NUREG/CR-6716

ORNL/TM-2000/385

Recommendations on

Fuel Parameters for

Standard Technical

Specifications for

Spent Fuel Storage

Casks

S. M. Bowman, I. C. Gauld, J. C. Wagner

Oak Ridge National Laboratory

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Washington, DC 20555-0001
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NUREG/CR-6716

ORNL/TM-2000/385

Recommendations on Fuel Parameters

for Standard Technical

Specifications for

Spent Fuel

Storage Casks

Manuscript Completed: February 2001

Date Published: March 2001 

Prepared by

S. M. Bowman, I. C. Gauld, J. C. Wagner, ORNL

Oak Ridge National Laboratory

Managed by UT-Battelle, LLC

Oak Ridge, TN 37831-6370

C. J. Withee, NRC Project Manager

Prepared for

Spent Fuel Project Office

Office of Nuclear Material Safety and Safeguards

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

NRC Job Code B0009

_________________________________________________________________

For sale by the U.S. Government Printing Office

Superintendent of Documents, Mail Stop: SSOP, Washington, DC 20402-9328

ISBN 0-16-050570-4
4
iii

ABSTRACT

The U.S. Nuclear Regulatory Commission (NRC) is currently reviewing the technical specifications for spent

fuel storage casks in an effort to develop standard technical specifications (STS) that define the allowable spent

nuclear fuel (SNF) contents. One of the objectives of the review is to minimize the level of detail in the STS

that define the acceptable fuel types. To support this initiative, this study has been performed to identify

potential fuel specification parameters needed for criticality safety and radiation shielding analysis and rank

their importance relative to a potential compromise of the margin of safety.
iv
v

CONTENTS

ABSTRACT ...........................................................................................................................................................iii

LIST OF FIGURES...............................................................................................................................................vii

LIST OF TABLES .................................................................................................................................................ix

ACKNOWLEDGMENTS......................................................................................................................................xi

1 INTRODUCTION ............................................................................................................................................1

1.1 Background ...............................................................................................................................................1

2 REVIEW OF FUEL SPECIFICATIONS FOR CRITICALITY SAFETY ......................................................3

2.1 Review Basis.............................................................................................................................................3

2.2 Technical Specification Parameters ..........................................................................................................3

2.3 Analysis Methods......................................................................................................................................3

2.4 Results.......................................................................................................................................................4

2.4.1 Principal Fuel Specification Parameters........................................................................................4

2.4.2 Secondary Fuel Specification Parameters .....................................................................................5

2.4.3 Less Significant Fuel Specification Parameters ............................................................................5

2.4.4 Examination of Anomalous or Unexpected Results......................................................................9

2.4.4.1 Fuel Pellet OD ...............................................................................................................................9

2.4.4.2 Fuel Stack Density.........................................................................................................................9

2.5 Recommendations...................................................................................................................................16

3 REVIEW OF FUEL SPECIFICATIONS FOR SHIELDING.........................................................................21

3.1 Review Basis...........................................................................................................................................21

3.2 Technical Specification Parameters ........................................................................................................21

3.3 Analysis Methods and Models ................................................................................................................23

3.4 Results.....................................................................................................................................................23

3.4.1 Principal Fuel Specification Parameters......................................................................................23

3.4.1.1 Burnup .........................................................................................................................................23

3.4.1.2 Enrichment...................................................................................................................................26

3.4.1.3 Cooling Time ...............................................................................................................................26

3.4.2 Secondary Fuel Specification Parameters ...................................................................................26

3.4.2.1 Fuel Assembly Type ....................................................................................................................26

3.4.2.2 Burnable Poison Assemblies........................................................................................................32

3.4.2.3 Uranium Mass..............................................................................................................................34

3.4.2.4 Specific Power.............................................................................................................................36

3.4.2.5 Moderator Density .......................................................................................................................38

3.4.2.6 Fuel Cladding Material ................................................................................................................38

3.4.3 Combined Burnup, Enrichment, and Cooling Time Effects .......................................................38

3.5 Recommendations...................................................................................................................................42

4 SUMMARY....................................................................................................................................................45

5 REFERENCES ...............................................................................................................................................47
vi
vii

LIST OF FIGURES

Figure Page

1 Normalized k-inf versus pellet OD for an infinite pin cell lattice..................................................................10

2 Normalized k-eff versus pellet OD for an infinite radial array of fuel assemblies in cask basket .................11

3 Normalized k-eff versus EALF for KENO V.a pellet OD calculations.........................................................12

4 Normalized 1-D k-inf versus fuel density for an infinite pin cell lattice at various enrichments...................13

5 Normalized 3-D k-eff versus fuel density results at various enrichments......................................................14

6 Comparison of 1-D and 3-D infinite pin cell results for normalized k-inf versus fuel density......................15

7 Normalized 3-D k-inf versus fuel density for the 3 geometric models ..........................................................18

8 Normalized 3-D k-inf versus fuel density for the 3 geometric models at 2 wt %..........................................19

9 Variation of the neutron and gamma dose rates as a function of burnup, for cooling times of 5, 20,

and 100 years. Calculated for a concrete storage cask with 3 wt % fuel.......................................................24

10 Variation of the total dose rates as a function of burnup, for cooling times of 5, 20, and 100 years

for the concrete (A) and steel (B) cask designs. Calculated assuming 3 wt % fuel. .....................................25

11 Variation of the neutron (A) and gamma (B) dose rate as a function of enrichment, for burnups of 20,

40, and 60 GWd/t for the concrete cask design, assuming 3 wt % enrichment and 5 year cooling...............27

12 Variation of the neutron (A) and gamma (B) dose rate as a function of enrichment, for burnups of 20,

40, and 60 GWd/t for the concrete cask design, assuming 3 wt % enrichment and 100 years cooling. ........28

13 Variation of the total dose rate as a function of enrichment, for burnup of 20, 40, and 60 GWd/t

for concrete and steel cask designs (5 year and 100-year cooling times shown) ...........................................29

14 Variation of the gamma (A) and neutron (B) dose rates as a function of cooling time for an enrichment

of 3.5 wt % and burnups of 20, 40, and 60 GWd/t, for the concrete cask design. .........................................30

15 Ratio of the neutron, gamma, and total dose rate for assemblies with and without burnable poisons...........35

16 Distribution of uranium mass for currently discharged PWR (A) and BWR (B) fuel assemblies.

The distributions reflect the actual inventory of spent fuel assemblies through 1998 from commercial

reactor operation.............................................................................................................................................37

17 Enrichment-burnup dose rate contour plots for concrete storage cask, 5-y cooling ......................................40

18 Burnup-cooling time dose rate contour plots for concrete storage cask ........................................................41
viii
ix

LIST OF TABLES

Table Page

1 Design specifications for selected fuel assembly types....................................................................................4

2 Westinghouse 17 × 17 fuel assembly parameter study results.........................................................................6

3 ABB-CE 16 × 16 fuel assembly parameter study results.................................................................................7

4 GE 8 × 8 fuel assembly parameter study results ..............................................................................................8

5 Fuel technical specification parameters and common restrictions.................................................................22

6 Design specifications and operational conditions for selected fuel assembly types (no burnable poisons)...31

7 Design specifications for integral burnable poison rod fuel assemblies ........................................................33

8 Summary of fuel specification study finding .................................................................................................43
x
xi

ACKNOWLEDGMENTS

This work was performed under contract with the Office of Nuclear Material Safety and Safeguards,

U.S. Nuclear Regulatory Commission (NRC). The authors thank Cecil V. Parks of Oak Ridge National

Laboratory (ORNL) for his guidance and insight and Carl J. Withee of the NRC Spent Fuel Project for his

review and comments. The authors also thank Willena Carter for her careful preparation of the final report and

Kay Lichtenwalter for producing the plots.
xii
Section 1 Introduction

1

1 INTRODUCTION

1.1 Background

Under the current licensing procedures for spent nuclear fuel (SNF) dry storage casks, vendors must identify all

fuel assembly types that may potentially be stored in their casks. The acceptable cask contents are limited to the

fuel specifications (i.e., dimensions or ranges of dimensions) as detailed in the Technical Specifications (TS).

Each time a candidate fuel assembly type or specification falls outside the range identified in the TS, a Part 72

license amendment request must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for review

and approval. This practice results in frequent license amendments that create an unnecessary burden on the

NRC and industry.

Recently, Section 72.48 in Title 10 of the Code of Federal Regulations (CFR) was revised, effective April 5,

2001, to better define the changes in the SNF storage cask design or procedures that can be made without a

license amendment request. In implementing this rule change, some control of the cask contents will be shifted

from the TS to the Final Safety Analysis Report (FSAR). The objective is to replace the current detailed TS

with more general Standard Technical Specifications (STS) that concentrate control on those fuel parameters

that are most important to maintaining safety. The remaining fuel parameters are of lesser importance and would

be handled under the Section 72.48 process, which allows the licensees to change those parameters by

performing additional safety analyses to update the FSAR. The licensee would notify the NRC of the FSAR

updates but no review or approval by NRC would be required.

As a result of these changes, the NRC is working to develop STS that minimize the level of detail that defines

the acceptable fuel types that can be loaded in a cask. To support this initiative, a study has been performed to

identify and rank potential fuel specification parameters needed for criticality safety and radiation shielding and

rank their importance relative to a potential compromise of the margin of safety. Minimizing the level of detail

in the STS could allow new fuel types and designs that are not significantly different than the design basis

assemblies to be added using a Section 72.48 change to the FSAR.

The following goals have been identified as being important to the development of STS for dry storage casks.

1. The STS should address all fuel design parameters that are important to criticality safety and radiation

shielding.

2. The STS, in combination with the Section 72.48 change process, must ensure that the relevant limits for

subcriticality and radiation dose are not exceeded.

3. The STS should enable allowance for additional fuel assemblies through the Section 72.48 change

process, and thus reduce or eliminate the need for license amendment requests for expanding fuel

specifications that do not have a significant impact on the relevant limits for subcriticality and radiation

dose.

Section 2 of this report documents a study of the importance of individual fuel design parameters to maintaining

subcriticality. Section 3 documents a similar parameter study relative to radiation shielding safety limits.

Finally, Section 4 presents a summary of recommendations regarding which parameters should be considered of

primary importance to criticality safety and radiation shielding in the STS.
Introduction Section 1

2
Section 2 Review of Fuel Specifications

3

2 REVIEW OF FUEL SPECIFICATIONS FOR

CRITICALITY SAFETY

2.1 Review Basis

A large number of fuel assembly design parameters can potentially influence the maximum neutron

multiplication factor (keff ) of a spent fuel storage cask. The importance of the parameters can be assessed by

determining their impact on the upper subcritical limit (USL), that calculated keff value above which loading of

the cask would not be allowed. The objective of this section is to determine which fuel assembly design

parameters would be included in the STS and which should be specified in the FSAR, based on the importance

to criticality safety. Note that no burnup credit is assumed in this study. The USL typically includes 5%

margin1

 to assure subcriticality. This margin may potentially be encroached during changes to the FSAR not

reviewed by the NRC. Thus, this margin is the basis of examination for compromises to safety.

2.2 Technical Specification Parameters

The following candidate fuel specification parameters for criticality safety analyses were reviewed in the

parametric study to determine the sensitivity of keff to each.

$ Enrichment

$ Fuel rod pitch

$ Fuel pellet outer diameter (OD)

$ Cladding thickness

$ Cladding OD

$ Guide/instrument tube thickness

$ Active fuel length

$ Fuel stack density

2.3 Analysis Methods

All calculations were performed using the SCALE code system.2 One-dimensional (1-D) radial

CSAS1X/XSDRNPM infinite pin cell lattices and three-dimensional (3-D) CSAS25/KENO V.a infinite fuel

assembly array lattices in poisoned cask basket geometries were modeled for Westinghouse pressurized water

reactor (PWR) 17 × 17, ABB-Combustion Engineering (ABB-CE) 16 × 16 PWR, and General Electric (GE)

boiling water reactor (BWR) 8 × 8 fuel assemblies. The PWR basket model is an infinite radial array of storage

cells based on the Holtec MPC-24 basket,3

 and the BWR basket model is based on the Holtec MPC-68 basket.3

Both baskets use boral absorber plates for criticality control. The nominal design specifications for each of the

fuel assembly types are listed in Table 1.
Review of Fuel Specifications for Criticality Safety Section 2

4

Table 1 Design specifications for selected fuel assembly types

Design parameter Westinghouse

17 × 17 OFA

ABB CE

16 × 16

GE BWR

8 × 8

Number of fuel rods 264 236 62

Number of instrument tubes 1 1 0

Number of guide tubes 24 4 2

Number of burnable poison rods 0 0 0

Fuel type UO2 UO2 UO2

Nominal enrichment (wt % 235U) 4.0 4.6 .0

Fuel density (g/cm3

) 10.412 10.412 10.5216

Clad type Zircaloy Zircaloy Zircaloy

Fuel rod outer diameter (cm) 0.7975 0.8262 1.0414

Clad inner diameter (cm) 0.8225 0.8432 1.0642

Clad outer diameter (cm) 0.9385 0.9702 1.1176

Fuel rod pitch (cm) 1.2675 1.2852 1.6256

Guide tube data

Inner radius (cm) 0.56130 1.143 0.6570

Outer radius (cm) 0.59686 1.2446 0.7080

Guide tube material Zircaloy Zircaloy Zircaloy

2.4 Results

2.4.1 Principal Fuel Specification Parameters

The change, in keff over the range of each parameter for different fuel assembly types, has been used to

determine which design parameters have the greatest impact on criticality safety. This approach assumes that

the fuel assembly type (i.e., assemblies with the same array size, number and cladding material of fuel rods, and

number and material of guide tubes) will be specified in the STS, because sensitivity of keff to the design

parameters is dependent on the fuel assembly design. The cladding and guide tube materials are considered

important to the definition of the fuel assembly type because of the potential impact on reactivity and neutron

spectrum in switching from one type of material to another (e.g., from stainless steel to Zircaloy). Loading a

fuel assembly type not specified in the STS would require a licensing amendment.
Section 2 Review of Fuel Specifications for Criticality Safety

5

Where possible, the parameter ranges for the Westinghouse and GE assemblies were based on actual fuel

dimensions.3

 For enrichment, fuel stack density, and fuel length, the parameter ranges were defined using an

arbitrary percentage change from the nominal value. Because ranges in actual fuel dimensions were not

available for the ABB-CE fuel, the percentage changes from nominal based on the Westinghouse fuel were

applied. Note the Westinghouse 17  17 study covers the full parameter range for both Standard and Optimized

Fuel Assembly (OFA) fuel assembly designs. Tables 2 through 4 show the calculated results and ∆k values for

the Westinghouse 17  17, ABB-CE 16  16, and GE 8  8 fuel assemblies, respectively. It is noted in the final

column of each table whether the minimum or maximum value of each parameter produces the limiting keff. The

most significant parameter, based on the change in keff over the entire range of parameter variation, is

enrichment (approximately 5% ∆k/wt % 235U). Due to the large sensitivity to enrichment, it is recommended

that maximum enrichment be required in the STS.

2.4.2 Secondary Fuel Specification Parameters

Other parameters in Tables 2 through 4 that cause changes in the calculated keff value of > 1% over the entire

range of parameter variation have been flagged as significant. These parameters include fuel rod pitch, fuel

pellet OD, and cladding thickness. Note that the classification of significant parameters would not change

notably if the criterion were reduced from 1% to 0.5%. It is recommended that changes in these parameters

would require an update to the FSAR via the Section 72.48 process. Based on the calculated results listed in

Tables 2 through 4, the maximum anticipated change in keff due to changes in all these parameters should be no

more than 1.5 to 2%.

Note that the assumption of no clad is worth approximately 5% ∆k. The assumption of no clad carries a very

high degree of conservatism. It is unlikely to be used by applicants, but it is presented for informational

purposes. However, as noted above, within the anticipated range of clad thickness, the worth is ≤ 1.2%.

2.4.3 Less Significant Fuel Specification Parameters

Cladding OD (assuming constant clad thickness), guide or instrument tube/water rod thickness, active fuel

length, and fuel stack density are shown to have relatively minor impact on keff (approximately 0.5% or less).

The worth of the guide and instrument tubes for PWRs and water rods and channels for BWRs is very small,

approximately 0.5%. For this reason, an applicant may find it desirable to eliminate them from consideration in

the FSAR. Because replacing them with water in the criticality safety model is conservative, their presence may

be neglected entirely in the criticality safety analysis. The additional conservatism associated with these model

simplifications has a minimal impact on the USL. If the licensee does include them in the FSAR analysis, these

parameters must be included in the Section 72.48 process.
Review of Fuel Specifications for Criticality Safety Section 2

6

Table 2 Westinghouse 17 

× 17 fuel assembly parameter study results

Parameter Minimum Nominal Maximum

1-D k-eff

min.

parameter

1-D k-eff

max.

parameter

3-D k-eff

min.

parameter1

3-D k-eff

max.

parameter1

% 

∆

k

1-D Significant

% 

∆

k

3-D 

Enrichment 3.500 4.000 4.500 1.42800 1.47885 0.9358 0.9862 5.09 Yes 5.04 Pitch 1.2570 1.2675 1.2780 1.45315 1.45890 0.9581 0.9666 0.58 0.85 Pellet OD 0.7680 0.7975 0.8270 1.46370 1.44587 0.9674 0.9603 -1.78 Yes -0.71 Clad thickness 0.0520 0.0580 0.0640 1.45884 1.45343 0.9694 0.9574 -0.54 -1.20 No clad 0.0000 0.0580 0.0640 1.48094 1.45343 1.0086 0.9574 -2.75 Yes -5.12 Clad OD3 0.9140 0.9385 0.9630 1.45740 1.45498 0.9644 0.9629 -0.24 -0.15 GT thickness 0.0320 0.0356 0.0391 0.9622 0.9620 N/A -0.02 No guide tube 0.0000 0.0356 0.0391 0.9681 0.9620 N/A -0.61 Fuel length 351.00 366.00 381.00 0.9632 0.9626 N/A -0.06 Fuel stackdensity (% TD)

93.00 95.00 97.00 1.45718 1.45506 0.9606 0.9661 -0.21 0.55 

Notes:1. The statistical uncertainties associated with the 3-D k-eff calculations are less than 0.1%.2. All dimensions are in cm.3. A fixed nominal clad thickness was maintained in the clad OD calculations.4. Limiting value based on 3-D results (see Section 2.4.4.2 for discussion).

6
Section 2 Review of Fuel Specifications for Criticality Safety

7

Table 3 ABB-CE 16 

× 16 fuel assembly parameter study results

Parameter Minimum Nominal Maximum

1-D k-eff

min.

parameter

1-D k-eff

max.

parameter

3-D k-eff

min.

parameter1

3-D k-eff

max.

parameter1

% 

∆

k

1-D Significant

% 

∆

k

3-D Si

Enrichment 4.025 4.600 5.175 1.44901 1.49365 0.9300 0.9738 4.46 Yes 4.38 Pitch 1.2746 1.2852 1.2958 1.46985 1.47738 0.9499 0.9606 0.75 1.07 Pellet OD 0.7956 0.8262 0.8432 1.48449 1.45968 0.9589 0.9467 -1.77 Yes -1.22 Clad thickness 0.0569 0.0635 0.0701 1.47718 1.47030 0.9607 0.9492 -0.69 -1.15 No clad 0.0000 0.0635 0.0701 1.50518 1.47030 1.0056 0.9492 -3.49 Yes -5.64 Clad OD3 0.9449 0.9702 0.9955 1.47735 1.47227 0.9590 0.9532 -0.51 -0.58 GT thickness 0.0914 0.1016 0.1118 0.9548 0.9535 N/A -0.13 No guide tube 0.0000 0.1016 0.1118 0.9591 0.9535 N/A -0.56 Fuel length 351.00 366.00 381.00 0.9531 0.9539 N/A 0.08 Fuel stackdensity (%TD)

93.00 95.00 97.00 1.47525 1.47221 0.9520 0.9566 -0.30 0.46 

Notes:1. The statistical uncertainties associated with the 3-D k-eff calculations are less than 0.1%.2. All dimensions are in cm.3. A fixed nominal clad thickness was maintained in the clad OD calculations.4. Limiting value based on 3-D results (see Section 2.4.4.2 for discussion).

7
Review of Fuel Specifications for Criticality Safety Section 2

8

Table 4 GE 8 

× 8 fuel assembly parameter study results

Parameter Minimum Nominal Maximum

1-D k-eff

min.

parameter

1-D k-eff

max.

parameter

3-D k-eff

min.

parameter1

3-D k-eff

max.

parameter1

% 

∆

k

1-D Significant

% 

∆

k

3-D Si

Enrichment 3.675 4.200 4.725 1.43951 1.48767 0.9213 0.9787 4.82 Yes 5.74 Pitch 1.6154 1.6256 1.6282 1.46365 1.46677 0.9470 0.9531 0.31 0.61 Pellet OD 1.0288 1.0414 1.0566 1.46920 1.46213 0.9519 0.9531 -0.71 0.12 Clad thickness 0.0729 0.0813 0.0897 1.46919 1.46313 0.9561 0.9480 -0.61 -0.81 No Clad 0.0000 0.0813 0.0897 1.49408 1.46313 0.9932 0.9480 -3.10 Yes -4.52 Clad OD3 1.2046 1.2268 1.2522 1.46729 1.46495 0.9543 0.9505 -0.23 -0.38 Water rodthickness 0.0635 0.0762 0.1016 0.9523 0.9521 0 -0.02 No water rod 0.0000 0.0762 0.1016 0.9553 0.9521 0 -0.32 Fuel stackdensity (%TD)

94.00 96.00 98.00 1.46738 1.46491 0.9499 0.9555 -0.25 0.56 

Notes:1. The statistical uncertainties associated with the 3-D k-eff calculations are less than 0.1%.2. All dimensions are in cm.3. A fixed nominal clad thickness was maintained in the clad OD calculations5. Limiting value based on 3-D results (see Section 2.4.4.2 for discussion).

8
Section 2 Review of Fuel Specifications for Criticality Safety

9

2.4.4 Examination of Anomalous or Unexpected Results

For fuel parameters where 1-D calculations are possible, the 1-D and 3-D results show generally good

agreement. However, this observation is not true for the fuel stack density. In addition, the fuel pellet OD

results contradicted the expected results. The variation of keff versus these parameters was studied in more

detail, as presented below. Because the fuel pellet OD is a more significant parameter, it is examined first.

2.4.4.1 Fuel Pellet OD

The minimum pellet OD was identified as the limiting value for criticality based on the results in Tables 2 and 3

and assuming nominal values for all other parameters, including enrichment. Because the maximum pellet OD

was expected to be the limiting value for keff, additional calculations were performed with both

CSAS1X/XSDRNPM and CSAS25/KENO V.a for the Westinghouse 17  17 fuel assembly design at various

pellet ODs and uranium enrichments to better understand the impact of the pellet OD on keff. The results of

these calculations are plotted as normalized keff versus percent of nominal pellet OD in Figures 1 and 2. The

maximum pellet OD is limited by the nominal inside diameter (ID) of the clad. The calculated keff values have

been normalized to the calculated keff value at the nominal pellet OD for the same uranium enrichment. This

normalization allows the trends in results to be easily compared in a single figure. The XSDRNPM 1-D infinite

pin cell lattice results and the KENO V.a fuel assembly cask basket results are fairly consistent. They show that

keff does increase as previously expected with increasing fuel pellet OD for low enrichments (e.g., 2 wt %), so

the optimum keff occurs at the maximum pellet OD. As the enrichment increases, the optimum point shifts

toward smaller pellet ODs. The optimum keff value for the 3.5 wt % results occurs in the middle of the range,

and at 5 wt %, the optimum occurs near the low end of the range.

The reason for this phenomenon may be understood by examination of Figure 3, which plots the normalized keff

values as a function of the spectral index, the energy of the average lethargy causing fission (EALF). The

results for all three enrichments have an optimum keff for an EALF of approximately 0.16 eV. As the size of the

pellet increases, the spectrum hardens. If the EALF is less than the optimum, increasing the pellet OD results in

an increased keff as shown in Figure 3. However, if the EALF is equal to or greater than the optimum,

increasing the OD results in a decreased keff value due to the associated spectral hardening. For cask analysis

performed at enrichments between 3.5 and 5 wt %, it appears that the optimum pellet OD will occur somewhere

between the nominal and the minimum OD; tending toward minimum with increasing enrichment.

2.4.4.2 Fuel Stack Density

The 1-D deterministic results for fuel stack density are inconsistent with the 3-D Monte Carlo results. Although

the change in keff over the range of fuel stack density is small, in both the 1-D and 3-D results, it seemed prudent

to understand the apparent discrepancy in results and to identify the optimum fuel density. As was done for the

pellet OD cases, additional calculations were performed with both XSDRNPM and KENO V.a for the

Westinghouse 17 × 17 fuel at various fuel stack densities and uranium enrichments to better understand the

impact of the fuel density on keff. The results of the 1-D pin cell calculations are presented in Figure 4. For low

enrichment (e.g., 2 wt % 235U) the reactivity increases with increasing density. However, as the enrichment

increases, the reactivity of the fuel decreases with increasing fuel density. This behavior was not observed with

the 3-D basket cell calculations, as shown in Figure 5. To compare the 1-D and 3-D methods directly, infinite

pin cell calculations were performed at the minimum and maximum densities with KENO V.a for the minimum

and maximum enrichments studied (2 wt % and 5 wt %). The 1-D and 3-D results are compared in Figure 6.
Review of Fuel Specifications for Criticality Safety Section 2

10

0.9800

0.9850

0.9900

0.9950

1.0000

1.0050

1.0100

88% 90% 92% 94% 96% 98% 100% 102% 104%

Effect of Pellet OD on k-inf

CSAS1X calculations

E=2 wt %

E=3 wt %

E=4 wt %

Normalized k-inf

Percent of Nominal OD

nominal OD, 0.78435 cm

Figure 1 Normalized k-inf versus pellet OD for an infinite pin cell lattice
Section 2 Review of Fuel Specifications for Criticality Safety

11

0.9800

0.9850

0.9900

0.9950

1.0000

1.0050

1.0100

88% 90% 92% 94% 96% 98% 100% 102% 104%

Effect Pellet OD on k-eff

KENO V.a calculations

E=2 wt %

E=3.5 wt %

E=5 wt %

Normalized k-eff

Percent of Nominal Pellet OD

nominal OD, 0.78435 cm

Figure 2 Normalized k-eff versus pellet OD for an infinite radial array of fuel assemblies in cask basket
Review of Fuel Specifications for Criticality Safety Section 2

12

0.9800

0.9850

0.9900

0.9950

1.0000

1.0050

1.0100

0.1 0.15 0.2 0.25 0.3

k-eff vs EALF for Variations in Pellet OD

KENO V.a calculations

k-eff=2 wt %

k-eff=3.5 wt %

k-eff=5 wt %

Normalized k-eff

EALF

Figure 3 Normalized k-eff versus EALF for KENO V.a pellet OD calculations
Section 2 Review of Fuel Specifications for Criticality Safety

13

0.9980

0.9985

0.9990

0.9995

1.0000

1.0005

1.0010

1.0015

1.0020

90% 92% 94% 96% 98% 100% 102%

Effect of Fuel Density on k-inf

CSAS1X calculations

E=2 wt %

E=3 wt %

E=4 wt %

E=5 wt %

Normalized k-inf

Percent of Theoretical Density

nom inal density, 96% TD

Figure 4 Normalized 1-D k-inf versus fuel density for an infinite pin cell lattice at various enrichments
Review of Fuel Specifications for Criticality Safety Section 2

14

-0.0080

-0.0060

-0.0040

-0.0020

0.0000

0.0020

0.0040

0.0060

0.0080

90% 92% 94% 96% 98% 100% 102%

Effect of Fuel Density on k-inf

KENO V.a calculations

E=2 wt % (Basket)

E=3.5 wt % (Basket)

E=5 wt % (Basket)

delta-k

Percent of Theoretical Density

nominal density, 96% TD (10.5216 g/cc)

Figure 5 Normalized 3-D k-eff versus fuel density results at various enrichments
Section 2 Review of Fuel Specifications for Criticality Safety

15

0.9980

0.9990

1.0000

1.0010

1.0020

90% 92% 94% 96% 98% 100% 102%

Effect of Fuel Density on k-inf

1-D vs 3-D Infinite Pin Cell Results

E=2 wt % (1-D)

E=5 wt % (1-D)

E=2 wt % (3-D)

E=5 wt % (3-D)

Normalized k-inf

Percent of Theoretical Density

nom inal density, 96% TD

Figure 6 Comparison of 1-D and 3-D infinite pin cell results for normalized k-inf versus fuel density
Review of Fuel Specifications for Criticality Safety Section 2

16

Assuming that the 3-D results vary linearly, the 3-D results show very good agreement with the 1-D results.

Thus, Figure 6 indicates that the differences between 1-D and 3-D results shown in Tables 2ñ4 are due to

spectral effects caused by geometrical differences between the infinite pin cell and the infinite cask basket

models, not differences between XSDRNPM and KENO V.a.

This conclusion was confirmed by performing KENO V.a calculations at 2 and 5 wt % for three different

geometry models based on the Westinghouse 17 × 17 fuel assembly design: (1) infinite pin cell lattice,

(2) infinite array of 17 × 17 fuel assemblies with a water gap between assemblies, and (3) infinite array of

17 × 17 fuel assemblies in poisoned cask baskets. The results are shown in Figure 7. The calculations were

performed at 0.5% increments in fuel density for the infinite cask basket model. Since the results were

approximately linear, calculations were only performed at the minimum and maximum densities for the infinite

pin cell and assembly models. The slope of the line for kinf versus fuel density decreases with increasing

enrichment. This behavior is consistent with that observed in Figures 4ñ6. Likewise, comparing the three

geometry models, the slope decreases from the cask basket model to the infinite assembly array to the infinite

pin cell array. This effect can be seen more clearly in Figure 8, which shows the results for a single enrichment.

The results in Figure 6 confirm that, for fuel up to 5 wt % enriched in a cask basket, the maximum fuel stack

density is the optimum for criticality.

2.5 Recommendations

Based on the calculated parameter sensitivities over the range from minimum to maximum for each parameter,

the following criticality safety parameters are recommended for inclusion in the STS:

1. Fuel type

a. Array size, number of fuel rods, including number of partial length rods (where applicable), and

cladding material

b. Number and material of guide and instrument tubes for PWRs and water rods for BWRs

2. Enrichment

Other parameters considered to be significant in criticality safety licensing calculations that should require

updates to the FSAR via the Section 72.48 process are listed below. Note that these are recommended

parameters based on their influence on keff. Applicants could be permitted to eliminate any of these parameters

with appropriate justification. For example, neglecting their presence in the safety analysis could eliminate

guide/instrument tube material.

3. Fuel rod pitch

4. Fuel pellet OD

5. Cladding thickness

6. Cladding OD
Section 2 Review of Fuel Specifications for Criticality Safety

17

7. PWR guide/instrument tube thickness

8. Active fuel length

9. Fuel stack density
Review of Fuel Specifications for Criticality Safety Section 2

18

0.9900

0.9950

1.0000

1.0050

1.0100

90% 92% 94% 96% 98% 100% 102%

Effect of Fuel Density of k-inf

KENO V.a calculations

E=2 wt % (Basket)

E=5 wt % (Basket)

E=2 wt % (Pin cell)

E=5 wt % (Pin cell)

E=2 wt % (Assembly)

E=5 wt % (Assembly)

Normalized k-inf

Percent of Theoretical Density

nom inal density, 96% TD

Figure 7 Normalized 3-D k-inf versus fuel density for the 3 geometric models
Section 2 Review of Fuel Specifications for Criticality Safety

19

0.9900

0.9950

1.0000

1.0050

1.0100

90% 92% 94% 96% 98% 100% 102%

Effect of Fuel Density on k-inf

KENO V.a calculations

E=2 wt % (Basket)

E=2 wt % (Pin cell)

E=2 wt % (Assembly)

Normalized k-inf

Percent of Theoretical Density

nom inal density, 96% TD

Figure 8 Normalized 3-D k-inf versus fuel density for the 3 geometric models at 2 wt %
Review of Fuel Specifications for Criticality Safety Section 2

20
Section 3 Review of Fuel Specifications for Shielding

21

3 REVIEW OF FUEL SPECIFICATIONS FOR SHIELDING

3.1 Review Basis

A large number of fuel assembly design, cask design, and operating parameters can potentially influence the

radiation dose rates for a spent fuel storage cask. The basis for establishing a minimum set of specifications

requires criteria against which the importance of the parameters can be assessed. Unlike criticality, which has a

relatively well-defined criterion for safety, the basis for establishing parameters of importance to cask radiation

dose rates is less straightforward. Because the acceptability criteria for a storage cask are based on the dose rate

at the public boundary of a licensed storage facility (i.e., no specific local limits on the dose rate for the cask), it

may be acceptable to allow larger variations in the specifications related to dose assessment than might be

acceptable for criticality safety.

Because the dose rate is dependent on the SNF compositions, the specifications may include irradiation and

decay parameters such as burnup and cooling time, in addition to assembly design parameters. The total dose

rate includes both gamma and neutron components, and each component exhibits very different

behavior/sensitivity in different spent fuel regimes. Thus, the cask design may become a key factor due to its

influence on the relative importance of the neutron and gamma components.

3.2 Technical Specification Parameters

The candidate technical specification parameters for shielding analyses that were reviewed in this study are

listed in Table 5. Each of the parameters identified in the table has been evaluated to determine the influence of

typical variations in the parameters on the shielded cask dose rate.

The parameters commonly used to define radiation source levels in SNF are burnup and cooling time. These

parameters can vary over a wide range for spent fuel storage applications and are clearly important in

determining the dose rates. The initial fuel enrichment may also have a large effect on cask dose rates.

The other fuel specifications considered in this study include assembly uranium mass, fuel assembly type,

burnable poison rod exposure, integral burnup poison assemblies, assembly hardware and cladding type, and

moderator density and specific power during fuel irradiation. As noted in Section 2, the fuel assembly type may

be defined for groups of assembly designs with common lattice types (e.g., 14 × 14, 15 × 15, etc.) and common

characteristics (e.g., number and location of fuel rods, water holes, guide tubes, etc.). The fuel assembly type

(beyond being either a PWR- or BWR-assembly type) is typically not included as shielding technical

specifications. Similarly, exposure to burnable poison rods, moderator density, and specific power has typically

not been specified in the TS. In general, the effect of variations in these parameters has been addressed by

assuming bounding values in the source term analysis. These parameters are included in this study to provide a

wide range of different candidate parameters, and serve as a baseline against which parameter importance can be

judged.
Review of Fuel Specifications for Shielding Section 3

22

Table 5 Fuel technical specification parameters and common restrictions

Technical specifications (candidate) Commonly used restrictions

Cooling time (years) Minimum cooling time

Assembly burnup (MWd/t) Maximum burnup

Initial enrichment (wt % 235U) Minimum enrichment

Assembly/cask uranium mass (kg U) Maximum uranium content

Fuel assembly type (14 × 14, 15 × 15, 17 ×17, etc.) Unrestricted

Integral burnable poison rods (IBAs) Unrestricted

Burnable poison rods (BPRs) Unrestricted

Assembly hardware: cladding type and structural

impurity levels

Unrestricted

Moderator density (g/cm3

) Unrestricted

Specific power (MW/t) Unrestricted
Section 3 Review of Fuel Specifications for Shielding

23

3.3 Analysis Methods and Models

Storage cask surface dose rates were calculated with the SCALE 1-D radial shielding sequence SAS1 that uses

the XSDRNPM transport code. Cross sections from the SCALE coupled 27-neutron/18-gamma group transport

library were used. Spent fuel isotopic compositions and the associated neutron and gamma-ray source terms

were generated using the ORIGEN-ARP methodology, which includes burnup-dependent cross sections created

for several assembly designs that are distributed with SCALE. The SAS2 depletion analysis sequence of

SCALE was used to create additional libraries for assembly designs that did not have ARP libraries already

available. The XSDRNPM transport calculations applied the neutron and gamma fixed sources calculated by

ORIGEN-S with the fission option in XSDRNPM enabled in order to calculate secondary fissions from

subcritical neutron multiplication.

The principal storage cask design used in this study was based on the HI-STORM (Holtec International)

concrete cask design3

 constructed with approximately 68 cm of normal concrete and 5 cm of steel. The inner

diameter of the cask model was about 187 cm. The internal canister contained Boral absorber plates. Several

calculations were also performed using a carbon steel/resin type cask model4

 similar to the TN24 cask to assess

the impact of using higher Z shielding materials.

The neutron and gamma-ray components of the dose rate were calculated separately, since the behavior of the

respective dose rate components can be significantly different for many of the parameters investigated. The

relative importance of the neutron and gamma-ray components is also influenced by the cask design. The

individual dose rate components are presented to provide a means of estimating the parameter importance for

cask designs that differ from those used in this study and allow the bounding variations (for any cask design) to

be estimated.

3.4 Results

3.4.1 Principal Fuel Specification Parameters

3.4.1.1 Burnup

Previous studies of the radionuclide importance to shielding5,6 have demonstrated that the dominant effect of

increasing burnup is the dramatic increase in the spontaneous fission neutron source, primarily from 244Cm. The

neutron dose rate (Dn) has been observed to increase approximately as the burnup (B) to the power of four, i.e.,

Dn ∝ B4

. The gamma dose rate increases nearly linearly with burnup. The variation in the neutron and gamma

dose rate on the external surface of the concrete cask model, as a function of burnup, is illustrated in Figure 9 for

3 wt % fuel. The slope, of the profiles (m) from the log-log plots yields the power relationship of the dose rates

to changing burnup (e.g., D ∝ Bm). The gamma ray dose rate exhibits some nonlinear behavior, which is due

entirely to the contribution of secondary gamma rays resulting from the neutron sources. The steel cask results

were also investigated and found to be similar to those of the concrete cask. Therefore, only the concrete cask

results are presented. It can be seen that neutrons become an increasingly larger component of the total dose

rate with increasing burnup. The variation in the total dose rate for both the concrete and steel casks with

burnup is illustrated in Figure 10 assuming a constant enrichment of 3 wt %. The variation in the total dose rate

with burnup is slightly larger for the steel cask because neutrons (and secondary gamma rays) contribute a larger

fraction of the total dose rate than for the concrete cask.
Review of Fuel Specifications for Shielding Section 3

24

Gamma Dose Rate

(arb units)

Gamma Dose Rate

(arb units)

(A) Gamma-ray dose rate (linear and logarithmic scales)

(B) Neutron dose rate (linear and logarithmic scales)

Neutron 

Dose 

Rate

 (arb units)

Neutron Dose Rate

 (arb units)

0

0.01

0.02

0.03

0.04

0.05

0 20 40 60 80

Burnup (GWd/t)

0.0001

0.001

0.01

0.1

10 100

Burnup (GWd/t)

0

0.001

0.002

0.003

0.004

0.005

0.006

0.007

0 20 40 60 80

Burnup (GWd/t)

0.000001

0.00001

0.0001

0.001

0.01

10 100

Burnup (GWd/t)

5 years 20 years 100 years

Figure 9 Variation of the neutron and gamma dose rates as a function of burnup, for cooling

times of 5, 20, and 100 years. Calculated for a concrete storage cask with 3 wt % fuel.
Section 3 Review of Fuel Specifications for Shielding

25

(arb units)

(arb units)

5 years 20 years

(A) Concrete cask design (linear and logarithmic scales)

(B) Steel cask design (linear and logarithmic scales)

(arb units)

 (arb units)

100 years

0

0.01

0.02

0.03

0.04

0.05

0.06

0 20 40 60 80

Burnup (GWd/t)

T

T

T

T

o

o

o

o

t t

t

a

t

a

a

a

l l

l

D

l

D

D

D

o

o

o

o

s

s

s

e

s

e

e

e

R

R

R

R

a

a

a

a

t t

t

e

t

e

e

e

0.0001

0.001

0.01

0.1

10 100

Burnup (GWd/t)

0.0001

0.001

0.01

0.1

10 100

Burnup (GWd/t)

0

0.02

0.04

0.06

0.08

0.1

0.12

0 20 40 60 80

Burnup (GWd/t)

Figure 10 Variation of the total dose rates as a function of burnup, for cooling times of 5, 20, and

100 years for the concrete (A) and steel (B) cask designs. Calculated assuming 3 wt % fuel.
Review of Fuel Specifications for Shielding Section 3

26

The relative sensitivity, or importance, of burnup to the dose rate is dependent on many factors; the cooling

time, enrichment, fraction of the neutron dose rate component (cask design), and the burnup level (i.e., the rate

of increase is significantly greater at higher burnup than lower burnup). Therefore, it is difficult to characterize

the sensitivity with a single value that captures the importance of the parameters in all the different regimes.

3.4.1.2 Enrichment

The variation in the gamma and neutron dose rate with changing enrichment (assuming constant burnup) is

shown in Figure 11 (5-year cooling) and Figure 12 (100-year cooling) for the concrete cask. The gamma-ray

dose rate is seen to exhibit a lower sensitivity to enrichment variations than the neutron dose rate, and the

gamma contribution is seen to be effectively independent of enrichment at low burnup values. At higher

burnups, the gamma dose rate increases as the enrichment decreases because of the contribution of secondary

gamma rays caused by neutron interactions in the shielding material. The variation of the neutron dose rate is

significantly greater than for gamma rays. The neutron dose rate increases by more than a factor of two for a

reduction in enrichment from 5 to 2.5 wt %, assuming a constant burnup of 60 GWd/t.

The enrichment effect is caused by the fact that, as the enrichment decreases, the fuel must be exposed to a

larger neutron fluence (typically longer irradiation times) to achieve the same burnup. Consequently, the

neutron and secondary gamma ray dose rate contribution will increase due to a larger actinide content, and,

therefore, a larger neutron source term. The change in the total dose rate for the concrete and steel cask designs

is illustrated in Figure 13.

3.4.1.3 Cooling Time

Cooling time is a primary fuel specification parameter that has significant influence on the neutron and gamma

source term and the spectra of the gamma source term. For safety analyses, a minimum cooling time

(sometimes based on the burnup) is typically used to define acceptable spent fuel assemblies for loading.

The variation in the dose rate with cooling time has been studied extensively (Ref. 5). The variation for the

individual neutron and gamma dose rate components is illustrated in Figure 14 for an enrichment of 3.5 wt %

and burnups of 20, 40 and 60 GWd/t, for the concrete cask design. The neutron dose rate decreases at a nearconstant exponential rate for all cooling times. The gamma dose rate decreases very rapidly between 5 and

20 years cooling as the short-lived fission products decay. After 20 years cooling the dose rate decreases

exponentially. The dose rate decreases by roughly two orders of magnitude from 5 to 100-years cooling.

3.4.2 Secondary Fuel Specification Parameters

3.4.2.1 Fuel Assembly Type

The importance of the fuel assembly type on the shielded neutron and gamma cask dose rates was investigated

for three PWR assembly designs: ABB-CE 14 × 14, Westinghouse 15 × 15, and Westinghouse 17 × 17.

A summary of the assembly design specifications is given in Table 6. These designs were used to estimate the

importance of the fuel assembly type on the cask dose rates. These designs do not fully represent the range of

commercial assemblies in use, but are intended only to provide a general estimate of the level of importance of

the fuel assembly type.
Section 3 Review of Fuel Specifications for Shielding

27

(A) Neutron dose rate (linear and logarithmic scales)

(B) Gamma dose rate (linear and logarithmic scales)

20 GWd/t 40 GWd/t 60 GWd/t

Neutr

on 

d

ose rate

(arb. units)

0

0.001

0.002

0.003

0.004

0.005

23456

Enrichment (wt %)

Neutron dose rate

(arb. units)

0.00001

0.0001

0.001

0.01

1 10

Enrichment (wt %)

0

0.01

0.02

0.03

0.04

23456

Enrichment (wt %)

Gamma dose rate

(arb. units)

0.001

0.01

0.1

1 10

Enrichment (wt %)

Gamma dose rate

(arb. units)

Figure 11 Variation of the neutron (A) and gamma (B) dose rate as a function of enrichment, for

burnups of 20, 40, and 60 GWd/t for the concrete cask design, assuming 3 wt % enrichment and 5 year

cooling.
Review of Fuel Specifications for Shielding Section 3

28

(A) Neutron dose rate (linear and logarithmic scales)

(B) Gamma dose rate (linear and logarithmic scales)

20 GWd/t 40 GWd/t 60 GWd/t

0

0.00005

0.0001

0.00015

0.0002

0.00025

23456

Enrichment (wt %)

Neutron dose rate

(arb. units

Neutron dose rate

arb. units

)

0.000001

0.00001

0.0001

0.001

1 10

Enrichment (wt %)

0

0.0002

0.0004

0.0006

0.0008

23456

Enrichment (wt %)

Gamma dose rate

(arb. units)

0.00001

0.0001

0.001

1 10

Enrichment (wt %)

Gamma dose rate

(arb. units)

Figure 12 Variation of the neutron (A) and gamma (B) dose rate as a function of enrichment, for

burnups of 20, 40, and 60 GWd/t for the concrete cask design, assuming 3 wt % enrichment and

100 years cooling.
Section 3 Review of Fuel Specifications for Shielding

29

20 GWd/t 40 GWd/t

(A) Concrete cask design, 5-year cooling

(C) Steel cask design, 5-year cooling

(B) Concrete cask design, 100-year cooling

(D) Steel cask design, 100-year cooling

60 GWd/t

0

0.01

0.02

0.03

0.04

0.05

23456

Enrichment (wt %)

Total dose rate

(arb. units)

0

0.0002

0.0004

0.0006

0.0008

0.001

23456

Enrichment (wt %)

Total dose rate

(arb. units)

0

0.0005

0.001

0.0015

0.002

0.0025

0.003

23456

Enrichment (wt %)

Total dose rate

(arb. units)

0

0.02

0.04

0.06

0.08

0.1

23456

Enrichment (wt %)

Total dose rate

(arb. units)

Figure 13 Variation of the total dose rate as a function of enrichment, for burnup of 20, 40, and

60 GWd/t for concrete and steel cask designs (5 year and 100 year cooling times shown)
Review of Fuel Specifications for Shielding Section 3

30

0.001

0.01

0.1

1

10

0 20 40 60 80 100

Cooling time (Y)

Neutron Dose Rate

(arb units)

0.01

0.1

1

10

100

0 20 40 60 80 100

Cooling time (Y)

Gamma Dose Rate

(arb units)

(A) Gamma dose rate

(B) Neutron dose rate

Figure 14 Variation of the gamma (A) and neutron (B) dose rates as a function of cooling time for

an enrichment of 3.5 wt % and burnups of 20, 40, and 60 GWd/t, for the concrete cask design.
Section 3 Review of Fuel Specifications for Shielding

31

Table 6 Design specifications and operational conditions for selected fuel assembly types

(no burnable poisons)

Design parameter ABB-CE

14 × 14

Westinghouse

15 × 15

Westinghouse

17 × 17

Assembly pitch (cm) 20.780 21.4503 21.4173

Number of fuel rods 176 204 264

Number of instrument tubes 0 1 1

Number of guide tubes 20 20 24

Number of burnable poison rods 0 0 0

Fuel type UO2 UO2 UO2

Enrichment range (wt % 235U) 2.5 ñ 5.0 2.5 ñ 5.0 2.5 ñ 5.0

Fuel density (g/cm3

) 10.045 9.44 10.32

Fuel temperature (K) 873 923 811

Clad type Zircaloy Zircaloy Zircaloy-4

Clad temperature (K) 620 595 620

Fuel rod data

Fuel rod outer diameter (cm) 0.9563 0.9294 0.81915

Gap outer diameter (cm) 0.9855 0.9484 0.83566

Clad outer diameter (cm) 1.1176 1.0719 0.94966

Fuel rod pitch (cm) 1.4732 1.430 1.25984

Guide tube data

Inner radius (cm) 0.6570 0.6502 0.57150

Outer radius (cm) 0.7080 0.6934 0.61214

Guide tube material Zircaloy Zircaloy Zircaloy-4

Moderator data

Average density (g/cm3) 0.7332 0.7135 0.7295

Average boron concentration (ppm) 331 653 550

Moderator temperature (K) 570 579 570
Review of Fuel Specifications for Shielding Section 3

32

The total mass of uranium in the assemblies (and cask) was assumed to be the same for all fuel assembly types.

Therefore, the variations in the dose rate are due only to the changes in actinide and fission product inventories

attributed to the different assembly designs (i.e., spectral differences in fuel region during irradiation). These

assemblies did not use burnable poison rods, which are addressed as a separate fuel parameter.

The neutron and gamma dose rates for each fuel assembly type were calculated for initial 235U enrichments of

2.5 to 5 wt % and burnups of 20, 40 and 60 GWd/t. Assembly cooling times of 5 years and 100 years were

evaluated. The total dose rate exhibited a maximum difference of about 6% between the ABB-CE 14 × 14 and

W 17 × 17 designs. Significantly less difference was observed between the W 15 × 15 and W 17 × 17 designs.

The gamma component exhibited less variability than the neutron component. The maximum variation in the

neutron dose rate was about 20% for the different assembly designs considered. For both the steel and concrete

cask designs studied, neutrons were generally a smaller component of the total dose rate than gamma rays and,

consequently, the variation of the gamma dose rate dominated the variation in the total dose rate. For cask

designs and spent fuel regimes with a larger neutron dose component, the effect of assembly type could be

larger than observed here, but in any case the differences are likely to be < 10% for any cask design.

3.4.2.2 Burnable Poison Assemblies

The effects of exposure to burnable poisons rods (BPRs) and integral burnable absorbers (IBAs) were

investigated to determine: (1) the effect on the spent fuel compositions and, consequently, the cask dose rates

caused by the shift in the neutron spectrum during irradiation for the fuel associated with the presence of neutron

poisons; and (2) the potential direct contribution from activated hardware components associated with the BPRs

to the dose rate, when irradiated BPRs are loaded in the cask assemblies.

Two removable BPR designs were considered: (1) a B4C-Al absorber design, and (2) a steel and borosilicate

glass burnable poison rod design. The spectral effects of the B4C absorbers and borosilicate glass absorbers are

expected to be similar, because both designs displace moderator in the guide tubes, which has been

demonstrated to be the dominant spectral factor affecting the spent fuel compositions.7

 However, the

borosilicate glass design uses a significant quantity of stainless steel, which will present a large potential

activation product (cobalt) source. Therefore, only the borosilicate glass BPR design was evaluated.

Two IBA designs were considered. Cask shielding evaluations were performed for a ABB-CE 16 × 16 design

using 8 integral Gd2O3 burnable poison rods, and an ABB-CE 14 × 14 design with 60 Er2O3 burnable poison

rods. These were only intended to represent some typical assembly burnable poison absorber designs to

estimate the potential impact of integral absorbers on shielded cask dose rates. The assumed 235U enrichment for

these studies was 4 wt % (uniform assembly enrichment). The poison loadings were 5 wt % Gd2O3 and 2 wt %

Er2O3, respectively. The IBA assembly design specifications used for these studies are listed in Table 7.
Section 3 Review of Fuel Specifications for Shielding

33

Table 7 Design specifications for integral burnable poison rod fuel assemblies

Design parameter ABB-CE

16 × 16

ABB-CE

14 × 14

Assembly pitch (cm) 20.78 20.8

Number of fuel rods 228 116

Number of water holes 5 5

Number of burnable poison rods 8 60

Burnable poison 5 wt % Gd2O3 2 wt % Er2O3

Fuel type UO2 UO2

Enrichment (wt % 235U) 4.0 4.0

Fuel density (g/cm3

) 10.44 10.44

Fuel temperature (K) 1000 1000

Clad type Zircaloy Zircaloy

Clad temperature (K) 620 620

Fuel rod data

Fuel rod outer diameter (cm) 0.82550 0.95631

Gap outer diameter (cm) 0.84328 0.97536

Clad outer diameter (cm) 0.97028 1.1176

Fuel rod pitch (cm) 1.285 1.470

Guide tube data

Inner radius (cm) 1.1430 1.1532

Outer radius (cm) 1.2446 1.2040

Guide tube material Zircaloy Zircaloy

Moderator data

Average density (g/cm3) 0.710 0.71

Average boron concentration (ppm) 650 650

Moderator temperature (K) 600 600
Review of Fuel Specifications for Shielding Section 3

34

The borosilicate glass absorber rod design used in the study was constructed with two concentric stainless steel

tubes with the inner region containing inert gas and the interstitial region containing the borosilicate glass

matrix. The nominal BPR design was obtained from Ref. 8 and used to derive representative absorber and

hardware masses for an assembly. This design was used with Westinghouse 17 × 17 assemblies in the North

Anna Unit 1 reactor. The assembly lattice contains 264 fuel rods and 24 guide tubes, and 1 instrument tube.

The different BPR clusters used in this reactor during cycle 5 contained 12, 16, or 24 rods. The shielding

assessment was performed assuming the assembly nominally contained 12 BPRs. The potential number of BPR

rods could range up to 24 rods per assembly. Therefore, this component of the dose rate evaluation will be

highly dependent on the number of rods assumed to be present in the assemblies, and the number of assemblies

containing irradiated BPR clusters. This analysis assumed that all assemblies in the storage cask contained

BPRs with 12 rods per cluster. The mass of stainless steel per MTU is about 5.6 kg. However, the thermal

neutron flux level in the absorber region is about 70% that of the fuel region. Consequently, the effective mass

of stainless steel used in the activation calculations was correspondingly reduced to 4 kg to account for the flux

difference between the regions. The stainless steel in the BPRs was assumed to have a nominal cobalt impurity

level of 800 ppm, a value associated with older assembly designs. The cobalt mass was therefore equal to 3.2 g

cobalt per MTU of fuel (0.0032 g/kg U). This value is actually less than the estimated 0.075 g/kg U cobalt level

associated with older PWR assemblies from inconel grid spacers and other assembly hardware.

The neutron and gamma dose rates for the concrete cask design were calculated using the different burnable

poison assembly models for a uniform fuel enrichment of 4 wt %, burnups of 20 and 60 GWd/t, and cooling

times of 5 and 100 years. These results are compared with the dose rates for a reference fuel design with no

burnable poisons (results for the Westinghouse 17 × 17 design listed in Table 6) in Figure 15. All calculations

assumed a uniform uranium mass per assembly. The variations in the dose rate are due entirely to the spectral

effects caused by burnable poison exposure during the irradiation, with the exception of the borosilicate glass

BPRs which are calculated with and without the contribution of activated cobalt to the gamma dose rate. These

latter calculations indicate that the level of cobalt associated with the borosilicate glass rods does not have a

significant effect on the gamma dose rates. However, the results will be strongly dependent on the cobalt level

in activated hardware and the number of rods stored in the cask assemblies.

Figure 15 plots the ratio of the dose rates for assemblies with burnable poisons to the reference assembly

without burnable poisons. The results indicate that the impact of the IBAs is greatest on the neutron dose rate.

This effect is most pronounced at the lower burnup. However, at low burnup the relative importance of the

neutron dose rate component is typically small. At higher burnup, the effects of the assembly poison are

negligible and the neutron dose rates are very similar to the reference assembly with no exposure to burnable

poisons. The total dose rate, which is typically dominated by gamma rays, is about 2% higher for the IBA

assemblies compared to the reference fuel with no burnable poisons. The BPRs have a slightly larger effect,

particularly at longer cooling time. In all cases, the dose rates increased by no more than about 10% for any of

the assemblies considered.

3.4.2.3 Uranium Mass

The total uranium mass of a storage cask depends on the number of assemblies that can be loaded into a

particular cask and the uranium mass of each assembly. Variations in the uranium mass for a given cask design

may result from the variations in the uranium mass for the range of assembly types approved for the cask. An

increase in the uranium mass per assembly results in a larger radiation source term. This is offset to some extent

by greater self-attenuation of the source due to the larger amount of uranium that acts as a self-shield.

Increasing the uranium mass also has a small effect on the neutron subcritical multiplication factor of the cask

that may increase the secondary fission source.
Section 3 Review of Fuel Specifications for Shielding

35

0.9

1

1.1

1.2

1.3

1.4

Gd2O3 IBA(20 GWd/t)Gd2O3 IBA(60 GWd/t)

Er2O3 IBA(20 GWd/t)Er2O3 IBA(60 GWd/t)

Borosilicate BPR without cobalt (20 GWd/t)Borosilicate BPR without cobalt (60 GWd/t)Borosilicate BPRwith cobalt(20 GWd/t)Borosilicate BPRwith cobalt(60 GWd/t)

R

R

a

a t t i i o

o ( ( p

p

o

o i i s

s

o

o

n

n

a

a

s

s

s

s

e

e

m

m

b

b l l y

y / / r r e

e f f a

a

s

s

s

s

e

e

m

m

b

b l l y

y ) )

0.9

1

1.1

1.2

1.3

1.4

1.5

1.6

1.7

1.8

1.9

Gd2O3 IBA(20 GWd/t)Gd2O3 IBA(60 GWd/t)

Er2O3 IBA(20 GWd/t)Er2O3 IBA(60 GWd/t)

Borosilicate BPR without cobalt (20 GWd/t)Borosilicate BPR without cobalt (60 GWd/t)Borosilicate BPRwith cobalt(20 GWd/t)Borosilicate BPRwith cobalt(60 GWd/t)

Figure 15 Ratio of the neutron, gamma, and total dose rate for assemblies with and without

burnable poisons
Review of Fuel Specifications for Shielding Section 3

36

The impact of variations in assembly uranium mass on the radiation dose rates was studied for the concrete and

steel type casks. The variation in uranium mass was obtained from a U.S. Department of Energy (DOE)

database9

 containing assembly design and irradiation history information for all spent fuel assemblies

discharged from commercial operating reactors in the United States through 1998. The database contains the

uranium mass for over 130,000 individual assemblies (approximately 76,000 BWR assemblies and 56,000 PWR

assemblies) currently in storage.

The distribution of uranium mass for PWR and BWR assemblies is illustrated in Figure 16. Note that although

the variation in mass is quite large when all assemblies are considered, most assemblies reside in a relatively

well-defined mass range. The PWR assembly outliers include assemblies from Yankee Rowe (most assemblies

below 300 kg U) and South Texas Units 1 and 2 (assemblies greater than 500 kg U) reactors. For the BWR fuel

assemblies, the outliers (below 150 kg U) include mainly assemblies from Dresden 1, La Crosse, and

Big Rock Point reactors.

The uranium mass variation for the large majority of assemblies is approximately ±20% (about the mean) for the

PWR assemblies and approximately ±10% for the BWR assemblies currently in storage. The variation is much

larger if the outlier assemblies were to be included in the distribution. However, the only group of assemblies

that have a significantly larger mass than the mean, and hence could result in dose rates well above the mean,

are roughly 800 assemblies from the South Texas PWR reactors.

The variation in the dose rate, as a function of the uranium mass in the casks, was calculated for the neutron and

gamma dose rate components by increasing the reference uranium mass by 10%. The maximum increase in the

neutron dose rate for both the concrete and steel casks was about 6%, while the gamma ray dose rate increased

by about 2%. These values were largely independent of enrichment and burnup. The maximum change in the

total dose rate for either cask design was only about 3% due to domination of the gamma dose rate for the cask

designs considered.

When these results are combined with the typical variation in the uranium mass for different assembly designs

(maximum of ±20% for PWR assemblies), the expected variation in the cask surface dose rate is less than ±10%

for any cask design. For BWR assembly designs, which exhibit lower relative variability, the dose rate variation

will be lower. For concrete storage casks, which typically have a lower neutron dose rate component compared

to a steel case design, the variation will also be lower (< ±5%) since the gamma dose rate has a lower sensitivity

to uranium mass than the neutron dose rate.

3.4.2.4 Specific Power

The specific power of the fuel during irradiation is typically not included in fuel specifications. The

predominant effect of the specific power (particularly near the end of life) is on the short-lived fission product

inventory, which affects the radiation source term and decay heat power. Variations in specific power may

occur as a result of many factors such as reactor operating history, changing assembly position in the core, and

axial power variations in an assembly.

The effect of specific power on the cask dose rates was investigated for 3.5 wt % PWR fuel irradiated to

40 GWd/t using specific powers of 20, 30, and 40 MW/t. The neutron and gamma dose rates were compared at

cooling times from 5 to 100 years. The results confirmed that the neutron dose is largely unaffected by the

specific power level. The gamma dose rate increases with higher specific power, with the effect most

pronounced at short cooling times. For 5-years cooling, the gamma dose rate increased by about 30% using a

specific power of 40 MW/t compared to 20 MW/t. After a 10-year cooling time the increase was 7%, and after

20 years the effect decreased to only 3%.
Section 3 Review of Fuel Specifications for Shielding

37

0

5000

10000

15000

20000

25000

30000

35000

40000

45000

0-100

100-110110-120120-130130-140140-150150-160160-170170-180180-190190-200

Uranium mass (kg U)

No. of assemblies

(A) PWR Assemblies

(B) BWR Assemblies

0

2000

4000

6000

8000

10000

12000

14000

16000

18000

200-220220-240240-260260-280280-300300-320320-340340-360360-380380-400400-420420-440440-460460-480480-500500-520520-540540-560

Uranium mass (kg U)

No. of assemblies

Figure 16 Distribution of uranium mass for currently discharged PWR (A) and BWR (B) fuel

assemblies. The distributions reflect the actual inventory of spent fuel assemblies through 1998 from

commercial reactor operation.
Review of Fuel Specifications for Shielding Section 3

38

3.4.2.5 Moderator Density

Reactor moderator density is another parameter not typically included as a shielding technical specification.

The relative importance of the moderator density in hardening the neutron spectrum and influencing the SNF

inventories was assessed for a range of moderator densities associated with axial variations in BWR assemblies

(typically 0.3 ñ 0.7 g/cm3

). The calculations assumed fuel with 4 wt % and 40 GWd/t. The neutron and gamma

dose rates were observed to increase with decreasing water moderator density. The neutron dose rate was found

to vary by more than 30% over the moderator density range, while the gamma dose rate varied by about 10%.

The net impact of moderator density on cask shielding is expected to be low for PWR fuels. However, the axial

variation in moderator density in BWR reactors can have a measurable effect on the axial dose rate profile on a

cask, and increase the dose rate near the top of the assemblies where the moderator density is lowest.

3.4.2.6 Fuel Cladding Material

The impact on the dose rates from loading assemblies with stainless steel fuel cladding was estimated by scaling

the results for the assemblies containing borosilicate glass and stainless steel BPRs (12 BPR rods per assembly).

The amount of stainless steel associated with the BPR rod design studied is about one half that of a fuel rod with

steel cladding. Therefore, the dose rate results for the borosilicate glass absorbers were divided by the ratio of

the number of fuel rods to BPR rods (12) and scaled by a factor of two. For example, in a 16 × 16 design with

228 fuel rods, the dose rates for the assembly containing 12 BPR rods (with stainless steel) were scaled by a

factor of roughly 40 to estimate the dose rate for an assembly containing 228 fuel rods clad with stainless steel.

The potential impact on the gamma dose rate would be very large for the cooling times during which 60Co is the

dominant gamma ray source (up to about 50 years). The steel clad fuel potentially increases the cask dose rate

by more than an order of magnitude over that from conventional Zircaloy clad fuel.

3.4.3 Combined Burnup, Enrichment, and Cooling Time Effects

The discharge burnup and cooling time are clearly dominant parameters for shielding, and established

combinations of burnup and cooling time for acceptable spent fuel assemblies have been used previously in

technical specifications. The neutron source exhibits a rapid increase with burnup, which is most pronounced

when fuel is irradiated well beyond a typical discharge burnup as determined by the initial enrichment. To limit

the increase in the neutron dose rate, technical specifications have sometimes imposed a minimum fuel

enrichment.

The variation of the neutron and gamma dose rate for combinations of burnup and enrichment was investigated

for a concrete storage cask and is shown in Figure 17. The contour plot lines indicate the regimes

(enrichment/burnup combinations) of constant dose rate. For example, the neutron dose rate for spent fuel with

4.5 wt % and 50 GWd/t is seen to be nearly equal to 2.5 wt % and 40 GWd/t. The figure illustrates the rapid

increase in the neutron dose component in the overburned (low enrichment and high burnup) region. The large

gradient in the high burnup region (rapid rate of change) indicates that the sensitivity of the dose rate to changes

in enrichment will be significantly greater for high burnup fuel than low burnup fuel. That is, a given change in

burnup, for example, will have a much larger affect on the dose rate for high burnup fuel than for low burnup

fuel. The figure also illustrates that combining enrichment and burnup as a single parameter may provide a

basis for a less restrictive and more general technical specification involving the correlated parameters of

enrichment, burnup, and cooling time. Such an approach would be similar to the loading curve proposed for

burnup credit whereby only spent fuel with a specified enrichment and burnup would be acceptable for loading.

As an example, loading could be restricted to spent fuel with a minimum enrichment-to-burnup ratio instead of

imposing a minimum enrichment alone. This might permit low enrichment fuel to be loaded in the cask if there

was a commensurate reduction in the burnup of the assembly. To illustrate such an approach, consider a

concrete storage cask licensed for 5-year-cooled fuel (Figure 17). If the shielding analysis for the cask
Section 3 Review of Fuel Specifications for Shielding

39

demonstrated spent fuel with 5 wt % and 60 GWd/t was acceptable for loading, it can be seen that any fuel with

an enrichment-to-burnup ratio greater than or equal to the licensing basis fuel (e.g., 3 wt % up to 36 GWd/t)

would generate a dose rate less than the safety analysis value.

A less restrictive prescription that provided a more realistic representation of the dose rate dependence would

allow a greater range of assemblies to be safely loaded in a storage cask. Such a loading curve, however, would

need to consider the specific cask design since the neutron and gamma dose rate behavior is significantly

different (e.g., the gamma dose rate varies more slowly than neutron dose rate with changes in enrichment).

The changing neutron and gamma dose rates for the concrete storage cask for different burnup and cooling time

combinations are illustrated in Figure 18. The total dose rate for the cask is dominated by gamma rays

(i.e., neutrons contribute less than 25% of the total for high burnup fuel and long cooling times) and; therefore,

the total dose rate profile is dominated by the gamma ray behavior.
Review of Fuel Specifications for Shielding Section 3

40

(A) Total Dose Rate (arb. units)

Burnup (

G

Wd/t)

Enrichment (wt %)

2.5 3 3.5 4 4.5 5

20

40

60

7.00E-02-8.00E-02

6.00E-02-7.00E-02

5.00E-02-6.00E-02

4.00E-02-5.00E-02

3.00E-02-4.00E-02

2.00E-02-3.00E-02

1.00E-02-2.00E-02

0.00E+00-1.00E-02

(B) Neutron Dose Rate (arb. units)

Burnup (GWd/t)

2.5 3 3.5 4 4.5 5

4.50E-02-5.00E-02

4.00E-02-4.50E-02

3.50E-02-4.00E-02

3.00E-02-3.50E-02

2.50E-02-3.00E-02

2.00E-02-2.50E-02

1.50E-02-2.00E-02

1.00E-02-1.50E-02

5.00E-03-1.00E-02

0.00E+00-5.00E-03

20

40

60

Enrichment (wt %)

Burnup (

GWd/t)

(C) Gamma Dose Rate (arb. units)

2.5 3 3.5 4 4.5 5

2.80E-02-3.20E-02

2.40E-02-2.80E-02

2.00E-02-2.40E-02

1.60E-02-2.00E-02

1.20E-02-1.60E-02

8.00E-03-1.20E-02

4.00E-03-8.00E-03

0.00E+00-4.00E-03

20

40

60

Enrichment (wt %)

Figure 17 Enrichment-burnup dose rate contour plots for concrete storage cask, 5-y cooling
Section 3 Review of Fuel Specifications for Shielding

41

5 1

Cooling time (years)

Burnup (GWd/t)

0 20 30 50 100

20

40

60 3.00E-02-3.20E-02

2.80E-02-3.00E-02

2.60E-02-2.80E-02

2.40E-02-2.60E-02

2.20E-02-2.40E-02

2.00E-02-2.20E-02

1.80E-02-2.00E-02

1.60E-02-1.80E-02

1.40E-02-1.60E-02

1.20E-02-1.40E-02

1.00E-02-1.20E-02

8.00E-03-1.00E-02

6.00E-03-8.00E-03

4.00E-03-6.00E-03

2.00E-03-4.00E-03

0.00E+00-2.00E-03

(A) Total Dose Rate (arb. units)

Cooling time (years)

Burnup (GWd/t) 20

40

60

5 10 20 30 50 100

2.60E-03-2.80E-03

2.40E-03-2.60E-03

2.20E-03-2.40E-03

2.00E-03-2.20E-03

1.80E-03-2.00E-03

1.60E-03-1.80E-03

1.40E-03-1.60E-03

1.20E-03-1.40E-03

1.00E-03-1.20E-03

8.00E-04-1.00E-03

6.00E-04-8.00E-04

4.00E-04-6.00E-04

2.00E-04-4.00E-04

0.00E+00-2.00E-04

(B) Neutron Dose Rate (arb. units)

Burnup (GWd/t)

20

40

60

5 10 20 30 50 100

2.80E-02-3.00E-02

2.60E-02-2.80E-02

2.40E-02-2.60E-02

2.20E-02-2.40E-02

2.00E-02-2.20E-02

1.80E-02-2.00E-02

1.60E-02-1.80E-02

1.40E-02-1.60E-02

1.20E-02-1.40E-02

1.00E-02-1.20E-02

8.00E-03-1.00E-02

6.00E-03-8.00E-03

4.00E-03-6.00E-03

2.00E-03-4.00E-03

0.00E+00-2.00E-03

(C) Gamma Dose Rate (arb. units)

Cooling time (years)

Figure 18 Burnup-cooling time dose rate contour plots for concrete storage cask
Review of Fuel Specifications for Shielding Section 3

42

3.5 Recommendations

A summary of findings is given in Table 8. The reported variations in the table are presented for a concrete

storage cask design, and may be somewhat different for alternate cask designs that different shielding materials

and/or have a larger relative neutron dose rate component. The variations in the dose rates reflect the limited

range of parameter variations and fuel assembly designs considered here. They are not intended to cover the full

range of spent fuel and assembly designs used in commercial PWR and BWR reactors. These results are

intended only to provide a rough guide to the importance of the respective candidate fuel specifications as they

affect typical storage cask dose rates to allow judgments to be made on the relative importance of the

parameters.

The parameters having the largest impact on the dose rate are burnup and cooling time. In addition, assemblies

containing significant quantities of stainless steel (e.g., steel cladding or BPR rods containing steel) may also

have a very large effect on the dose rates. The enrichment level is also observed to be important. The

magnitude of the enrichment effect, like many of the parameter effects, is dependent on the relative neutron-togamma ray dose rate for the cask, since the gamma ray dose rate exhibits a smaller sensitivity to changes in

enrichment.

Assemblies with integral burnable absorbers were found to have dose rates similar to those for nonpoison

assemblies. Larger effects were observed for assemblies exposed to burnable poison rods during irradiation,

caused primarily by the spectral effects of displacing moderator in the assembly. However, these effects were

most pronounced for the neutron-dose rate, and had only small effects on the total shielded dose rate for the

cask. The largest potential effect from assemblies residing in a cask that contains irradiated BPR clusters is

from activated component hardware (mainly activated cobalt in steel). For BPR designs containing stainless

steel, the impact on the gamma dose rate can be large.

The uranium mass was found to be of intermediate importance to shielding. The variation in the mass for

different assembly designs is typically less than ±20% and results in a maximum variation in the cask exterior

surface dose rate of less than 10%. The variation in the dose rate due to typical variations in fuel assembly type,

various integral burnable poison designs, BPR exposure, moderator density, and specific power were all found

to be less than about 10% over most of the parameter range studied.
Section 3 Review of Fuel Specifications for Shielding

43

Table 8 Summary of fuel specification study finding

Technical specifications (candidate) Parameter range studied Observed variationa Bo

Initial enrichment (wt % 235U) 2.5 ñ 5.0 100% MinimCooling time (years) 5 ñ 100 > 100% MinimAssembly burnup (MWd/t) 20 ñ 60 > 100% MaAssembly/cask uranium mass (kg U) mean 

± 20 % 20% M

Fuel assembly type (no burnable poisons) ABB-CE 14 

× 14, W 15 

× 15,

W 17 

× 17

5% 17 

Integral burnable poison rods (IBAs) 5 wt % Gd2O3, 2 wt % Er2O3 5% Maxim

Burnable poison rods (BPRs) 12 borosilicate glass and

stainless steel rods

10% Maximum 

cobalt l

Assembly structural materials stainless steel cladding > 100% MaximModerator density (g/cm3) ñ BWR 0.3 ñ 0.7 10% Minimum 

Specific power (MW/t) 20 ñ 40 10% after 5 years

< 5% after 10 years

Maxim

a Approximate maximum variation in total dose rate expressed as percentage difference over full range = (max/min ñ 1) 

storage cask design

43
Summary Section 6

44
45

4 SUMMARY

The use of STS for spent fuel storage casks would benefit both the NRC and licensees because of the reduced

personnel time and costs associated with the reduction in license amendment submittals. Applicants should

perform safety analysis for unique fuel types (i.e., array size, number of fuel rods, cladding material) and

develop bounding specifications based on the safety analysis that include all parameters important to safety.

These specifications should be defined in the FSAR. The safety analysis should include justification for the

exclusion of any parameters from the FSAR. Fuel assemblies that satisfy the STS, but are not explicitly covered

in the FSAR (e.g., one or more parameters are outside of the defined range in the FSAR), could be added using

the Section 72.48 process.

As stated, this type of approach does not enable the addition of new fuel types (i.e., assemblies with differing

array sizes, numbers of fuel rods, or number and location of guide tubes) or increases in enrichment limits.

Consequently, the burden still rests on the applicants to provide general analyses that consider the unique fuel

types that may potentially be stored in their cask(s).

This report has studied the important parameters that influence criticality safety and radiation shielding doses.

The following parameters are recommended for inclusion in the STS.

1. Fuel type

a. Array size, number of fuel rods, including number of partial length rods (where applicable), and

cladding type

b. Number and material of guide and instrument tubes;

2. Enrichment (maximum for criticality safety, minimum for radiation shielding);

3. Maximum burnup;

4. Minimum cooling time;

5. Maximum uranium mass; and

6. Maximum cobalt level.

Other criticality safety parameters (pitch, pellet OD, clad thickness, clad OD, guide tube and water rod

thickness, and fuel stack density) could be updated via the Section 72.48 process. Likewise, radiation shielding

parameters (e.g., maximum poison loading, minimum moderator density (BWR), and maximum specific power)

could be updated via the Section 72.48 process.
46
47

5 REFERENCES

 

1. H. R. Dyer, C. V. Parks, Recommendations for Preparing the Criticality Safety Evaluation of

Transportation Packages, NUREG/CR-5661 (ORNL/TM-11936), U.S. Nuclear Regulatory Commission,

Oak Ridge National Laboratory, April 1997.

2. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing

Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2R5), Vols. I, II, and III, May 2000.

Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as

CCC-545.

3. Safety Analysis Report for the HI-STORM 100 System, Holtec International Report HI-951312. February 1999.

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PWR Spent-Fuel Dry Storage Cask Loaded with Consolidated Fuel, EPRI NP-6191 (PNL-6631),

February 1989.

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Importance to Functional Requirements Related to Transport and Long-Term Storage of LWR Spent Fuel,

ORNL/TM-12742, Martin Marietta Energy Systems, Oak Ridge National Laboratory, June 1995.

6. I. C. Gauld and J. C. Ryman, Nuclide Importance to Criticality Safety, Decay Heating, and Source Terms

Related to Transport and Interim Storage of High-Burnup LWR Fuel, NUREG/CR-6700 (ORNL/TM2000/284), U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, January 2001.

7. J. C. Wagner and C. V. Parks, "Impact of Burnable Poison Rods on PWR Burnup Credit Criticality Safety

Analyses," ANS/ENS 2000 International Winter Meeting and Embedded Topical Meetings, November 12ñ

16, 2000, Washington, D.C. Trans. Am. Nucl. Soc. 83, 130ñ134, 2000.

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Volume 5 – North Anna Unit 1 Cycle 5, ORNL/TM-12294/V5, Lockheed Martin Energy Research Corp.,

Oak Ridge National Laboratory, October 1996.

9. Department of Energy, Office of Civilian Radioactive Waste Management, Form RW-859 Nuclear Fuel

Data File, collected by Energy Information Administration EIA.
 
49

NUREG/CR-6716

ORNL/TM-2000/385

INTERNAL DISTRIBUTION

1. S. M. Bowman, 6011, MS-6370 15. C. V. Parks, 6011, MS-6370

2. B. L. Broadhead, 6011, MS-6370 16. L. M. Petrie, 6011, MS-6370

3. W. C. Carter, 6011, MS-6370 17. C. E. Pugh, 9201-3, MS-8063

4. M. D. DeHart, 6011, MS-6370 18. C. E. Sanders, 6011, MS-6370

5. M. E. Dunn, 6011, MS-6370 19. J. C. Wagner, 6011, MS-6370

6. K. R. Elam, 6011, MS-6370 20. R. M. Westfall, 6011, MS-6370

7. M. B. Emmett, 6011, MS-6370 21. Laboratory Records-RC

8. I. C. Gauld, 6011, MS-6370 4500N, MS-6285

9. S. Goluoglu, 6011, MS-6370 22. Central Research Library

10. D. F. Hollenbach, 6011, MS-6370 4500N, MS-6191

11. C. M. Hopper, 6011, MS-6370 23%50. Return extra ORNL copies to: 

12. M. A. Kuliasha, 6025, MS-6435 W. C. Carter, 6011, MS-6370

13. A. Loebl, 6025, MS-6435

14. B. D. Murphy, 6011, MS-6370

EXTERNAL DISTRIBUTION

51. M. G. Bailey, U.S. Nuclear Regulatory Commission, NMSS/SFPO/SLID, MS T8-A23,

Washington, DC 20555-0001

52. D. E. Carlson, U.S. Nuclear Regulatory Commission, NMSS/SFPO/TRD, MS O13-D13,

Washington, DC 20555-0001

53%54. Office of Scientific and Technical Information, U.S. Department of Energy, P.O. Box 62,

Oak Ridge, TN 37831

55. N. L. Osgood, U.S. Nuclear Regulatory Commission, NMSS/SFPO/SLID, MS O13-D13,

Washington, DC 20555-0001

56. B. H. White IV, U.S. Nuclear Regulatory Commission, NMSS/SFPO/TRD, MS O13-D13,

Washington, DC 20555-0001

57. C. J. Withee, U.S. Nuclear Regulatory Commission, NMSS/SFPO/TRD, MS O13-D13,

Washington, DC 20555-0001
50
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION

(2-89)

NRCM 1102 BIBLIOGRAPHIC DATA SHEET

3201, 3202 ( See instructions on the reverse)

1. REPORT NUMBER

(Assigned by NRC, Add Vol., Supp.,

Rev., and Addendum Numbers,

 if any.)

NUREG/CR-6716 

ORNL/TM-2000/385 2. TITLE AND SUBTITLE

Recommendations on Fuel Parameters for Standard Technical Specifications 

for Spent Fuel Storage Casks 3. DATE REPORT PUBLISHED

MONTH

March 

YEAR

2001

4. FIN OR GRANT NUMBER

B0009

5. AUTHOR(S)

 S. M. Bowman, I. C. Gauld, J. C. Wagner

 

6. TYPE OF REPORT

Technical

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION — NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing

address; if contractor, provide name and mailing address.)

Oak Ridge National Laboratory

Managed by UT-Battelle, LLC

Oak Ridge, TN 37831-6370

9. SPONSORING ORGANIZATION — NAME AND ADDRESS (If NRC, type “Same as above”; if contractor, provide NRC Division, Office or Region, U.S. Regulatory

Commission, and mailing address.)

Spent Fuel Project Office

Office of Nuclear Material Safety and Safeguards

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

C. J. Withee, NRC Project Manager

11. ABSTRACT (200 words or less)

The U.S. Nuclear Regulatory Commission (NRC) is currently reviewing the technical specifications for spent fuel 

storage casks in an effort to develop standard technical specifications (STS) that define the allowable spent nuclear 

fuel (SNF) contents. One of the objectives of the review is to minimize the level of detail in the STS that define the 

acceptable fuel types. To support this initiative, this study has been performed to identify potential fuel specification 

parameters needed for criticality safety and radiation shielding analysis and rank their importance relative to a 

potential compromise of the margin of safety.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)

 cask, license, technical specifications, standardized, criticality, shielding, 72.48, Part 72

13. AVAILABILITY STATEMENT

 unlimited

14. SECURITY CLASSIFICATION

(This Page)

unclassified

(This Report) unclassified

15. NUMBER OF PAGES

16. PRICE

15&)250