8100-6 7.8223-6 0.16 1.1998-6 1.1844-6

-1.28 5.7913-5 6.1167-4 95 7.0719-6 5.4469-7 1.9374-5 7.1435-6 5.4389-7 2.0823-5 1.01 -0.15 7.48 I 97.5 4.1492-6 4.3728-6 1.5537-7 1.6700-7 4.3880-6 4.9377-6 5.39 7.49 12.5 I

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TABLE A.5 FUEL ASSEMBLY NORMALIZED MONTE CARLO FLUX I Croup E g(ev) E (ev) 1.80+7 H

6.06+6 0%

8.62-3 60%

9.12-3 90%

8.31-3 95% 97.5%

1 7.65-3 6.44-3 7.74-2 8.92-3 8.86-3 8.18-3 6.12-3

-10.2 -2.19 6.62 6.93 -4.97 6.06+6 8.21+5 2.67-1 3.34-1 2.84-1 2.71-1 I

2 2.60-1 2.57-1 3.00-1 2.77-1 2.55-1 2.40-1

-3.75 -10.2 -2.46 -5.90 -7.69 I 3 8.21+5 1.11+5 1.88-1 1.91-1 1.60 2.26-1 2.41-1 6.64 2.74-1 2.62-1 4.38 2.83-1 3.12-1 2.85-1 0.70 3.23-1 3.53 4 1.11+5 2.50+4 6.38-2 6.90-2 7.32-2 8.49-2 1.01-1 6.56-2 7.09-2 7.56-2 8.68-2 1.02-1 2.82 2.75 3.28 2.24 1.00 5 2.50+4 5.53+3 4.88-2 5.10-2 4.99-2 5.63-2 6.45-2 4.91-2 5.23-2 5.01-2 5.67-2 6.49-2 0.61 2.55 0.40 0.71 0.62 6 5.53+3 9.61+2 5.00-2 4.91-2 4.67-2 5.35-2 6.03-2 5.14-2 5.15-2 5.35-2 5.72-2 6.40-2 2.80 4.89 14.6 6.92 6.13 7 9.61+2 7.89+1 6.62-2 5.96-2 5.66-2 6.33-2 6.83-2 KEY: SAM /CE I 6.68-2 0.90 6.33-2 6.21 5.95-2 5.12 6.52-2 3.00 6.89-2 0.88 KENO-IV

%1 8 7.89+1 1.07+1 4.59-2 3.77-2 3.42-2 3.79-2 3.63-2 4.66-2 4.06-2 3.59-2 3.89-2 3.73-2 1.52 7.69 4.97 2.64 2.75 9 1.07+1 2.38+0 3.19-2 2.43-2 2.31-2 2.53-2 2.22-2 3.09-2 2.46-2 2.34-2 2.61-2 2.20-2

-3.13 1.23 1.29 3.61 -0.90 10 2.38+0 6.50-1 2.97-2 2.22-2 2.41-2 2.61-2 2.34-2 2.93-2 2.24-2 2.45-2 2.82-2 2.36-2 1.35 0.90 1,66 3.16 0.85 11 6.50-1 3.50-1 1.50-2 1.06-2 1.16-2 1.21-2 9.14-3 1.41-2 1.07-2 1.14-2 1.23-2 9.27-3 6.00 0.94 -1.72 1.65 1.42 12 3.50-1 1.00-5 1.85-1 1.08-2 1.15-1 7.92-2 3.68-2 1.90-1 1.13-2 1.19-1 8.00-2 3.85-2 2.70 4.63 3.48 1.01 4.62 A-8

I APPENDIX B Uncertainty Calculations The uncertainty calculation for the KENO results consists of combining the KENO uncertainty and the methodology uncertainty such that there is a 95% confidence that the actual K-effective is below the calculated average plus the uncertainty. This calculated average plus combined uncertainty should be below the .95 NRC limit for the flooded condition and below .98 for the optimum moderation condition.

The 95/95 one sided uncertainty for the NITAWL-KENO-IV methodology has been calculated to be 0.008 from calculations of benchmark critical experiments (9). Since a large number of histories, 20,000, are used the the KENO analysis, a 95/95 one sided tolerance factor of 1.65 is multiplied by the KENO calculated standard deviation to get the KENO uncertainty.

For the fully flooded condition, KENO calculated:

K-eff = .8442 1 0058 Combining the KENO uncertainty and the methodology uncertainty gives:

= [(.008)2 + (1.65*.0058)2 35

= .0125 Thus, the K-effective with uncertainties for the vault fully loaded with 3.5 Wt% U235 fuel and in a flooded condition is

.8567 which is well below the .95 NRC limit.

For the optimum moderation condition, KENO calculated:

K-eff = .8775 i .0046 Combining the uncertainties gives:

= [(.008)2 + (1.65*.0046)2]h B-1 I _- _ .

= .0110 l

Thus, the K-effective of the vault fully loaded with 3.5 Wt%

U235 fuel and in the optimum moderation condition is .8885 which is well below the .98 limit.

4 l

4 i

I 1 .

i l

1 4

I iI i

l I

l I B-2 i

J

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ML20211L381

Text

1.0 INTRODUCTION

1.0 INTRODUCTION

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