Final Report under IAEA Research Agreement No.15951

Title of Project: Validation and Improvement of COPERNIC Code at Extended

Burnup

Title of Report

CALCULATIONS OF FISSION GAS RELEASE DURING

RAMP TESTS USING COPERNIC CODE

Host organization: Nuclear Fuel R&D Center, China Nuclear Power Technology

Research Institute (CNPRI)

Chief Scientific Investigator: Dr. Tong Liu

December, 2011

I

TABLE OF CONTENTS Executive summary...........................................................................................................................1 1. INTRODUCTION .....................................................................................................................2 2. CODE DESCRIPTION..............................................................................................................3 3. DATA DESCRIPTION OF THE RAMP TESTS .......................................................................4

3.1 Introduction......................................................................................................................4 3.2 Rod Design and Fabrication.............................................................................................5 3.3 General Assumption Used................................................................................................9

4. RESULTS AND DISCUSSIONS.............................................................................................13 4.1 Power Histories of the Ramp Tests ................................................................................13 4.2 Sensitivity Study of the FGR Models.............................................................................17 4.3 Application of COPERNIC to Real-time FGR ..............................................................22

5. CONCLUSIONS......................................................................................................................26 REFERENCES................................................................................................................................27

II

LIST OF TABLES Table 3-1 Summary of the central segment characteristics ...............................................................6 Table 3-2 Main characteristics of PK rods ........................................................................................7 Table 3-3 Main fuel rod component dimensions and rod characteristics ..........................................8 Table 4-1 Irradiation information of each rod.................................................................................14 Table 4-2 The diffusion parameters used in different COPERNIC FGR models............................17

LIST OF FIGURES Figure 3-1 Schematic diagram of the segment fuel rod stringer .......................................................6 Figure 3-2 Bump test power history of rod AN4 provided by the data file of AN4 document .......10 Figure 3-3 Bump test power history of rod AN4 provided by the pdf format document................11 Figure 3-4 Modified bump test power history of rod AN4 .............................................................11 Figure 3-5 The base irradiation history of rod group PK1 ..............................................................12 Figure 4-1 The power history of the ANF fuel rods........................................................................13 Figure 4-2 Power histories of PK1 series rods................................................................................15 Figure 4-3 Power histories of PK2 series rods................................................................................15 Figure 4-4 Power histories of PK4 series rods................................................................................16 Figure 4-5 Power histories of PK6 series rods................................................................................16 Figure 4-6 Comparison of the model predictions with the in-pile measurement data for ANF fuel........................................................................................................................................................18 Figure 4-7 Comparison of the model predictions with the experimental data for rods PK1, PK4 and PK6...........................................................................................................................................19 Figure 4-8 Comparison of the model predictions with the experimental data for rod PK2 ............19 Figure 4-9 The predicted versus measured fission gas release for ANF and PK fuel. ....................20 Figure 4-10 Comparison between predicted and measured FGR for rods AN1, AN3, AN4 and AN10...............................................................................................................................................22 Figure 4-11 Comparison between predicted and measured fuel centerline temperature of AN3 and AN10...............................................................................................................................................23 Figure 4-12 Measured fission gas release fraction of AN4 during temperature change .................24

1

Executive summary

The report performed under IAEA research contract No.15951 describes the

results of fuel performance evaluation of LWR fuel rods operated at ramp conditions

using the COPERNIC code developed by AREVA. The experimental data from the

Third Risø Fission Gas Project and the Studsvik SUPER-RAMP Project presented in

the IFPE database of the OECD/NEA has been utilized for assessing the code itself

during simulation of fission gas release (FGR).

Standard code models for LWR fuel were used in simulations with parameters

set properly in accordance with relevant test reports. With the help of data adjustment,

the input power histories are restructured to fit the real ones, so as to ensure the

validity of FGR prediction. The results obtained by COPERNIC show that different

models lead to diverse predictions and discrepancies. By comparison, the COPERNIC

V2.2 model (95% Upper bound) is selected as the standard FGR model in this report

and the FGR phenomenon is properly simulated by the code. To interpret the large

discrepancies of some certain PK rods, the burst effect of FGR which is taken into

consideration in COPERNIC is described and the influence of the input power

histories is extrapolated.

In addition, the real-time tracking capability of COPERNIC is tested against

experimental data. In the process of investigation, two main dominant factors

influencing the measured gas release rate are described and different mechanisms are

analyzed. With the limited predicting capacity, accurate predictions cannot be carried

out on abrupt changes of FGR during ramp tests by COPERNIC and improvements

may be necessary to some relevant models.

2

1. INTRODUCTION

The incentive of developing extended burnup fuel is to maintain the

competitiveness of nuclear energy. With increasing demand for economics, the

industry needs to reduce maintenance and fuel cycle costs, while enhancing safety

features. Extended burnup is one of the methods applied to meet these objectives.

There are a number of issues that need resolution to be able to successfully implement

extended burnup.

The performance of the critical fuel components is the result of a complex

interaction of a large number of variables that challenge the evaluation of the

mechanisms in progress and the prediction of their behavior at extended and more

severe conditions (e.g. higher burnup). In order to guarantee the rod integrity and

operation safety, a number of fuel rod performance codes are developed, such as

URANUS[1,2], TRANSURANUS[3,4], FRAPCON-3[5], PAD, COPERNIC[6],

FEMAXI-V[7], FALCON[8] and so on. They are widely used by research centers as

well as nuclear safety authorities to assess specific fuel rod design and offer a good

tool for evaluating a limited number of variables and mechanisms and have provided

some valuable data.

The International Atomic Energy Agency (IAEA) has initiated a new

Coordinated Research Project on Fuel Modeling at Extended Burnup (FUMEX-III) in

which all principal investigators exchange information and experience in testing and

developing their fuel performance codes against data and cases provided by the IAEA

and OECD/NEA. As a participant, the present work of CNPRI is focus on the

validation of the COPERNIC prediction capability in simulating burnup, cladding

waterside corrosion, cladding creep, fuel temperature, FGR, fuel axial elongation and

rod internal pressure for LWR fuel rods operated at steady-state or transient

conditions. Based on the experimental data from the Third Risø Fission Gas Project

and the Studsvik SUPER-RAMP Project, numerous rods are tested and certain

mechanisms are investigated.

3

2. CODE DESCRIPTION

The COPERNIC code is an advanced fuel rod design tool for evaluating fuel rod

thermal-mechanical performance. It incorporates several state-of-the-art models

developed by FRAMATOME and borrows the European TRANSURANUS code

architecture which is based on a mathematical framework of the heat transfer and

mechanical models. The main objectives of COPERNIC are to describe the behavior

of the fuel accurately and provide credible data for fuel rod design[9].

Using the fuel rod fabrication and irradiation data, COPERNIC evaluates the fuel

performances under a continuously varying power history by introducing the time

discretization and spacial discretization method. Each macro time step is further

divided into additional micro ones by the code itself during the time discretization

process. As for the spacial discretization of rod geometry, the fuel rod is expressed in

sequential axial slices or sections. In each slice, there are several coarse concentric

rings (iso-volume discretization) which are subdivided into many fine ones.

During a micro time step, the time-dependent quantities (linear power, fast flux,

thermohydraulics data) are considered as constant in each axial slice, the young's

modulus and the poisson's ratio are assumed constant in each coarse ring (cladding or

pellet) while the thermal conductivity is considered as constant in each fine ring. The

fuel temperature, fission gas release, densification, swelling, cladding stress, strain

and cladding corrosion are calculated separately for each axial slice, and the effects

are integrated to obtain the overall rod void volume and internal pressure. The code

runs iteratively due to the complex interaction of individual model until convergence

occurs[10].

The physical models in COPERNIC code were developed from the latest

knowledge of fuel in-pile phenomena based on experimental data base. The database

is obtained from different rod design and under different operation conditions.

4

3. DATA DESCRIPTION OF THE RAMP TESTS

3.1 Introduction

The data sets used in this report come from two ramp (or bump) tests. During

these tests, thermomechanical fuel behaviour such as fission gas release and fuel

microstructural changes is investigated with the help of COPERNIC code. Brief

introduction of these projects is conducted as follows:

1) the third Risø fission gas project

The Risø National Laboratory in Denmark have carried out three irradiation

programs of slow ramp and hold tests, so called ‘bump test’ to investigate fission gas

release and fuel microstructural changes. The Third Risø Fission Gas Project was

organized as joint international project with sponsorship from a number of

organizations in Europe, Japan and the USA. The project was executed by Risø

between 1986 and 1990.

Two Risø projects have already generated comprehensive data sets with high

burnup fuel. Part of these data, such as transient fission gas release as a function of

time, is described in detail. In addition, detailed characterizations are provided of

local fission product distributions and fuel structures resulting from the power

transients.

In the Third Risø Fission Gas Project, previously irradiated power reactor fuel

was refabricated into test fuel pins that were instrumented with pressure transducers

and fuel thermocouples. This enabled continuous monitoring of fission gas release

and fuel temperature during power transients. As in the previous projects, extensive

hotcell examinations before and after the transient testing provided detailed data on

local fission product distribution and fuel structure.

The project has provided valuable data derived from the PIE process. In the

project, extensive hotcell examinations were performed to characterize the tested fuels

after base irradiation and after bump testing. Some examinations were carried out at

the Institute for Transuranium Elements in Karlsruhe, including electron probe micro

5

analysis as well as optical and electron microscopy.

There were three kinds of fuel rods used in the project: IFA-161 irradiated in the

Halden BWR with burnups from 13 to 46 MWd/kgUO2, GE BWR fuel irradiated in

Quad Cities 1 and Millstone 1 with burnups from 20 to 40 MWd/kgUO2 and ANF

PWR fuel irradiated in Biblis A to 38 MWd/kgUO2. In this report, only the ANF fuel

is discussed.

2) the Studsvik SUPER-RAMP project

The Studsvik SUPER-RAMP Project investigated the failure propensity of

typical LWR fuel in the form of test rods when subjected to power ramps, after base

irradiation to high burn-up[11].

The Project power ramped 28 individual PWR rods and 16 BWR rods. The 28

fuel rods of PWR subprogram consisted of 6 groups of rods with variations in design

and material parameters. 4 groups among them, PK1, PK2, PK4 and PK6 are

simulated with COEOPRNIC code. Group PK1, PK2, PK4, PK6 consisting of 19 rods

(18 of them were modeled in this activity) were delivered by Kraftwerk Union

AG/Combusting Engineer (KWU/CE). The PK rods were base irradiated in power

reactor environment KK Obrigheim at time averaged heat ratings mainly in the range

14-26 kW/m to peak burn-ups in the range 33-45 MWd/kgU and were subsequently

ramp tested in the research reactor R2 at Studsvik, Sweden.

After irradiation, all rods underwent a thorough examination program,

comprising characterisation prior to base irradiation, examination between base and

ramp irradiation and examination after ramp irradiation.

3.2 Rod Design and Fabrication

1) the third Risø fission gas project

Fuel rod stringers with zirconium lined barrier cladding rod segments were

irradiated as part of the Biblis A XN-3 fuel reload. These rods were fabricated by the

Advanced Nuclear Fuels Corporation. Each rod has six segments, but only the central

four segments are characterized and have zircaloy-4/zirconium barrier cladding [12].

6

Each of the central segments is ~653.4mm long. Every segment contains a fuel stack

of enriched pellet, one natural UO2 depressor pellet and one insulator disk on each

end of the fuel stack, plenum spring and zircaloy-4 end fittings that are welded to the

cladding. The characterized segments contain solid core dished pellets. The nominal

cladding-pellet gap is 0.19mm. The schematic diagram of the segment fuel rod

stringer is shown in Figure 3-1 and the summary of the central segment characteristics

is listed in Table 3-1

Figure 3-1 Schematic diagram of the segment fuel rod stringer

Table 3-1 Summary of the central segment characteristics

Parameter description Mean value Standard deviation

Segment

Length(mm) 653.5

Active fuel length(mm) 541.8

Plenum length(mm) 61.0

Diametral gap(mm) 0.19

Fill pressure, He(MN/m2) 2.52

Pellet

Process ADU

Outer diameter(mm) 9.053 0.005

Length(mm) 6.909 0.071

Enrichment(w/o) 2.95

Density(g/cm3) 10.274

T.D.(%) 93.74 0.434

7

Resintered density(%) 93.92

Dish volume(%) 0.97 0.07

Grain size(μm) 6.0

Open porosity(%) 0.07 0.05

Cladding

Material Zr-4, Zirconium lined

Treatment Coldworked-Stress relieved

Manufacturer Sandvik Special Metals

Outer diameter(mm) 10.82

Inner diameter(mm) 9.25

Thickness(mm) 0.787

Barrier thickness(mm) 0.057

CSR-R value 1.797

Burst hoop stress(MN/m2) 887.0

0.2% Yield strength(MN/m2) 535

Ultimate strength(MN/m2) 728

Elongation(%) 19.5

ASTM grain size 12.0

Hydride orientation <0.1

2) the Studsvik SUPER-RAMP project

18 PWR fuel rods of PK rods were simulated by COPERNIC code. The rods

formed four groups containing 4 to 5 rods of identical design. The main

characteristics of PK rods are shown in Table 3-2.

Table 3-2 Main characteristics of PK rods

Group Type Number of Rods Cycles Rod Average Burnup

MWd/tU

PK1 Standard 4 KWO Cycle 9, 10, 11 35

PK2 Standard 5 KWO Cycle 8, 9, 10, 11 44

8

PK4 Standard 4 KWO Cycle 8, 9, 10 33

PK6 Large grain 5 KWO Cycle 8, 9, 10 36

The main fuel rod component dimensions and rod characteristics are shown in

Table 3-3.

Table 3-3 Main fuel rod component dimensions and rod characteristics

PK1 PK2 PK4 PK6

Pellet type Standard Standard 4% Gd2O3 Large grain

He fill pressure bar

22.5 22.5 22.5 22.5

UO2 stack length mm

311 318 314 315

Overall rod length mm

388 390 390 390

Cladding OD mm

10.67 10.75 10.77 10.74

Cladding ID mm

9.31 9.38 9.28 9.29

Diametral gap μm

197 145 169 146

Pellet fabrication AUC AUC AUC DC Enrichment wt% 235U

3.2 3.21 3.19 2.99

Gd content wt% Gd2O3 in UO2

4.09±0.05

Pellet OD mm

9.11 9.14 9.11 9.14

Pellet L/D 1.25 1.24 1.19 1.21 Increase after densification test

% 0.4 0.7 0.7 none

Average grain size μm

6 5.5 5.5 22

Final density g/cm3

10.36 ±0.012

10.34 ±0.019

10.295 ±0.01

10.42 ±0.04

Pellet OD mm

9.110 ±0.006

9.138 ±0.003

9.113 ±0.004

9.144 ±0.008

Pellet length mm

11.35 ±0.60

11.34 ±0.40

11.799 ±0.65

11.1 ±0.90

O/U 2.00±0.01 2.00±0.01 2.037±0.01 2.00±0.01

9

Open porosity %

50±5 55.8±3.1 2.1±0.7 42.8±7.8

IFPE database also provide detailed information of some of the rod

characteristics for fuel rods of PWR subprogram which may differ from the table

above slightly due to fabrication tolerance.

3.3 General Assumption Used

The following assumptions were made to obtain the necessary data for

evaluation of fuel rod performances.

Fuel pellet and fuel rod data:

– Total number of plenum spring turns, spring external diameter and spring wire

diameter are substituted by a multiplicative factor on upper plenum volume.

Then, the multiplicative factor is adjusted to fit the rod void volume. This

method is applied to all of the tested rods.

– The fill gas temperature of 20℃ is used.

– Dish depth and pellet dish spherical radius, which are used to calculate the

dish volume, are replaced by the ratio between the dish and pellet volume.

– The pellet chip volume used to calculate void volume inside the rod was set as

0.3% for all the rods in this project.

– The large gain pellets of PK6 rods were manufactured by DC process, which

is a dry conversion procedure. Since COPERNIC code dose not have a DC

pellets model, the pellets of PK rods are regarded as IDR (which is also a dry

conversion procedure) pellets in calculation.

– The COPERNIC code employed a linear interpolation method to gain the

power histories for calculation between two adjacent power steps inputted by

users. Thus, the bump test power history of rod AN4 which is provided by the

data file of Risø is treated as Figure 3-2. Whereas, it is not consistent with the

power history (illustrated in Figure 3-3) provided by the pdf format document

which apparently has several step like power level changes. Therefore, the

10

power history of Figure 3-2 is modified to construct the power history in

Figure 3-3. Several steps are added in order to interpolate the history desired

and the modified power history is showed in Figure 3-4. The modified power

history is almost the same with the history from pdf document.

For PK rods, the same problem occurs. Similarly, the history of PK rods is

modified, but the image of PK rods power history in SuperRamp document

(illustrated in Figure 3-5) is not clear. Therefore, the adjusted power history of

PK rods may have a deviation from the actual experimental power history.

– The power history data can be obtained from explanatory files. The coolant

pressure is assumed to have no change with time.

0 10 20 30 40 50 60 70 800

10

20

30

40 Average linear heating rate

Ave

rage

line

ar h

eatin

g ra

te (k

W/m

)

Time (hours)

Figure 3-2 Bump test power history of rod AN4 provided by the data file of AN4

document

11

Figure 3-3 Bump test power history of rod AN4 provided by the pdf format

document

0 10 20 30 40 50 60 70 800

10

20

30

40

Ave

rage

line

ar h

eatin

g ra

te (k

W/m

)

Time (hours)

Average linear heating rate

Figure 3-4 Modified bump test power history of rod AN4

12

Figure 3-5 The base irradiation history of rod group PK1

13

4. RESULTS AND DISCUSSIONS

4.1 Power Histories of the Ramp Tests

1) the third Risø fission gas project

The segmented stringers with ANF rods were irradiated in the Biblis A reactor,

during cycles 7, 8, 9 and 10. The power and fast flux distribution for the assemblies (4

radial and 18 axial nodes per assembly) were generated through a synthesis of 3D full

core calculation. The rod powers were generated from a fitted surface of the 4 radial

LHGR’s for the assembly and a local power factor. ANF neutronics calculations of

relative assembly power distributions have been extensively compared to measured

values in many PWR cores. Agreement is typically within ±5%, although larger

discrepancies have been observed. Comparisons on a rod-by-rod basis are not

available, however agreement within ±10% at any given point on time would be

expected and on an integrated basis (i.e. at discharge) agreement within ±5% would

be expected.

0 20 40 60 800

200

400

600

Line

ar P

ower

Rat

e (W

/cm

)

Time (h)

AN1 AN2 AN3 AN4 AN8 AN10 AN11

Figure 4-1 Power history of the ANF fuel rods.

Then these test fuel rods were transferred to the DR3 reactor where transient

tests were carried out in water-cooled rigs. The total test time was in the range of 5h

14

to 3d. The test power was determined from calorimetric measurements during the test.

The axial power shape was obtained from 140La axial gamma scans (T/2=40.2h, and

12.79d for the 140Ba pre cursor). Axial power variations were small for the

refabricated tests, due to the short stack length. Figure 4-1 shows the power history of

the ANF fuel.

2) the Studsvik SUPER-RAMP project

The irradiation information of PK rod is shown in table 4-1.

Table 4-1 Irradiation information of each rod

Type Cycle Burnup

MWd/kgUrod

Burnup MWd/kgU

Ramp peak power kW/m

Peak power duration

h 1 35.4 37.2 11 2 35.6 39.5 11 3 35.2 42.6 11

PK1 Standard 9, 10, 11 33-36

4 33.1 42.6 11 1 45.2 36.8 11 2 45.1 41.3 11 3 44.6 43.9 11 4 41.4 39.5 very short

PK2 Standard 8, 9, 10, 11 41-45

S 43.4 39.5 11 1 33.7 34 12 2 33.8 38.9 12 3 33.6 42.8 36

PK4 4wt% Gd2O3

8, 9, 10 33-34

S 32.5 36.2 11 1 36.8 35.7 very short 2 36.8 35.7 very short 3 36.5 38.4 11

PK6 Large grain 8, 9, 10 34-37

S 35.9 36.6 11

The manufacturing dimensions of PK series rods are similar. PK6 is

characterized by large grain size which reduces the amount of released fission gas,

and PK4 is Gd-doped UO2 fuel. The PK1 and PK2 are both standard design, while

PK2 was extended to 4 cycle irradiation. The power histories of each group are

depicted in Figure 4-2 through 4-5.

15

0 5000 10000 15000 200000

10

20

30

40

Line

ar H

eat R

ate

(kW

/m)

Time (h)

PK1-1 PK1-2 PK1-3 PK1-4

Figure 4-2 Power histories of PK1 series rods

0 5000 10000 15000 20000 25000 300000

10

20

30

40

Line

ar H

eat R

ate

(kW

/m)

Time (h)

PK2-1 PK2-2 PK2-3 PK2-4 PK2-S

Figure 4-3 Power histories of PK2 series rods

16

0 5000 10000 15000 200000

10

20

30

40Li

near

Hea

t Rat

e (k

W/m

)

Time (h)

PK4-1 PK4-2 PK4-3 PK4-S

Figure 4-4 Power histories of PK4 series rods

0 5000 10000 15000 200000

10

20

30

40

Line

ar H

eat R

ate

(kW

/m)

Time (h)

PK6-1 PK6-2 PK6-3

Figure 4-5 Power histories of PK6 series rods

17

The power histories of PK rods are similar without large fluctuation due to the

same operation condition in same reactor. The 3 irradiation cycles of rods PK1, PK4

and PK6, and 4 irradiation cycles can be recognized in these figures, and the power of

first cycle of PK4 is lower than PK1 and PK6 because of the dope effect of Gd2O3.

4.2 Sensitivity Study of the FGR Models

Analogical FGR models are incorporated in the COPERNIC code, the main

differences lie in certain parameters which may be influential to simulation results. In

order to derive the constituent parameters and to validate the current FGR models,

in-pile measurement data was set available in the COPERNIC code while changing

the models. In COPERNIC, attention was paid to models from different versions, and

the parameters used in them are listed in Table 4-2 (without parameters of

TRANSURANUS model).

Table 4-2 The diffusion parameters used in different COPERNIC FGR models

FGR Models D01

(m2/s)

D02

(m2/s)

D03

(m2/s)

T01

(K)

T02

(K)

L2b

(m2/s)

COPERNIC V2.2 model (95%

Upper bound) 7.6×10-10 1.77×10-15 2×10-21 35000 13800 10-15

COPERNIC V2.1 model 3.9×10-6 1.77×10-15 2×10-21 45275 13800 10-15

COPERNIC V1 model 3.9×10-6 1.77×10-15 4×10-21 45275 13800 10-15

1) the third Risø fission gas project

The fraction of FGR during the bump test is derived from pressure measurement

under hot condition. Using the FGR models built in COPERNIC, each model

prediction is checked against the measurement data and the other model predictions.

As shown in Figure 4-6, the prediction of COPERNIC V2.2 model (95% Upper

bound) is in good agreement with the experimental measurement data rather than

those of other models built in the code.

18

40.0 40.5 41.0 41.5 42.00

5

10

15

20

25

30

35

40

45 Measured data COPERNIC V2.2 model(95% Upper Bound) COPERNIC V2.1 model COPERNIC V1 model TRANSURANUS model

Fiss

ion

gas

rele

ase

(%)

Pellet average burnup (MWd/kgU) Figure 4-6 Comparison of the model predictions with the in-pile measurement data for

ANF fuel rods

2) the Studsvik SUPER-RAMP project

To determine appropriate FGR model in the predictions of the PK rods, several

results of model predictions are compared with each other in Figure 4-7 and Figure

4-8, respectively. Among the four models, COPERNIC V1 model and COPERNIC

V2.2 model (95% Upper bound) are more precise with respect to FGR predictions of

PK rods. Combined with the Risø results, the V2.2 model is conceived more

reasonable to reflect FGR phenomenon during ramp conditions with burnup in the

range of 32~46 MWd/kgU.

19

32 33 34 35 36 370

20

40

60

Measured data

COPERNIC V2.2 model(95% Upper Bound)

COPERNIC V2.1 model

COPERNIC V1 model

TRANSURANUS model

Fiss

ion

gas

rele

ase

(%)

Pellet average burnup (MWd/kgU) Figure 4-7 Comparison of the model predictions with the experimental data for rods

PK1, PK4 and PK6

41 42 43 44 45 460

10

20

30

40

50

Measured data COPERNIC V2.2 model(95% Upper Bound) COPERNIC V2.1 model COPERNIC V1 model TRANSURANUS model

Fiss

ion

gas

rele

ase

(%)

Pellet average burnup (MWd/kgU)

Figure 4-8 Comparison of the model predictions with the experimental data for rod

PK2

20

In this study, influence of diffusion parameters can be considered dominant in

FGR simulations. According to the previous work mentioned above, COPERNIC

V2.2 model (95% Upper bound) is chosen as standard model to predict FGR

phenomenon of ANF and PK rods. The predicted versus measured results is presented

in Figure 4-9.

1 10 1001

10

100

ANF PK1 PK2 PK4 PK6

Pre

dict

ed F

issi

on G

as R

elea

se (%

)

Measured Fission Gas Release (%) Figure 4-9 The predicted versus measured fission gas release for ANF and PK fuel.

Neglecting the influences of the initial fill gas, the COPERNIC predicted results

fit the measured data well at EOL of these ramp tests. The relative errors obtained

from the results lie within -5%~50%, except for some certain rods.

The fission gas release model integrated in COPERNIC code takes the burst

effect into consideration. When a fuel rod undergoes a transient condition, the

evolution of the gas concentration is in part due to diffusion and to burst release. The

burst release is modeled as an exponential decrease of the grain boundary

concentration with a relaxation time proportional to the diffusion coefficient.

Reference [13] presented the equation of the gas concentration evolution at time t

during transient condition:

21

( )[ ] ⎟⎟⎠

⎞⎜⎜⎝

⎛ −−−−−=

6

02000 exp1

Ctt

aDcttFcc BTR

where,

0t : Start time of the transient

0c : Total concentration retained in the fuel at time 0t .

0Bc : Grain boundary concentration at time 0t

D : Diffusion coefficient

a : Diameter of the grain

6C : Model parameter fitted from experimental observations.

The first term on the right side of the equation reflects the gas release due to

diffusion, and the second term represents the concentration attenuation caused by the

burst effect. Above all, the equation explains the excess release of gas in the gain

boundary during transient. Then, in the subsequent normal condition, the gain

boundary will be filled with gas until the saturation is reached again.

The calculation strategy of FGR in transient is different from that of the steady

state condition in COPERNIC code. Thus, the FGR calculation of COPERNIC

depends greatly on the accuracy of the input power history. As discussed in chapter

3.3, the power histories form different sources of SuperRamp documents are not

consistent. This deviation may cause the calculation inaccuracy PK rods in Figure 4-9.

22

4.3 Application of COPERNIC to Real-time FGR

To make good use of the Risø experimental data, the real-time FGR rate is

employed in extended research on the real-time tracking capability of COPERNIC.

Since there are only four ANF rods which were mounted on the pressure

transducers in the refabrication process, the measured FGR is compared with the

COPERNIC predicted one in Figure 4-10.

0 20 40 60 800

10

20

30

40

50

AN1 Measured AN1 Predicted

FGR

(%)

Time (h)0 20 40 60 80

0

10

20

30

40

50

AN3 Measured AN3 Predicted

FGR

(%)

Time (h)

0 20 40 60 800

10

20

30

40

50

AN4 Measured AN4 Predicted

FGR

(%)

Time (h)0 20 40 60 80

0

10

20

30

40

50

AN10 Measured AN10 Predicted

FGR

(%)

Time (h)

Figure 4-10 Comparison between predicted and measured FGR for rods AN1, AN3,

AN4 and AN10.

23

0 20 40 60 800

300

600

900

1200

1500

1800

Cen

terli

ne T

empe

ratu

re (℃

)

Time (h)

AN10 Measured AN10 Predicted AN3 Measured AN3 Predicted

Figure 4-11 Comparison between predicted and measured fuel centerline temperature

of AN3 and AN10.

The measured FGR jumps up rapidly at a certain time which the fuel power and

temperature undergoes noticeable changes (see Figure 4-11).

Firstly, more pellet fragmentation due to excessive stress will occur during

temperature transient. The pellet fragmentation produced in temperature transient will

increase the free surface of fuel pellet by creating more cracks. The interlinked grain

boundaries which have not formed a tunnel to the free surface may connect with the

crack surface, then the fission gas of all the interlinked grains will release.

Secondly, the pellet and cladding may contact more tightly at high temperature

due to the thermal expansion of pellet, then, the tight contact may block the fission

gas convection and cause the maldistribution of fuel rod internal pressure. As can be

seen from the technical report, the measured FGR is derived from local pressure and

the local pressure is proportional to the local fission gas release, that is, proportional

to the local power and temperature. Probably, the pressure in the upper part of the rod

where the pressure transducer is installed is lower than the pressure of the middle part,

24

or lower than the average pressure of the rod. When the temperature of pellet

descends, the contact force between pellet and cladding may decreases, the internal

pressure of the fuel rod will be balanced, that, a part of the gas in the middle section

of the rod will flow into the upper section to increase the pressure rapidly.

The difference between the two possible reasons which cause the steep rise in the

fission gas release is: the former phenomenon may be dominant when the temperature

increases, while the later one takes place during the temperature decline. Figure 4-12

illustrates the measured fission gas release fraction of AN4 during the temperature

change. The fission gas release fraction increases quickly as soon as the temperature

falls down. Then, during the process of temperature raise, the fission gas release

fraction continues to increase, but in a much slower speed. It can be concluded that,

the fuel fragmentation and the tight pellet and cladding contact during high power

level may cause the sharp increase of the measured FGR fraction; however the

reduced pellet and cladding contact pressure may contribute more.

48.0 48.5 49.0 49.5 50.0 50.5 51.0800

1000

1200

1400

1600 Fuel Rod Center Hole Temperature FGR Fraction

Time (hours)

Fuel

Rod

Cen

ter H

ole

Tem

pera

ture

(o C)

20

25

30

35

40

45

FGR

Fraction (%)

Figure 4-12 Measured fission gas release fraction of AN4 during temperature change

Skimming over this section, conclusions can be drawn that FGR predictions by

25

COPERNIC only consider the contributions of Xe and Kr, regardless of the influence

of the initial fill gas. In addition, with the limited predicting capacity, real-time

tracking cannot be carried out on abrupt changes of FGR during ramp tests by

COPERNIC, improvement may be necessary to some relevant models taking into

account the gas block effect of pellet and cladding contact and the fission gas release

caused by pellet fragmentation.

26

5. CONCLUSIONS

The feasibility of the COPERNIC code for predicting fuel rod performance

during ramp conditions has been assessed. In the investigation, the tested rods are

base irradiated to high burn-up and then subjected to ramp tests[11][12].The ramp

approach to the bump terminal level (usually around 350-400 W/cm) was mostly

made in small steps to obtain extensive measurement data. The hold time at most of

these steps was several hours. In the process of simulations, several input parameters

are adjusted in accordance with different fuel rod geometric dimensions or power

histories.

To be comprehensive, different FGR models are compared with each other and

tested in the simulations. In order to derive the constituent parameters and to validate

the current FGR models, in-pile measurement data was set available in the

COPERNIC code while attention was paid to models from different versions, and the

parameters used in them are described in the report. With changes of principal model

parameters, the prediction results alter a great deal. By comparison, the COPERNIC

V2.2 model (95% Upper bound) is considered as a more precise one and the

predictions are in good agreement with the experimental data.

According to the results of the research on the real-time tracking capability of

COPERNIC, the measured FGR jumps up rapidly at a certain time which the fuel

power and temperature undergoes noticeable changes. The predicted results are

compared with the measured ones and two possible reasons are proposed to explain

the discrepancies between them. Accordingly, the models are considered to need

improvements on abrupt changes of FGR, at least during ramp tests.

27

REFERENCES

[1] K.Lassmann, A computer programme for the thermal and mechanical analysis of

the fuel rods in a nuclear reactor, Nuclear Engineering and Design, 45 (1978)

325-342.

[2] K.Lassmann, An advanced method for transient temperature calculation in fuel

element structural analysis, Nuclear Technology, 60 (1983) 406-419.

[3] K.Lassmann, TRANSURANUS: a fuel rod analysis code ready for use, Journal

of Nuclear Materials, 188 (1992) 295-302.

[4] K.Lassmann, H.Blank, Modelling of fuel rod behaviour and recent advances of

the TRANSURANUS code, Nuclear Engineering and Design, 106 (1988)

291-313.

[5] D.Lanning, C.Beyer and C.Painter, FRAPCON-3: Modification to fuel rod

material properties and performancer models for high-burnup application,

NUREG/CR-6534, PNNL-11513,1997.

[6] Ch.Garnier, et al., Recent modeling features in the COPERNIC3 AREVA NP fuel

rod performance code, 2006 International meeting on LWR fuel performance,

22-26 October 2006.

[7] M.Suzuki, Light water reactor fuel analysis code FEMAXI-V (ver. 1),

JAERI-Data/Code 2000-030,2000.

[8] Y. Rashid, R.Dunhman and R.Montgomery, FALCON MOD01: Fuel analysis and

licensing code-New, Volume1: Theoretical and numerical bases, ANA-04-0666

Report, EPRI, Palto Alto, CA, and ANATECH Corp., San Diego, CA, 2004.

[9] H.Segura, COPERNIC - FUEL ROD DESIGN CODE USER MANUAL,

FF/DC/01858, FRAMATOME ANP，March 2005.

[10] J.L.Jacoud, Ph.Vesco, Description and qualification of the COPERNIC/

TRANSURANUS (UPDATE OF MAY 2000) FUEL ROD DESIGN CODE,

TF.JC/DC/1556, Rev. A, May 2000.

[11] Seved Djurle, The Super-Ramp Project, STUDSVIK-STSR-32, December 1984

[12] K.R.Merckx, THE THIRD RISØ FISSION GAS PROJECT –

28

CHARACTERIZATION, BASE IRRADIATION HISTORIES AND POOL

INSPECTION REPORTS OF ANF BARRIER CLADDING SEGMENTS,

RISØ-FGP3-ANF,Pt.1, September 1990.

[13] L.C. Bernard, J.l. Jacoud, P. Vesco, An efficient model for the analysis of fission

gas release, Journal of Nuclear Material, 302 (2002) 125-134.