NUREG/IA-0199

IPSN 01-16 

NSI RRC 2241 

International 

Agreement Report 

Mechanical Properties of Unirradiated 

and Irradiated Zr-1% Nb Cladding 

Procedures and Results of Low Temperature Biaxial Burst 

Tests and Axial Tensile Tests 

Prepared by 

E. Kaplar, L. Yegorova, K. Lioutov, A. Konobeyev, N. Jouravkova 

Nuclear Safety Institute of Russian Research Centre "Kurchatov Institute" 

Kurchatov Square 1, Moscow 123182 Russian Federation 

and 

V. Smimov, A. Goryachev, V. Prokhorov, 0. Makarov, S. Yeremin, A. Svyatkin 

State Research Centre "Research Institute of Atomic Reactors" 

Dimitrovgrad 433510, Russian Federation 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

April 2001 

Prepared for 

U.S. Nuclear Regulatory Commission, Institute for Protection and Nuclear Safety (France), and 

Ministry of Science and Technologies (Russian Federation) 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0199 

IPSN 01-16 

NSI RRC 2241 

International 

Agreement Report 

Mechanical Properties of Unirradiated 

and Irradiated Zr-i % Nb Cladding 

Procedures and Results of Low Temperature Biaxial Burst 

Tests and Axial Tensile Tests 

Prepared by 

E. Kaplar, L. Yegorova, K. Lioutov, A. Konobeyev, N. Jouravkova 

Nuclear Safety Institute of Russian Research Centre "Kurcbatov Institute" 

Kurchatov Square 1, Moscow 123182 Russian Federation 

and 

V. Smimov, A. Goryacbev, V. Prokhorov, 0. Makarov, S. Yeremin, A. Svyatkin 

State Research Centre "Research Institute of Atomic Reactors" 

Dimitrovgrad 433510, Russian Federation 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

April 2001 

Prepared for 

U.S. Nuclear Regulatory Commission, Institute for Protection and Nuclear Safety (France), and 

Ministry of Science and Technologies (Russian Federation) 

Published by 

U.S. Nuclear Regulatory Commission
ABSTRACT

This report contains the description of the test conditions, procedures, and main results of the following 

types of mechanical tests of unirradiated and irradiated Zr-1 %Nb cladding: 

"* biaxial burst tests with liquid pressurized tube specimens in the temperature range 293-723 K; 

"* uniaxial tensile tests in axial direction with tube specimens in the temperature range 293-1200 K. 

Results of these tests are compared with test results obtained in 1997-1999 using simple ring specimens 

(293-1200 K) and gas pressurized tube specimens (900-1400 K).

iii
TABLE OF CONTENTS 

Page 

1. EXECUTIVE SUMM ARY ............................................................................................................................... 1.1 

1.1. Introduction .......................................................................................................................................... 1.1 

1.2. Program of tests .................................................................................................................................... 1.1 

1.3. Selection of test param eters and ranges ............................................................................................... 1.4 

2. CONDITIONS, PROCEDURES AND RESULTS OF BIAXIAL BURST TESTS ................................................... 2.1 

2.1. Test facility for biaxial burst tests ........................................................................................................ 2.1 

2.2. Test procedures and measured param eters .......................................................................................... 2.4 

2.3. Results of biaxial burst tests ................................................................................................................. 2.7 

3. CONDITION, PROCEDURES AND RESULTS OF AXIAL TENSILE TESTS .......................... 3.1 

3.1. Test procedure and specimen design ................................................................................................... 3.1 

3.2. Results of axial tensile tests ................................................................................................................. 3.5 

4. CONCLUSIONS ............................................................................................................................................ 4.1

V
LIST OF FIGURES 

Page 

Fig. 1.1. Structure of the program to obtain generalized data base ............................................................... 1.2 

Fig. 2.1. Burst test facility ............................................................................................................................. 2.1 

Fig. 2.2. Post-test appearance of specimens pressurized by gas and liquid .................................................. 2.2 

Fig. 2.3. Sealing devices ................................................................................................................................ 2.3 

Fig. 2.4. Specimen temperatures versus specimen length and test temperature ........................................... 2.4 

Fig. 2.5. Typical time history of burst test .................................................................................................... 2.5 

Fig. 2.6. M ethod to determine the axial radius of burst region curvature ..................................................... 2.6 

Fig. 2.7. Burst pressure versus temperature .................................................................................................. 2.10 

Fig. 2.8. Circumferential elongation versus temperature .............................................................................. 2.11 

Fig. 2.9. True hoop stress at rupture versus temperature ............................................................................. 2.11 

Fig. 3.1. Design of tube specimen according to the U.S. Standard ............................................................... 3.1 

Fig. 3.2. Design of tube specimen according to the French Standard ........................................................... 3.1 

Fig. 3.3. Design of tube specimen according to the Russian Standard ......................................................... 3.2 

Fig. 3.4. Design of Zr-l%Nb tube specimen used for axial tensile tests ...................................................... 3.3 

Fig. 3.5. Load-displacement curve ................................................................................................................ 3.4 

Fig. 3.6. Increment of interval between marks versus specimen length ....................................................... 3.4 

Fig. 3.7. Ultimate strength versus temperature ............................................................................................. 3.6 

Fig. 3.8. Yield stress versus temperature ...................................................................................................... 3.7 

Fig. 3.9. Total elongation versus temperature ............................................................................................... 3.8 

Fig. 3.10. Uniform elongation versus temperature (i=2.10-3 s-') .............................................................. 3.8 

Fig. 3.11. US/YE ratio versus temperature ................................................................................................... 3.9

vii
LIST OF TABLES

Page 

Table 1.1. M atrix of burst tests ..................................................................................................................... 1.2 

Table 1.2. Matrix of axial tensile tests .......................................................................................................... 1.3 

Table 1.3. Parameters of unirradiated specimens ......................................................................................... 1.3 

Table 1.4. Parameters of irradiated specimens ............................................................................................. 1.3 

Table 2.1. Sensitivity of specimen temperature at the burst to the type of working medium ....................... 2.3 

Table 2.2. Set of procedures to determine specimen geometric parameters after burst tests ....................... 2.6 

Table 2.3. Appearance of unirradiated cladding specimens versus biaxiality and temperature ................... 2.7 

Table 2.4. Appearance of irradiated cladding specimens versus biaxiality and temperature ....................... 2.8 

Table 2.5. Appearance of cross-sections of unirradiated and irradiated cladding specimens in the burst 

region versus biaxiality and temperature ........................................................................................ 2.8 

Table 2.6. Major results of burst tests ........................................................................................................... 2.9 

Table 2.7. Correlations for burst parameters versus temperature, biaxiality and type of cladding .............. 2.11 

Table 3.1. Appearance of unirradiated specimens after tensile tests ............................................................ 3.5 

Table 3.2. Main results of axial tensile tests ................................................................................................. 3.5 

Table 3.3. Results of simple ring tensile tests on irradiated Zr-l%Nb cladding (strain rate 0.002 1/s) ....... 3.7

ix
1. EXECUTIVE SUMMARY 

1.1. Introduction 

There is significant research today on the confirmation of fuel element safety in commercial light-water 

reactors under accident conditions up to high fuel burnup (50-70 MWdfkg U). The development of a data

base on mechanical characteristics of highly irradiated fuel claddings is one of the relevant components of 

this work. With the mechanical characteristics determining the behavior of cladding during a reactivity

initiated accident (RIA) and during the initial stage of the loss-of-coolant accident (LOCA), research to de

velop a database of mechanical characteristics of Zr-l%Nb claddings was therefore initiated in 1997. This 

research was conducted with financial support of Ministry of Science and Technologies of the Russian Fed

eration, AO Mashinostroitelny zavod (Russia, Elektrostal), U.S. Nuclear Regulatory Commission (U.S.), and 

Institute for Protection and Nuclear Safety (France). The work consists of four basic stages: 

1. uniaxial tensile tests in transverse direction with simple ring specimens at temperature range 293

1200 K; 

2. biaxial burst tests with gas pressurized tube specimens at temperature range 900-1400 K; 

3. biaxial burst tests with liquid pressurized tube specimens at temperature range 293-723 K; 

4. uniaxial tensile tests in axial direction with tube specimens at temperature range 293-1200 K. 

The first two stages were completed in 1998. The outcomes of these tests were used to make correlations of 

the basic mechanical characteristics of the unirradiated and irradiated Zr-i%Nb alloy claddings versus tem

perature and strain rate. These correlations were introduced in thermal mechanical codes FRAT-T6 and 

SCANAIR. The upgraded versions of both codes were verified by the results of re-fabricated fuel elements 

test in the IGR reactor under RIA conditions. The detailed description of these researches and their out

comes are presented in [1, 2, 3]. 

Later stages of this research were held in 1999 and 2000. Tasks, techniques and outcomes of these tests are 

described in the present report. 

1.2. Program of tests 

The recent stages of mechanical tests of unirradiated and irradiated Zr-1 %Nb claddings were held to accom

plish the following basic tasks: 

a. to update the previously obtained data base by outcomes of axial tensile tests and biaxial burst tests 

(with different biaxiality ratio); 

b. to conduct the cross-verification of all types of tests results; 

c. to estimate the anisotropy effects for unirradiated and irradiated Zr-i%Nb claddings. 

In accordance with these tasks were determined the following technical requirements to each test stage (see 

Fig. 1.1): 

"* biaxial burst tests: 

:= temperature range 293-723 K; 

=> biaxiality ratio 1 and 2; 

=> pressure increase rate 1 MPa/s; 

"* axial tensile tests:

1.1
=> temperature range 293-1200 K;

Sstrain rate 2-10-3 1/s. 

Unirmadiated and irradiated 

Zr-1*lNb cladding 

-Previo'us obtained- data. -base 

(First and second stages of the program) 

stage of Girdas pressurized tube, Ring specimens, 

Liquid pressurized biaxiality a/a,=2, transverse direction, Tube specimens, tube, biaxiality T=900-1400 K, T=293-1200 K, T s, tube, aelal, pressure increase rate strain rate 0.002-0.5 1/s _axial direction, 

/=2, 1, 0.01-1 NINA/ T=293-723 K, , T=293-1200 K, 

pressure increase strain rate 

rate 1 MPa/s Biaxial data: Uniaxial data: 0.0021/s 

- pressure - ultimate strength - temperature - yield stress 

- strain - uniform elongation 

- total elongation 

Fig. 1.1. Structure of the program to obtain generalized data base. 

These requirements were used as the basis to explore burst and axial test matrixes presented in Table 1.1, 

Table 1.2.

Table 1.1. Matrix of burst tests.

1.2
Table 1.2. Matrix of axial tensile tests. 

Number of tests 

Temperature (K) Unirradiated Irradiated 

293 1 

323 - 1 

423 1 2 

543 1 2 

723 2 2 

1000 1 1 

1200 1 

E___ 1 7 8 

It should be noted that cladding specimens for mechanical tests of both types were made of a material simi

lar to that of the cladding specimens used in the first two stages of the program. So, by analogy with the 

previously performed tests, standard fully recrystallized VVER-1000 commercial tubes were used for unir

radiated specimens (see Table 1.3). 

Table 1.3. Parameters of unirradiated specimens. 

Chemical composition of Zr

1 %Nb cladding Specimens characteristics 

(% by weight) 

"* Nb 0.9-1.1 0 outer diameter 9.11 mm 

"* Fe 0.05 * inner diameter 7.72 mm 

"* Cr 0.02 0 outer ZrO2 thickness (less than) 1 gzm 

"* Ni 0.05 0 length: 

"* Fe+Cr+Ni 0.12 • burst tests 150 mm 

* 02 0.11 = axial tests 50 mm 

The same approach has been applied for irradiated specimens manufactured from two commercial VVER

1000 fuel elements (#165, 154) irradiated in Unit 5 of NovoVoronezh Nuclear Power Plant during four years 

as a part of fuel assembly #4108. Average burnup of fuel elements sections used for manufacturing of 

specimens was about 48 MWd/kg U. The main parameters of irradiated cladding are presented in Table 1.4. 

Table 1.4. Parameters of irradiated specimens.

Specimens characteristics

* outer diameter 

* cladding thickness 

* outer ZrO2 thickness 

* inner ZrO2 thickness (less than) 

* hydrogen concentration (ppm) 

* hydride orientation coefficient 

* length: 

= burst tests 

=. axial tests

r I

9.068-9.096 mm 

0.54-0.84 mm 

3-5 gim 

1 R.m 

51-57 

0.42 

150 mm 

50 mm

1.3

Cladding microstructure

r 0-'
1.3. Selection of test parameters and ranges

The selection of test parameters and ranges was substantially determined by the fact that these tests were 

conducted with the purpose of updating and improvement of the previously obtained database. The inter

coupling of all test types for development of a generalized database is illustrated on Fig. 1.1. 

Tests with different stress biaxiality ratio. 

"* in the high-temperature zone the material is practically isotropic; 

"* in the high-temperature zone the loading of a cladding is determined by the gas pressure, the biaxiality 

ratio at such loading equals 2; 

"* there are no data on burst tests in low-temperature zone; 

"* the low-temperature zone is of interest from the point of view of the pellet cladding mechanical interac

tion at which the biaxiality ratio can vary from 2 to 1, depending on the contact condition; 

"* the pellet cladding mechanical interaction is characterized by high values of the cladding strain rate. 

Axial direction tests. 

The temperature range of axial direction tests has been set the same as for ring tensile tests, because the 

power law parameters are determined just by results of uniaxial tests. The tests were conducted only for a 

basic strain rate, because the strain rate sensitivity exponent describing the influencing of the velocity has 

been determined earlier with a significant amount of experimental data.

1.4
2. CONDITIONS, PROCEDURES AND RESULTS OF BIAXIAL BURST TESTS 

2.1. Test facility for biaxial burst tests 

A special test facility was designed for conducting biaxial tube burst tests in compliance with the following 

requirements: 

"* Temperature range 20-450'C; 

"• Working pressure range 40-200 MPa; 

"* Pressure increase rate - up to 2 MPa/s; 

"* Axial load up to 10 kN; 

"* Tangential and axial stress synchronization at a pre-selected ratio; 

"* Non-uniformity of the specimen temperature not exceeding of ±10°C. 

Schematic Diagram of the test facility, developed to meet these requirements, is shown in Fig. 2.1.

1. H>ydraulic pressure generator 

2. Electric furnace 3. Axial loading mechanism

4. Sensors 5. Control system 

6. Sealing device

Fig. 2.1. Burst test facility. 

The specimen was placed inside the electric furnace to ensure the required specimen temperature. Pressure 

in the specimen was created with the use of a hydraulic pressure generator. A special mechanism was used 

to provide additional axial stress, which varied in synchronism with the pressure inside the specimen at a

2.1

!
pre-selected ratio between tangential and axial stresses (1 or 2). With the axial loading mechanism switched 

off, gas pressure in the specimen provided a biaxiality ratio of 2. With the axial loading mechanism switched 

on, the biaxiality ratio was maintained at 1 with the use of the control system including axial loading control 

and biaxiality ratio control. Special sealing device prevented any leakage from the specimen during the test. 

A series of special research was conducted during development of the test facility to resolve the following 

problems: 

"* selection of working medium inside test specimens; 

"* development of sealing devices for specimens; 

"* ensuring stability of selected temperature. 

Selection of working medium. 

Analysis showed that results of specimen gas pressurization would be difficult to interpret. This is caused 

by the fact that high kinetic energy of gas, flowing from the specimen at burst, produces additional bending 

and tearing of the ruptured area. Therefore, decision was taken to give up gas medium and use liquid pres

surization instead, because, the liquid medium pressure drops practically instantaneously and causes no ad

ditional damage (see Fig. 2.2). However, liquids are not normally used as working medium in the installa

tions of this kind, since it is very difficult to find a liquid, which is free from phase transformations, and will 

not inflame or chemically react with the specimen material throughout the entire temperature range. To 

solve this problem we had to conduct special analytical research, which allowed us to recommend several 

types of liquids for this installation. Analysis of the results of preliminary tests performed on these liquids 

made it possible to select silicone oil "Sopolymer-3" as the working medium. Boiling point, ignition tem

perature and flash point of this oil exceed 450'C. 

Gas pressurization 

Liquid pressurization 

Fig. 2.2. Post-test appearance of specimens pressurized by gas and liquid. 

Sealing problem of specimen. 

Development of a sealing device for specimens also was a challenging task. The main challenge was to seal 

the specimen having high internal pressure at axial loads. Welding of irradiated specimens was unaccept

able, because it could lead to annealing of irradiation damage, which could change mechanical properties of 

these specimens. As a result of the design and process analyses, and a series of scoping tests, three types of 

sealing devices were developed for three types of specimens used (unirradiated specimen, irradiated speci

men without axial loading, irradiated specimen with axial loading). For the design of these sealing devices, 

see Fig. 2.3. All three designs use one sealing principle - sealing with a copper gasket placed between the 

cone and the nut. The difference is mainly in the number of parts used to increase the required axial load.

2.2
Unirradiated specimen

Irradiated specimen 

without axial load

Irradiated specimen 

with axial load

:1 3

-1 

A4

1. Copper gasket 

3. Nut

2: Cladding 

4. Cone

Fig. 2.3. Sealing devices. 

Stability of specimen temperature. 

Analysis of the first version of the burst test facility showed that deformation of a pressurized specimen 

resulted in the increase of its volume, which, in its turn, increased the inflow of working medium from the 

supply pipeline, where the working medium was at an ambient temperature. As a result, the working me

dium and the specimen temperature dropped. The quantitative effect is different for gas and liquid. Results 

of calculations, presented in Table 2.1, demonstrate that replacement of gas with liquid causes a more sig

nificant temperature drop in the specimen. This happens due to higher specific heat and density of liquids. 

Table 2.1. Sensitivity of specimen temperature at the burst to the type of working medium. 

Test p r Working medium 

parameter LTest dGas 

Test temperature (°C) 270 270 

Burst pressure (MPa) 60 60 

Drop of working medium temperature ('C) -53 -53 

Drop of specimen temperature (°C) -35 -9 

To prevent this effect from influencing the test results the facility design was modified to introduce a heater 

for the supply pipeline. The next step in this research was aimed at checking uniformity of specimen tem

perature versus length. Results of special temperature measurements performed in steps of 10 mm at differ

ent nominal temperature levels are shown in Fig. 2.4. The obtained results indicate that temperature non

uniformity over the specimen length for all the tested condition does not exceed ±5°C.

2.3
1100

Fig. 2.4. Specimen temperatures versus specimen length and test temperature. 

2.2. Test procedures and measured parameters 

The burst test procedure included three steps: 

1. preparation; 

2. testing; 

3. post-test examinations. 

Test preparation procedure was as follows: 

"* sealing devices were attached to the specimen; 

"• vertically installed specimen was filled with liquid and sealed; 

"* specimen pressurization was performed to test it for leakage; 

"* the specimen was installed in the electric furnace heated to the specified test temperature; 

"* axial loading mechanism was attached to the lower end of the specimen for the axial loading test; 

"* specimen temperature stabilization was provided by holding it in the furnace for the specified time pe

riod. 

The main step of the test was initiated upon actuation of the hydraulic pressure generator (see Fig. 2.5). Hy

draulic pressure generator operated to increase pressure in the specimen to be in compliance with the test 

scenario. The average pressure increase rate was 1 MPa/s. The following parameters were measured during 

this test versus time: 

"• temperature in electric furnace; 

"* temperature of specimen; 

"* pressure inside specimen; 

"• axial load; 

"* increment of specimen outer diameter.

2.4

Test temperature 4 1 T-i-- 373

"~, ad90 900-.... _ -A- 475 .. . ............ 

2 -O--578 

-- 623 

-- 673 E 7 0 0 . . . . . . . . . . . . . 

500 

t f.grf 

300 

0 20 40 60 80 100 120 140 160 

Specimen length (mm)
Heating Pressurization Cooiin

600 

Time (s)

800

600 

Time (s)

I

Fig. 2.5. Typical time history of burst test. 

During the test, the temperature in electric furnace was maintained with an accuracy of ±5°C. Specimen 

temperature measurement accuracy was also ±5°C. Liquid pressure measurement accuracy was ±1 MPa up 

to 100 MPa, and ±5 MPa above 100 MPa. Axial load measurement accuracy was ±0.2 kN. 

After the biaxial tube burst tests were completed all specimens were subjected to post-tests examinations, 

including the following procedures: 

"* photography; 

"* profilometry; 

"* preparation and photographing of metallographic cross-section of the burst region; 

"* computer processing of the whole set of photographs for: 

Sdetermination of the middle line profile; 

Smeasurement of the cladding thickness;

2.5

60 

40 

20

U

0 

a

0 200 400

Heating Pressurization Cooling

..........................

pressure ..... temperature

" 600 

500 

-400a 

-300 

200 

1000 1200 

ation Cooling -1.00 

-0.75 E 

0.50 

0.25

3-
=> measurement of the axial radius of the burst region curvature; 

=> measurement of the circumferential elongation; 

=> measurement of the circumferential radius of the burst region curvature. 

Previously developed methodological approaches to this research are described in [1]. Brief characteristics 

of the procedures used for the data determination are described in Table 2.2 and Fig. 2.6. 

Table 2.2. Set of procedures to determine specimen geometric parameters after burst tests. 

Specimen parameter Basis of procedure 

1. Outer diameter of cladding Profilometry of specimen in four azimuth directions 

2. Middle line profile Splitting of computer image of the cladding cross-section into sections 

at adequately small azimuth intervals, finding the points corresponding 

to half cladding thickness in the center of each section and plotting the 

envelope through all the obtained points. 

3. Cladding thickness Measuring of cladding thickness in the middle of previously determined 

azimuth intervals, subsequent integration of the obtained data file to 

determine average value of the cladding thickness. 

4. Axial radius of the burst re- See Fig. 2.6. 

gion curvature (r,) 

5. Circumferential elongation of Finding absolute difference between the length of the middle line before 

specimen (ee) and after the test with normalization for the middle line length before 

test (in percent). 

6. Circumferential radius of the ( E 

burst region curvature (re) r = r° (1 +00 

where ro = radius of middle line of cladding (as fabricated) 

o= circumferential elongation.

Fig. 2.6. Method to determine the axial radius of burst region curvature.

2.6
2.3. Results of biaxial burst tests

In accordance with research plan, 28 burst tests were conducted. Qualitative test results are contained in 

Table 2.3-Table 2.5. 

Table 2.3. Appearance of unirradiated cladding specimens versus biaxiality and temperature.

2.7
Table 2.4. Appearance of irradiated cladding specimens versus biaxiality and temperature.

Table 2.5. Appearance of cross-sections of unirradiated and irradiated cladding specimens in the 

burst region versus biaxiality and temperature.

Type of cladding 

and type of biaxi

ality

Unirradiated clad

ding (biaxiality=2)

Unirradiated clad

ding (biaxiality=l1)

[

293-323 K 

#H20-1

Irradiated cladding 

(biaxiality=2)

Temperature and test number 7

423 K 543 K

2.8

723 K

AW-AICM-A
Type of cladding Temperature and test number 

and type of biaxi

ality 293-323 K 423 K 543 K 723 K

Irradiated cladding 

(biaxiality= 1)

I AM lqn"')T #fl27flg_1 T. #O4Oztq-_T.

Analysis of these results allows us to draw the following conclusions regarding the trends in the cladding 

mechanical behavior under these conditions: 

"* circumferential elongation reduces with the increase of temperature; 

"* biaxiality ratio drop from 2 to 1 causes significant reduction of circumferential elongation of unirradi

ated cladding, while irradiated cladding remains less sensitive to this parameter. 

Quantitative assessment of the test results was performed on the basis of the source data given in Table 2.6. 

Table 2.6. Major results of burst tests.

Test No.

Tem

perature 

(K)

Burst 

pressure 

(MPa)

Axial 

load at 

burst 

(N)

Circumfer

ential elon

gation 

(%)

Unirradiated______ 

H20-1 293 88 - 60 0.54 30 6.75 

H20-3 293 94 - 67 0.57 30 7.03 

E50 323 85 - 79 0.54 30 7.54 

E50L 323 94 4630 25 0.58 00 5.30 

E150 423 68 - 77 0.56 Not measured Not measured 

E150L 423 76 3750 21 0.57 Not measured Not measured 

H270-3 543 52 - 74 0.54 34.7 7.34 

H270-4 543 54 - 75 0.55 36.7 7.38 

H270-5L 543 60 3240 19 0.56 00 5.01 

H270-6L 543 61 2910 16 0.59 0 4.90 

E350 623 45 - Not measured Not measured Not measured Not measured 

E350L 623 48 2380 Not measured Not measured Not measured Not measured 

E450 723 40 - 56 0.59 Not measured Not measured 

E450L 723 41 2000 18 0.52 Not measured Not measured 

H450-2 723 43 - 62 0.52 30 6.86 

H450-5 723 42 - 69 0.55 30 7.15 

H450-6 723 40 - 61 0.54 30 6.78 

H450-7L 723 44 2190 16 0.58 1 4.92

2.9
Axial Circumfer- Circumfer

T.ern Burst load at ential elon- Wall thick- ofiauradus ential radius 

Test No. perature pressure burst gation ness (mm) of curvature of curvature 

(K) (MPa) M(%) (r) (mm) (ro) (mm) 

_______ _______ _______ _______ Irradiated_ _ _ _ _ _ 

0150-2 423 127 - 18 0.64 0o 4.93 

0150-IL 423 127 5910 23 0.62 00 5.16 

0150-2L 423 130 6100 11 0.61 00 4.66 

0270-2 543 97 - 14 0.65 00 4.76 

0270-1L 543 93 4650 11 0.62 00 4.63 

0270-2L 543 94 4650 9 0.65 00 4.57 

0450-3 723 71 - 10 0.67 cc 

0450-4 723 70 - 7 0.67 0c 

0450-IL 723 66 3210 8 0.66 00 4.50 

0450-2L 723 66 3300 6 0.66 cc 4.44 

All these data were obtained by measuring the parameters of each test during post-test examinations. Further 

research was aimed at determining correlation of burst pressure, circumferential elongation and true hoop 

stress at rupture versus temperature, type of cladding, and biaxiality ratio (see Fig. 2.7-Fig. 2.9, Table 2.7). 

In doing so, the true hoop stresses at rupture was calculated in compliance with the procedures previously 

developed for the high temperature burst tests and described in detail in [1].

Fig. 2.7. Burst pressure versus temperature.

2.10

150 

100 

• 50

0 + 

200 300 400 500 600 700 800

500 Temperature (K)

o biaxiality=2 (unirradiated) 

* biaxiality=l (unirradiated) 

* biaxiality=2 (irradiated) 

A biaxiality-l (irradiated)

0 0"

1-

bu Pr.grf

0 200 300 400 600 700 goo
100 100

o60 60 

Unirradiated 

"40 -H biaxiality2 40 

0 - biaxiaii =l 

.2 20 . 20 

200 300 400 500 600 700 800 0

L me ie ( K I 

•00 300 400 500 600 700 800 

Temperature (K)

Fig. 2.8. Circumferential elongation versus temperature.

Temperature (K)

Fig. 2.9. True hoop stress at rupture versus temperature. 

Table 2.7. Correlations for burst parameters versus temperature, biaxiality and type of cladding. 

Parameter Type of specimen 

unirradiated irradiated 

Burst pressure Pb = 151.8598267 - 0.2129816731-T - Pb = 343.7481701 - 0.7161358707-T + 

(biaxiality=2) 2.886992355.10-'.T 2+ 1.5397377. 107 .T3 0.0004830832312-T2 

Burst pressure Pb = 135.3007207 - 0.06537602349-T - Pb = 400.4533641 - 0.9828090696.T + 

(biaxiality=l) 0.0002851021788-T2 + 2.7312947 10 7 "T3 0.00091965903T 2 - 2.7630515"10-7"T 3 

Circumferential 

elongation eo = 13.97035999 + 0.2589695178-T 

0.0002665684037.T2 = -0.0315744-T + 31.079106 

(biaxiality=2) II

2.11

1200

S1000 BURST TESTS 

A biaxiality=l (unirradiated) 

0 biwaxility-l (unlim-diatecD 

. 800 - A biaxiality=2 (indiated) 

2 * biaxiality=1 (irradiated) 

" 600 0 biaxiality2 (un6irrAgas) 

* biaxiality=2 (irjd.,gas) 

400

" 200 00-0 th s2:g rf 

0 

200 400 600 800 1000 1200 1400 1600 

Temperature (K)

100 100
Type of specimen unirradiated irradiated 

Circumferential 

elongation E9 = 52.50524937 - 0.1112964864.T + 

8.598062059.10--T2 = -0.0354367T + 30.16835 

(biaxiality= 1) 

True hoop stress 293_<T<723 423_<T<723 

(biaxility=l, 2) cr=2016.268-5.2948"T+0.00627"T 2

- a=4178.356-12.894"T+0.0154"T2

2.8233.10 6-T3 6.5545.10--T3 

973•_T<1190 

True hoop stress a=exp(-0.0065753-T). 116139.0182 

1190T<M 1473 

(biaxility=2) a---exp(-0.004283-T)-7611.8232 

Analysis of the array of these results showed the following: 

1. Strength of irradiated cladding was significantly higher, than that of the unirradiated ones, with the burst 

pressure difference being approximately 40 MPa throughout the entire temperature range of the tests. 

2. With the temperature increase, the burst pressure reduced monotonically, and was less dependent on 

biaxiality ratio. 

3. True hoop stress at rupture for irradiated cladding was significantly higher than that of the unirradiated 

one, however, with the temperature increase this difference was reducing monotonically. We can con

sider with adequate reliability, that the true hoop stress at rupture for irradiated cladding is independent 

of the biaxiality ratio. As for the unirradiated cladding, the growth of the biaxiality ratio from 1 to 2 

causes an increase of the hoop stress at rupture. However, significance of this effect reduces as the tem

perature rises, and the difference practically disappears at the temperature of 723 K. The general trend 

for the correlations obtained for irradiated and unirradiated specimens within the temperature range 293

723 K agrees with the data previously obtained for the high-temperature region (973-1473 K) [1]. This 

proves both the correctness of the experimental and analytical procedures used, and the absence of sig

nificant additional physical effects influencing the cladding behavior. 

4. Circumferential elongation of unirradiated claddings is highly sensitive to the biaxiality ratio. When the 

biaxiality ratio increases, the cladding elongation sharply increases up to 60-75 %. Sensitivity of irradi

ated claddings to this ratio is very low. However, irradiated claddings showed dependence of hoop elon

gation versus temperature. Elongation was monotonically reduced with the increase of temperature.

2.12
3. CONDITION, PROCEDURES AND RESULTS OF AXIAL TENSILE TESTS 

3.1. Test procedure and specimen design 

Axial tensile tests with tube specimens are specified in Russia with a special standard [4]. Still accounting 

for the international character of this program it was decided to perform comparative reassessment for the 

standards used for these purposes in the Russian Federation, U.S. and France. 

U.S. Standard ASTM E8M-93 [51. 

According to this standard small diameter tubes (D<25 mm) are usually tested as sections of a bulk cross

section tube. The main standard requirements to specimen and tooling, are shown on Fig. 3.1.

Fig. 3.1. Design of tube specimen according to the U.S. Standard. 

Cone grips or collet grips are used for tests. Plugs should be used in both ends of the specimen. Two sizes 

are strictly stipulated: 

"• the spacing interval between the butt ends of the plugs equals the gauge length; 

"* the spacing interval between the machine jaws should be longer than the gauge length by four internal 

diameters of the specimen. 

French Standard EN10002-5:1991 [61. 

The specimen sizes according to the French standard are shown on Fig. 3.2. The main differences from the 

American standard are: 

"* the plugs used are of the cylindrical shape; 

"* the gauge length (1a) is set smaller than the spacing interval between the butt ends of the plugs (L.); 

"* the length of tube specimens (Lt) should be longer than the gauge length by two external diameters (D.).

Fig. 3.2. Design of tube specimen according to the French Standard.

Russian Standard GOST 10006-80 [41. 

The requirements of the Russian standard to the shape and sizes of tubular specimens are shown in Fig. 3.3.

3.1

UT"'/''/--/''/•Gauge g . 1-cone grip or collet 

2-tube specimen 

3-specimen plug
Fig. 3.3. Design of tube specimen according to the Russian Standard.

This standard integrates the requirements of both U.S. and French standards: 

"* the use of both conical and cylindrical plugs is enabled; 

"* every plug should go into a sample on a length not less than four diameters; 

"* the gauge length (lg) is set smaller than the spacing interval between butt ends of the plugs (as in the 

French standard); 

"* the spacing interval between machine jaws should be longer than the gauge length by four diameters of 

the tubular specimens (as in the American standard), but the external ones (as in the French standard); 

"* a special marking is put on the specimen length (Lm) exceeding the gauge length by two external di

ameters (in the French standard it is the overall sample length) and allows afterwards to conduct the re

finement of the elongation of a sample by matching the point of fracture with the middle of the gauge 

length. 

Specimen design. 

Analysis results of the of the reviewed standards requirements and features of the used testing machines 

were allowed for specimen design. The gauge length was set in accordance with the requirement of the Rus

sian standard based on the following term: 

Ig =5.65 E(D2-d2). 

ý4 

Two designed types of plug and specimen attachment are shown on Fig. 3.4. Plugs with special head for 

machine mandrels were used in case of unirradiated specimens (see Fig. 3.4a). In this case plugs and speci

men were joined by means of welding. Such a design is a compromise with respect to all reviewed stan

dards. In this case the deviation in the spacing interval between machine jaws has no value as the effort is 

transmitted by weld joint, instead of friction. 

Collet grips were used for irradiated samples in the temperature range of 293-723 K. Such a design does not 

also greatly contradict the reviewed standards and allows one to eliminate the weld joint which would have 

lead to the annealing of the irradiated material and, as a result, to systematic errors in the measurements. 

Nevertheless, mechanical tests of irradiated specimens of temperatures above 1000 K were performed for 

the same specimen design as for the unirradiated cladding because annealing of irradiation damage in the 

whole of the specimen will most evidently take place under these high temperatures. 

Special marks were put on all samples as hairlines of 30-40 gtm depth and -5 mm arc length in order to 

validate and to adjust the gauge length of the samples. These marks were put on along the length of 45 mm 

with a step of 1 mun. Besides, additional scoping tests were performed to certify test devices. Results of 

these tests have demonstrated that: 

=> measurement uncertainty of the axial load was ±1%; 

=> measurement uncertainty of the test temperature was ±3 K for the temperature range of 293-723 K, 

and ±8 K for the temperature range of 1000-1200 K.

3.2
Unirradiated Irradiated

Fig. 3.4. Design of Zr-i %Nb tube specimen used for axial tensile tests. 

Two test machines were used for the whole set of research. Low temperature tests were performed with the 

air inside the machine. Vacuum (up to 10-4 mm mercury) was used to reach high temperature test conditions 

(1000-1200 K). 

Approach to adjust the specimen gauge length. 

As was mentioned above, special attention was paid to the specimen gauge length while developing proce

dures for the axial tensile tests. 

Special test was performed with irradiated specimen according to the following scenario: 

=* load relief was arranged periodically during the loading test (see Fig. 3.5); 

= the first load relief was arranged close to the yield stress point; 

* the second one was arranged in the section between the yield stress and the ultimate strength; 

= the third relief was arranged close to the ultimate strength. 

In addition to the major goal, this test was used to verify elastic properties of the mechanical system (in

cluding the specimen and machine). Data presented in Fig. 3.5 indicate that the apparent elastic modules of 

this system (tgu,) preserve high stability during the whole process of loading.

3.3

S• 2 - plug 

S~1 - nut 

3 - collet 

.. t-1t 4- cone 

If 5-nut 

6 - specimen 

a) b)
Fig. 3.5. Load-displacement curve. 

This check has proved that the obtained load displacement curves, processed according to the standard pro

cedures, allows one to determine the correct values of the specimen elongation. 

An additional procedure was developed to use this test for validation of the specimen gauge length. This 

procedure included measurements of the intervals between special marks put onto the specimen (with a 

space of 1 mm) before the test. After that, the increment of the interval initial length was determined for 

three representative points of the load displacement curve: 

=, yield stress area (after the first load relief); 

=> intermediate area between the yield stress and the ultimate strength (after the second load relief); 

> rupture area (after the test). 

Fig. 3.6 presents the results of these tests.

1.2 

, 1.0 

P 

0.8 

i .

S0.6 

S0.4 

E 0.2 

0.0

-0.2

0 5 10 15 20 25 30 

Number of interval

Fig. 3.6. Increment of interval between marks versus specimen length.

3.4

6000 

400-T=423 K 

4000 

z 

0 

2000 

tggcý=-guog 

0 • 

30 35 40 45 

Displacement of mandrel (mm)

A ntft gauge ngth T=423 K 

_____ ____ ______ -i- after first load relief 

....- after second load relief 

-.... after rupture 

____ __ ___ ____ _ ___ _ ____ __ ___ __ ___ risk.g~f

35 40 45 50
One could see that the deformation distribution along the sample length is of a symmetrical nature, the fail

ure has taken place in the central part of a sample, and the main portion of deformation is concentrated 

within the bounds of the gauge length during the all test. 

3.2. Results of axial tensile tests 

Typical appearance of the specimen after uniaxial tests is represented in Table 3.1, Table 3.2. 

Typical appearance of two types of the uniaxial tests is presented in the above tables to compare mechanical 

behavior of Zr-1 %Nb cladding versus the load direction and temperature. 

Table 3.1. Appearance of unirradiated specimens after tensile tests.

Temperature Tube specimen (axial tests) Ring specimen 

(K) (transverse tests) 

293 

423 - LI~1~ -. J II 

543 ~l 

723 (623 K) 

723M 

1000 _ b 

(923 K) 

1200 _ IW____ ( 

(1223 K)

These results indicate that the earlier revealed (for ring specimens) tendency of sharp increase of the sample 

elongation is observed around 1000 K. With further temperature increase, the elongation noticeably de

creases. 

Processing of all stress-strain curves, measured in axial tests, allowed us to obtain the set of mechanical 

properties of Zr-1 %Nb cladding presented in Table 3.2. 

Table 3.2. Main results of axial tensile tests. 

Yield stress Ultimate strength Total elongation Uniform elonga Number of test T (K) (MPa) (MPa) (%) tion (%) 

Unirradiated 

T20-2 293 237 378 45.1 18.5 

T150-1 423 196 311 43.1 19.4 

T270-5 543 118 225 54.1 28.3 

T450-3 723 100 176 49.1 21.7 

T450-7 723 99 184 54.1 20.2 

T727-2 1000 56 59 7.9 

T927-2 1200 8.6 11.6 80.6 5.7

3.5
Yield stress Ultimate strength Total elongation Uniform elonga Number of test T (K) (MPa) (MPa) (%) tion (%) 

__________.__ :Irradiated 

T50-0 323 520 585 18.3 2.8 

T150-0-1 423 498 553 9.4 2.0 

T150-0-2 423 516 572 13.0 3.1 

T270-0-1 543 440 485 11.5 2.5 

T270-0-2 543 409 461 14.2 2.0 

T450-0-w 723 300 342 20.0 1.7 

T450-0-2 723 317 326 12.9 1.2 

T727-0-w 1000 51 52 - 3.3 

The generalized data on uniaxial tensile tests in transverse and axial directions are presented in Fig. 3.7 

Fig. 3.10 in order to perform comparative analysis of all the uniaxial test results. 

It should be noted, that these plots contain the results of recent additional ring tensile tests of irradiated 

specimens in temperature range of 673-773 K. The tests were performed in order to define more exactly the 

temperature dependencies of mechanical properties in transverse direction and to compare with axial test 

data. Ring specimens have been made from the irradiated claddings, which were used for axial tensile tests 

(see section 1.2 of this report). The results of additional ring tensile tests do not practically affect previously 

obtained temperature dependencies, so Fig. 3.7 - Fig. 3.10 contain the same correlations that were published 

in [1]. The new results of transverse tests are also presented in Table 3.3. 

600

500- Ring specimens (transverse) 4,00 ____..____ unirradiated 0.• 0 irradiated : 

4"best fits 

S3A 00Tube specimens (axial) •300 A unirradiated 

.200 - irradiated " 

100" 

0 1usOlnn.grf• , 

200 400 600 800 1000 1200 1400 

Temperature (K) 

Fig. 3.7. Ultimate strength versus temperature. 

Circles indicate the results of tests in the transverse direction; triangles indicate those in the axial direction. 

Light marks relate to unirradiated specimens, the black ones to irradiated specimens. The correlations pre

sented were obtained from the results of tests in the transverse direction. The procedure for the obtainment 

of those is described in detail in [1]. 

The analysis of the ultimate strength data allows to reveal the following. At low temperatures, the ultimate 

strength in the axial direction is approximately 10% higher than that in the transverse direction. With tem

perature increase, this difference disappears. In the high temperature region, the values of the ultimate 

strength in the transverse and axial directions coincide. The noted tendency is equally valid both for unirra

diated claddings and for the irradiated ones. Thus, the increase in the ultimate strength due to the irradiation 

in the reactor is similar in the transverse and axial directions.

3.6
Table 3.3. Results of simple ring tensile tests on irradiated Zr-i %Nb cladding (strain rate 0.002 11s). 

Temperature Ultimate Yield stress Total elongation. Uniform elon

(K) strength (MPa) (MPa) M gation 

324 296 12.3 3.9 

673 

334 319 13.0 2.7 

313 293 11.0 2.4 

698 

301 280 12.3 4.2 

299 270 11.4 4.9 

723 

278 258 10.5 4.0 

299 278 10.8 3.4 

748 

275 254 14.4 4.2 

246 222 12.8 3.6 773 

258 240 18.4 3.5 

The effect of the base irradiation on the yield stress in the transverse and axial directions manifests itself 

differently. In the low temperature range, the yield stress in the axial direction for unirradiated samples is 

approximately 30% less than that in the transverse direction, and for irradiated specimens, the yield stress in 

the axial direction is approximately 10% higher than that in the transverse direction. With temperature in

crease, this difference decreases as well as for the ultimate strength. Thus, the increase in the yield stress 

due to the reactor base irradiation is significantly higher in the axial direction than in the transverse one. 

The observed loading direction effect on the yield stress and the ultimate strength for unirradiated samples is 

a consequence of Zr-1%Nb alloy anisotropy and quantitatively well matches the published data [7]. For 

irradiated samples, the effect of the loading direction essentially decreases indicating that the irradiation 

reduces the anisotropy of the yield stress for Zr-l%Nb alloy.

Fig. 3.8. Yield stress versus temperature.

3.7

Ring specimens (transverse) 

500o 0 o unirradiated 

A * irradiated 

~400- Abest fits 

4300 ___ _ __ :Tube specimens (axial) 

.• •A: 300 tirradiated 

SA &"A"irradiated : 

.2200-A 

100 

ys-02nn.grf 

0

200 400 600 800 1000 1200 1400 

Temperature (K)
160

Fig. 3.9. Total elongation versus temperature. 

Comparative analysis of the test data, characterizing total elongation of the cladding versus irradiation de

gree and temperature, leads to the following observations: 

of the unirradiated Zr-1 %Nb cladding specimen takes place in the low temperature zone; 

pronounced anisotropy of irradiated cladding was registered in any of the temperature regions. 

= total elongation versus temperature for both specimen types and in both directions can be divided 

into three typical zones, the boundaries of which correspond to the temperatures of allotropic phase 

transformations of Zr-l%Nb alloy (see Fig. 3.9): 

* total elongation of the a phase of Zr-1 %Nb alloy slightly depends on the temperature; 

* abrupt increase of the total elongation takes place during a + 13 phase of Zr-l%Nb alloy; 

* total elongation gets significantly less when transferring to the 03 phase of Zr-I%Nb alloy.

30 1 

A Ring specimens (transverse) 

25 0 unirradiated 

A - irradiated 

=20- best fits 2 A Tube specimens (axial) 

r_ 1 A unirradiated 

0 15 ___ 0 A, irradiated 

• •1• "A •ueO4nn.grf 

0 

200 400 600 800 1000 1200 1400 

Temperature (K) 

Fig. 3.10. Uniform elongation versus temperature (i=2-10- s-'). 

The analysis of the uniform elongation data shows that this parameter depends strongly on the loading di

rection. For unirradiated samples, the uniform elongation in the axial direction under low temperatures sig

nificantly (several times) exceeds that in the transverse direction. For irradiated samples, the uniform elon

gation in the axial direction is moderately less than that in the transverse direction. As the temperature in3.8

Ring specimen (transverse) 

0I unirradiated

o'120 -- :° -irradiated _ _ __ _ _ _ _ _ _ _ _ _ _ _ 

-best fits 

Tube specimens (axial) 

S80- A unirradiated 

A irradiated 0 A I/ 

0 ,_te _O 3nn.grf 

200 400 600 800 1000 1200 1400 

Temperature (K)
creases, the influence of the irradiation and loading direction over the uniform elongation diminishes as in 

the case for the ultimate strength, the yield stress and the total elongation. 

The revealed sharp decrease of the uniform elongation in the axial direction due to the irradiation is the im

portant result. A careful analysis of the experimental procedures showed up the lack of significant system

atic errors in measurements. 

It should be also noted that a sharp decrease of the uniform elongation in the axial direction due to the irra

diation is qualitatively in accord with the results of measurements of ultimate strength (US) and yield stress 

(YS). Fig. 3.11 illustrates US/YS ratio versus the temperature for the irradiated and unirradiated samples 

tested in the transverse and axial directions. The US/YS ratio characterizes the material potentiality for the 

strain hardening and, consequently, must correlate with the uniform elongation value. As can be seen from 

Fig. 3.11, the US/YS ratio for unirradiated samples tested in the axial direction under the temperatures lower 

than 800 K is significantly higher than that for the other specimens. 

M 2.00-I 

= 1.80 A Ring specimens (transverse) 

W 0 unirradiated (measured) 

__.• 1.60 A A L irradiated (measured) 

Tube specimens (axial) 

41 A unirradiated (measured) 1.40

,,A irradiated (measured) 

1.20 

' S1.00 

200 400 600 800 1000 1200 1400 

Temperature (K) 

Fig. 3.11. US/YE ratio versus temperature. 

The obtained results indicate an essential change in the anisotropy of Zr-I %Nb alloy mechanical properties 

at high burnup. In the course of a base irradiation, the cladding is not only subjected to radiation damages 

but also accumulates residual strains caused by the coolant excessive pressure and PCMII during power 

ramps. The total influence of those factors leads to different changes of mechanical properties in the axial 

and transverse directions. 

Results of the axial tensile tests presented in this section, as well as the results of the comparative analysis 

of both types of uniaxial tensile tests allow us to consider that this database together with the database ob

tained due to biaxial burst tests can be the basis to determine anisotropy coefficients, deformation law, and 

to develop failure criteria for Zr-1 %Nb cladding.

3.9
4. CONCLUSIONS

New research was performed to study mechanical properties of unirradiated and irradiated Zr-l%Nb clad

ding. The objective of the previous program was to measure mechanical properties of Zr-i%Nb cladding in 

the transverse direction, and under the high temperature internal gas loading conditions (burst tests). The 

present program was designed to determine mechanical properties in the axial direction, and to obtain me

chanical properties of Zr-I%Nb cladding for low temperature biaxial loading (burst tests). 

Results of these tests lead to the following conclusions: 

1. Axial tensile tests showed the same general tendencies of the cladding mechanical properties obtained 

with uniaxial tests in the transverse direction. These are: irradiation of Zr-i %Nb cladding leads to the in

crease of the cladding strength and to the decrease of the cladding ductility in the low temperature range. 

The difference between mechanical properties of unirradiated and irradiated claddings disappears com

pletely at the temperatures above 860 K. 

2. Low temperature burst tests have demonstrated high sensitivity of the circumferential elongation of unir

radiated claddings to biaxiality ratio. Strength parameters of both types of claddings and circumferential 

elongation of irradiated cladding are independent of the biaxiality ratio. 

3. Analysis of all the summarized mechanical data shows that the anisotropy effect is insignificant for irra

diated cladding in the whole temperature range studied. Anisotropy effect is very evident in unirradiated 

claddings for the yield stress, uniform elongation, and total elongation at low temperatures. 

4. Results of the axial tensile tests and low temperature biaxial burst tests along with the results of the pre

viously performed transverse tensile tests, and high temperature burst tests will be used as the basis to 

determine the anisotropy coefficients of Zr-l%Nb cladding versus temperature, as well as to develop the 

deformation laws for thermal mechanical codes. In addition to that, the test results could be used to de

velop the cladding failure criteria under low and high temperature conditions. 

References 

[1] L.Yegorova, V.Asmolov, G.Abyshov, V.Malofeev, A.Avvakumov, E.Kaplar, K.Lioutov, A.Shestopalov, 

A.Bortash, L.Maiorov, K.Mikitiouk, V.Polvanov, V.Smirnov, A.Goryachev, V.Prokhorov, and A.Vurim 

"Data Base on the Behavior of High Burnup Fuel Rods with Zr-I%Nb Cladding and U0 2 Fuel (VVER 

Type) under Reactivity Accident Conditions", RRC "Kurchatov Institute" report NSI RRC 2179, Vol.2, 

1999 (also USNRC report NUREG/LA-0156 and IPSN report IPSN 99/08 - 2). 

[2] A.Shestopalov, K.Lioutov, L.Yegorova, G.Abyshov, K.Mikitiouk "Modification of U.S. NRC's FRAP

T6 Fuel Rod Transient Code for High Burnup VVER Fuel", RRC "Kurchatov Institute" report NSI 

RRC 2180, 1999 (also USNRC report NUREGJIA-0164 and IPSN report IPSN 99/10). 

[3] K.Mikitiouk, A.Shestopalov, K.Lioutov, L.Yegorova, G.Abyshov "Modification of IPSN's SCANAIR 

Fuel Rod Transient Code for High Bumup VVER Fuel", RRC "Kurchatov Institute" report NSI RRC 

2181, 1999 (also USNRC report NUREG/IA-0165 and IPSN report IPSN 99/09). 

[4] GOST 10006-80 Metallic tubes - Procedure of tensile tests (ms.) 

[5] ASTM E8M-93 Standard Test Methods for Tension of Metallic Materials (Metric). 

[6] EN10002-5:1991 Materials metalliques- Essai de traction-Partie 5: Methode d'essai a temperature elevee. 

[7] A.Zaimovskyi, A.Nikulina, N.Reshetnikov "Zirconium Alloys in the Nuclear Industry". Moscow, Ener

goizdat, 1981 (rus). 

4.1
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add VoL, Supp., Rev., 

NRCM 1102, and Addendum Nurbe. If any.) 

3201, 3202 BIBLIOGRAPHIC DATA SHEET 

(See intstiucions on Me reverse) NUREG/IA-0199 

2. TITLE AND SUBTITLE IPSN 01-16 

NSI RRC 2241 

Mechanical Properties of Unirradiated and Irradiated Zr-1 %Nb Cladding 3. DATE REPORT PUBLISHED 

Procedures and Results of Low Temperature Biaxial Burst Tests and Axial MONTH YEAR 

Tensile Tests April 2001 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

E. Kaplar, L. Yegorova, K. Lioutov, A. Konobeyev, N. Jouravkova, NSIRRC 

V. Smimov, A. Goryachev, V. Prokhorov, 0. Makarov, S. Yeremin, A. Svyatkin, RIAR Technical 

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. NuclearRegulatory Commission, and mailing address. if contractor, 

provide name and mailing address.) 

Nuclear Safety Institute of Russian Research Centre "Kurchatov Institute" Kurchatov Square 1, Moscow 123182 Russian Federation

State Research Centre "Research Institute of Atomic Reactors" Dimitrovgrad 433510, Russian Federation

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, type 'Same as aboy if contrador provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,

9. S PO0NSORI NG 0ORGA NIZATI ON - NAM E AN D AD DRESS (If NRC, type -Same as above" -, f contractor, provide NRC Division, office or Region, u. S. Nuclear Regulatory Commission, 

and mailing address.) 

Division of Systems Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

"1 U. I""LEME-A Ii T rNO I 1=5

11. ABSTRACT (200 words orless) 

The present program was designed to determine mechanical properties in the axial direction, and to obtain mechanical 

properties of Zr-1%Nb cladding for low temperature biaxial loading (burst tests). 

Results of these tests lead to the following conclusions: 

1. Axial tensile tests showed the same general tendencies of the cladding mechanical properties obtained with uniaxial tests 

in the transverse direction. These are: irradiation of Zr-1 %Nb cladding leads to the increase of the cladding strength and to the 

decrease of the cladding ductility in the low temperature range. The difference between mechanical properties of unirradiated 

and irradiated claddings disappears completely at the temperatures above 860 K. 

2. Low temperature burst tests have demonstrated high sensitivity of the circumferential elongation of unirradiated claddings 

to biaxiality ratio. Strength parameters of both types of claddings and circumferential elongation of irradiated cladding are 

independent of the biaxiality ratio. 

3. Analysis of all the summarized mechanical data shows that the anisotropy effect is insignificant for irradiated cladding in the 

whole temperature range studies. Anisotropy effect is very evident in unirradiated claddings for the yield stress, uniform 

elongation, and total elongation at low temperatures.

12. KEY WORDS/DESCRIPTORS (List words or phrases that wivlassist researchers in locating the report.) 13. AVAJLABILITY STATEMENT 

Zirconium-1% Niobium alloy unlimited 

burst tests 14. SECURITY CLASSIFICATION 

material properties Mhis Page) 

cladding strain and ballooning unclassified 

test procedures 

post-test examinations (This Report) unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89)

d
Federal Recycling Program
NUREG/IA-0199 MECHANICAL PROPERTIES OF UNIRRADIATED AND IRRADIATED ZR-I% NB CLADDING APRIL 2001

UNITED STATEb 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300