Criticality Safety Calculation and Analysis for NPP Transportation of

Fuel Assemblies

Dajie Zhuang

China Institute for Radiation Protection, Taiyuan, 030006, China

Keywords: Criticality Safety, Benchmark Certification, LEU (Low-Enriched Uranium), Rods Lattice.

Abstract: Nuclear criticality safety was calculated by MC code for transportation activity of fuel assemblies to

Sanmen Nuclear Power Plant. Calculation result shows that the transportation of fuel assemblies meets the

corresponding criticality safety requirements. In the calculation, eight criticality benchmark experiments for

Low-Enriched Uranium rods lattice from NUREG/CR-6361 of the U.S. NRC was selected, and was

validated calculation by SuperMC. Thereby, the result of criticality calculation for transportation of fuel

assemblies with SuperMC code becomes more reliable.

1 INTRODUCTION

Nuclear criticality safety is an important issue in the

storage and transportation of fissile materials.

Regulations for the Safe Transport of Radioactive

Material

(GB11806-2019) has clear requirements for

nuclear criticality safety in the transportation of

fissile materials, such as fuel assemblies, etc. The

IAEA SSR6 also provides a detailed introduction to

the criticality safety assessment, including the

criticality safety analysis model, method, calculation

and experiments (SSG, 2012).

The Monte Carlo method can better model the

geometric structure in the criticality safety analysis

and is widely used. However, when using the MC

program to calculate the criticality safety, various

uncertainties must be considered to give the bias of

the program, such as model size, fuel enrichment,

section data, calculation method, etc. (LI, 2019) In

addition, since it is necessary to model and write

input files when using MC program for criticality

safety calculation, the calculation results of the

program may vary from person to person. Therefore,

when using the MC program for critical safety

calculation, the bias of the program must be

determined first.

2 CALCULATION PROGRAM

AND DETERMINATION OF

SUBCRITICAL LIMIT

2.1 Calculation Program

This project is supported by Super Monte Carlo

Program for Nuclear and Radiation Simulation,

named SuperMC, which is developed by Institute of

Nuclear Energy Safety Technology, Chinese

Academy of Science/the FDS Team. SuperMC is a

general, intelligent, accurate and precise simulation

software system for the nuclear design and safety

evaluation of nuclear system (WU, 2009; WU, 2015).

2.2 Subcriticality Benchmark

Experiment and Simulation

Calculation

This paper selects a group of eight subcriticality

benchmark experiments in Criticality Benchmark

Guide for Light-Water-Reactor Fuel in

Transportation and Storage Packages (NUREG/

CR-6361) (Lichtenwalter, 1997) of the US Nuclear

Regulatory Commission, and Dissolution and

Storage Experimental Program with UO

Rods

(Manaranche, 1979), which are ANS33AL1,

ANS33AL3, ANS33EB1, ANS33EB2, ANS33EP1,

ANS33EP2, ANS33SLG and ANS33STY

respectively. Fuel assembly dimensions, fuel rod

characteristics, material parameters and criticality

data are described in detail in the literature.

Zhuang, D.

Criticality Safety Calculation and Analysis for NPP Transportation of Fuel Assemblies.

DOI: 10.5220/0012145900003562

In Proceedings of the 1st International Conference on Data Processing, Control and Simulation (ICDPCS 2023), pages 51-57

ISBN: 978-989-758-675-0

 2023 by SCITEPRESS – Science and Technology Publications, Lda. Under CC license (CC BY-NC-ND 4.0)

Table 1: Simulation Results of Subcriticality Benchmark Test by SuperMC.

Subcriticalit

benchmark experimen

eff

±σ

ANS33AL1 1.00071±0.00186

ANS33AL3 0.99958±0.00169

ANS33EB1 0.99835±0.00173

ANS33EB2 1.00496±0.00162

ANS33EP1 1.00121±0.00180

ANS33EP2 0.99939±0.00183

ANS33SLG 0.99769±0.00166

ANS33STY 0.99510±0.00169

For fuel rods, the bottom end plug is simulated

as an aluminum cylinder with a diameter of 0.94 cm

and the top plug is 1.3 cm. The spring is not

simulated but replaced by air. The stainless steel fuel

grid with a thickness of 0.25cm at the bottom is also

not considered.

Use SuperMC to calculate these criticality

benchmark models, and the calculation results are

listed in Table 1. The cross section data used in the

program calculation is mainly from the endf60

continuous energy neutron cross section library, and

250 iterations are used in the program calculation,

and the results of the first 50 iterations with poor

statistics are omitted. The average value of the

results of the last 200 iterations is used, and the

number of particles simulated in each iteration is

1000.

2.3 Analysis of Simulation Results

The simulation results of the eight critical

benchmark experiments using SuperMC program are

very close to the test results, and more close to the

critical value 1.0, with the maximum deviation of

0.496% and the minimum deviation of 0.042%. It

can be seen from Table 1 that the average deviation

b and the calculated standard deviation

of the

SuperMC program for the calculated value of the

critical benchmark simulation are:

00210.0

)(

≈

−



ieff

(1)

()

00272.0

)(

≈

−



ieff

(2)

Then consider various uncertainties of the

program (including fuel enrichment, model geometry,

material cross-section data, experimental data,

calculation methods, etc.), conservatively add 3

times of the standard deviation

to the above

average deviation b , that is, the bias of the

SuperMC program

3010.03 ≈+=

σσ

(3)

3 NUCLEAR CRITICALITY

SAFETY CALCULATION FOR

FUEL ASSEMBLY

TRANSPORTATION

3.1 Description of Transport Container

The shape of PWR fuel assembly transport container

is a tubular structure, mainly including two parts:

outer cylinder and inner cylinder. The structure is

shown in Figure 1.

(1) Outer cylinder

The outer cylinder consists of an upper cover and

a base. It is a "steel foam plastic steel" laminated

structure composed of an austenitic stainless steel

shell, an inner shell and a rigid polyurethane foam

plastic between two shells. The upper cover can be

opened or removed, and the base is fixed on a

forklift bracket. The base of the outer cylinder is

connected with the hinges on both sides by 24

hexagon bolts, and the upper cover is also connected

with the hinges by 24 bolts. When all 48 bolts are

fastened, the upper cover and base are fixed together.

When the 12 bolts connecting the upper cover and

hinge on one side are removed, the upper cover

connected through the hinge on the other side can be

just like a door. Acrylic glass fiber sealing gasket is

set between the joint surface of upper cover and base

to prevent rainwater from entering the package.

There is no pressure seal design between the

transport container package and the surrounding

environment, so there will be no differential pressure

in the package.

ICDPCS 2023 - The International Conference on Data Processing, Control and Simulation

Figure 1: Fuel Assembly Shipping Container Component Drawing.

The outer shell of the outer cylinder is used to bear

the structural strength of the container, and the lower

part is designed with a forklift platform to lift, stack

and tie down the container during transportation;

The foam plastic of the outer cylinder interlayer is

used for heat insulation and impact protection. Some

polyethylene blocks are attached to the inner shell of

the outer cylinder for critical safety; The two ends of

the outer cylinder are equipped with shock absorbers,

which are made of 20 pcf polyurethane foam

covered with stainless steel leather.

(2) Inner Cylinder

The inner cylinder is a rectangular box composed

of an aluminum V-shaped positioning plate, two

aluminum plate doors, bottom and top plates, and a

multi-point cam hinge and latch device. It plays the

role of protecting the built-in fuel assembly

structurally; The V-shaped locating plate and two

aluminum doors are connected by continuous (11

cams) hinges. One aluminum door is equipped with

a cam lock plate, which is locked with the other door

by turning at right angles; The fixing structure of the

top plate of the inner cylinder and the inner cylinder

wall is designed with flat head hexagon hole screws,

nuts and recessed seams. The bottom plate is also

fixed to the inner cylinder through screws, and can

be closed by connecting the nut and groove with the

inner shell door. The inner surface of the inner

cylinder is attached with a neutron absorption plate,

which is installed on the inner surface along the full

length of the four sides of the inner cylinder and is

fixed on the inner wall with threaded fasteners.

The inner cylinder is also a part of the container

restraint system, which can protect and constrain the

fuel assembly under different transportation

conditions. A rubber pad is set on the inner axial

position of the inner cylinder door to constrain the

lateral movement of the assembly. The top of the

inner cylinder is equipped with an adjustable thread

clamping component, which can provide the top

axial restraint for the fuel assembly or the rod tube.

When the fuel assembly is placed in it, additional

restraint devices are added to fix it.

3.2 Fuel Assembly

PWR new fuel assembly consists of the fuel rod and

fuel assembly skeleton arranged in a 17×17 square

shape. The fuel assembly skeleton comprises an

upper tube socket, a lower tube socket, a grid, a

guide tube and a neutron flux measuring tube. Each

fuel assembly consists of 289 grid cells, 24 of which

are occupied by the guide tube, one by the neutron

flux measuring tube, and the remaining 264 are

loaded into the fuel rod or the overall burnable

poison rod. The fuel rod is loaded into the fuel

assembly framework and clamped by the grid to

keep it at the specified axial and radial positions, and

the fuel rod is allowed to expand freely along the

axial direction. Sufficient clearance shall be reserved

between the end of fuel rod and upper and lower

tube sockets to compensate for different thermal

Criticality Safety Calculation and Analysis for NPP Transportation of Fuel Assemblies

expansion and irradiation growth between fuel rod

and guide tube. When the fuel assembly is loaded

into the core, it is assembled by the locating hole on

the lower tube socket and the locating pin on the

lower plate of the core to make it stand upright in the

core. When the upper core plate is in place, press

down the four groups of plate compression springs

of the upper tube socket to provide enough

compression force to position the fuel assembly on

the designated position of the core, and it will not

move upward under hydraulic scouring. The axial

load applied on the fuel assembly and the weight of

the fuel assembly are transferred to the lower plate

of the core through the guide tube and the lower tube

socket; The lateral load applied to the fuel assembly

is transferred to the core support structure through

the locating pins on the upper and lower core plates.

See Table 2 for specific parameters.

3.3 Criticality Calculation Model

The calculation model includes two models: single

package and infinite package array. Some

conservative assumptions were adopted in the

establishment of the critical safety calculation model,

and the following conditions were mainly considered

in the calculation:

(1) Under normal transportation conditions, there

will be no water in the transport container;

(2) Under accident conditions, all the spaces

inside the transport container are filled with water;

(3) The maximum 235U enrichment of fuel for

shipment is 5% and UO2 density is 10.96 g/cm3;

(4) A 30 cm thick water reflecting layer is falsely

set outside the transport container;

(5) The neutron poison plate is modeled

according to 75% of the actual density of boron

aluminum material, 1.942 g/cm3;

Table 2: Main parameters of fuel assembly.

component Main parameters

Fuel pellet

Material Uranium dioxide ceramics

Maximum

235

U enrichment(%) 4.80

Diameter (mm)

Outer diameter (mm) 8.19

inner diameter (mm) 3.94

Length(mm)

Normal pellet 9.83

Axial regeneration zone pellet 12.70

Pellet density(g/cm

) 10.41

Fuel rod

Cladding material ZIRLO

Rod length(mm) 4583.20

Outer diameter (mm) 9.50

Cladding wall thickness(mm) 0.57

Fuel

assembly

Arrangement form 17×17 square

Number of cells 289.00

Number of fuel rods 264.00

Fuel rod center distance (mm) 12.60

Transverse overall dimension (mm) 213.97×213.97

Total length of assembly (mm) 4798.70

Height of active segment (mm) 4267.20

Total weight of single component (kg) 794

Metal uranium weight of single component (kg) 541

ICDPCS 2023 - The International Conference on Data Processing, Control and Simulation

Table 3: Main materials and parameters.

Material Chemical composition and correspondin

atomic densit

(10

/cm

) densit

(

/cm

)

U-238( 2.32E-2) U-235(1.24E-3) O(4.89E-2) 10.96

Wate

H(6.68E-2) O(3.34E-2) 1.00

Boron aluminum

material

B-10(4.78E-3) B-11(1.94E-2) C(6.04E-3) AL(4.32E-2) 2.59

Pol

eth

lene C(3.95E-2) H(7.91E-2) 0.92

Foa

O(9.65E-4 H(9.57E-3) O(5.63E-3) N(2.76E-4) 0.16

Al 100% 2.70

Fe 100% 7.94

100% 6.56

igure 2: Model of single package under normal and accident

conditions.

Figure 3: Cross section of 151

acka

es arra

(6) The polyethylene material is modeled

according to 90% of the actual density, 0.828 g/cm3;

(7) In order to simplify the calculation model,

the positioning grid, upper and lower tube sockets

and damping frame are not considered in the fuel

assembly.

The main materials and corresponding

parameters used in the model are listed in Table 3.

Under normal and accident conditions established by

SuperMC, when the outermost layer of the package

is set as specular reflection. The section of the

calculation model is shown in Figure 3.

3.4 Calculation Results and Analysis

The calculation results are listed in Table 4.

Since the enrichment of

235

U in the simulated

fuel rod UO

pellet is 5%, the actual enrichment is

less than 4.8%, the UO

density selected for

calculation is conservative. It can be seen from Table

4 that under accident conditions, the keff value of the

infinite cargo package array first decreases, then

increases, and then decreases with the decrease of

water density. When the water density is 1.0 g/cm

the keff has a maximum value.

4 CONCLUSION

In this paper, the Monte Carlo software SuperMC is

used to analyze the critical safety performance of

new fuel transportation containers under normal

transportation conditions and transportation

accidents. Based on the characteristics of new fuel

and possible accident scenarios, eight benchmark

test cases that meet the requirements are selected for

verification. According to the statistical analysis of

benchmark test calculation results, the bias of the

SuperMC is 0.0103.

For a single package, water will not enter the

package under normal transportation conditions, and

the maximum k

eff

value after considering 3 times of

standard deviation and the standard deviation of the

critical calculation program is 0.22733; The

maximum k

eff

value of water entering the internal

clearance of the cargo package under accident

conditions is 0.85,704 after considering 3 times of

the standard deviation and the standard deviation of

the critical calculation program, which belongs to

subcritical.

Criticality Safety Calculation and Analysis for NPP Transportation of Fuel Assemblies

Table 4: The calculation case and results of criticality safety.

Calculation conditions

Water density

(g/cm

)

eff

σ k

eff

+3σ

eff

+3σ+

Normal conditions of single

package

1.0

0.2159 0.00039 0.21707 0.22733

Accident conditions of single

package

1.0

0.84414 0.00088 0.84678 0.85704

Normal conditions of unlimited

package array

1.0

0.27262 0.00049 0.27409 0.28435

Unlimited package array accident

conditions

1 0.89921 0.00088 0.90185 0.91211

0.995 0.89678 0.00084 0.8993 0.90956

0.99 0.89381 0.00084 0.89633 0.90659

0.98 0.88681 0.00079 0.88918 0.89944

0.97 0.88395 0.00078 0.88629 0.89655

0.96 0.87901 0.00086 0.88158 0.89184

0.95 0.87456 0.00095 0.87741 0.88767

0.94 0.86960 0.00078 0.87194 0.8822

0.93 0.86550 0.0008 0.8679 0.87816

0.92 0.86103 0.00082 0.86346 0.87372

0.91 0.85311 0.00079 0.85547 0.86573

0.9 0.84866 0.00087 0.85127 0.86153

0.8 0.79102 0.00083 0.79351 0.80377

0.7 0.72706 0.00084 0.72958 0.73984

0.6 0.65704 0.00085 0.65959 0.66985

0.5 0.58353 0.00079 0.5859 0.59616

0.4 0.50542 0.00071 0.50755 0.51781

0.3 0.43179 0.00057 0.4335 0.44376

0.2 0.37139 0.00053 0.37298 0.38324

0.1 0.33845 0.0005 0.33995 0.35021

Under normal transportation conditions of 151 cargo

bag arrays, the value of k

eff

is 0.28435, Under

accident conditions k

eff

is 0.91211, which both are

lower than 0.95，the limit of allowed by regulations.

Nuclear criticality safety of new fuel assembly

transportation activities is guaranteed.

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ICDPCS 2023 - The International Conference on Data Processing, Control and Simulation

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Criticality Safety Calculation and Analysis for NPP Transportation of Fuel Assemblies