AMES-REPORT No 5Stuttgart, December 1995

Jürgen FöhlEUR 16313 EN

SURVEY OF NATIONAL REGULATORY REQUIREMENTSAND IDENTIFICATION OF EXISTING, PLANNED AND

REQUIRED STANDARDS AT EUROPEAN LEVELRELEVANT WITH IRRADIATION DAMAGE

AND MITIGATION METHODS

PART II

SURVEY OF EXISTING, PLANNED AND REQUIRED STANDARDS

EUROPEAN NETWORK ON AGEING MATERIALS EVALUATIONAND STUDIES

AMES

contact for copies: JRC-IAM of EC, Ulrik von Estorff, P.B. 2, NL-1755 ZG PettenTel.: +31.22456.5325, Fax: +31.22456.1568, E-mail: vestorff@jrc.nl

PrefaceThis work was financed by the Directorate General XI of the European Commission

through the contract COSU-CT-0036 with TRACTEBEL ENERGY ENGINEERING :

“Survey of National Regulatory Requirements and Identification of Existing,Planned and Required Standards at European Level Relevant withIrradiation Damage and Mitigation Methods“.

The present report, covering the survey of the National Standards has been prepared at

MPA Stuttgart.

The information necessary to compile the present report was gathered largely by means

of detailed questionnaires. The author acknowledges with gratitude the contributions of

the different national correspondents:

Dr. A. Fabry SCK·CEN, Belgium

Prof. Törrönen VTT, Finland

Mr. J. Koban Siemens (KWU), Germany

Dr. Ch. Leitz Siemens (KWU), Germany

Dr. A. M. Kryukov KURCHATOV INSTITUTE, Russia

Dr. A. Ballesteros TECNATOM, Spain

Dr. K. Gott SKI, Sweden

Dr. C. J. Bolton Nuclear Electric, U.K.

The author also wishes to express his gratitude to Mr. Robert Gerard for reviewing this

report and giving valuable comments and Acad. Myrddin Davies for conducting a peer

review of this report.

Table of contents

Preface

page

1 Introduction 12 Summary 33 Identification of Existing, 5

Planned and Required Standards3.1 Determination of Mechanical Properties 5

3.1.1 Specimen Sampling 6

3.1.2 Specimens and Test Procedures 13

3.1.2.1 Tensile Test 13

3.1.2.2 Charpy-V-Notch Test 13

3.1.2.3 Drop-Weight Test 16

3.1.2.4 Fracture Mechanics Test 16

3.1.2.5 Small Scale Specimen Testing 22

3.1.2.6 Reconstituted Specimens 30

3.2 Determination of Neutron Exposure 34

3.2.1 Neutron Field Calculation 36

3.2.2 Exposure Units 38

3.2.3 Neutron Dosimetry 39

3.2.4 Evaluation of Surveillance Results 41

3.3 Determination of Irradiation Temperature 42

4 Conclusions 44

5 Future work 47

References 49

Standards [S] 48

Regulatory Requirements [R] 53

Literature [L] 54

1

1 Introduction

The objective of this report is to provide a survey of national Standards used in European

countries to perform and to evaluate test results in order to describe degradation of reactor

pressure vessel steels and welds of light water reactors during service.

For the safety assessment conservative material data are needed to evaluate the safety

margin at any point in time in the life of a plant. The material has to be characterized in the

initial state and after irradiation and the data must be applied to the RPV wall, which

requires knowledge on neutron field and effective temperature at both the surveillance

position and RPV wall, Fig. 1.1. The material characterization comprises

• adequate and representative material sampling

• conservative and lower bound specimen orientation

• determination of strength and toughness properties

The Standards in use cover most of the necessary testing and evaluation methods.

However, due to limited amount of irradiated materials and specimens additional test

procedures had to be applied in the past which are not yet sufficiently covered neither by

national nor by international Standards. The survey is mainly a compilation of Standards as

they are applied in different European countries derived from the questionnaires which were

returned from Belgium, Finland, Germany, Russia, Spain, Sweden and the United Kingdom.

Commonly used and diverse procedures are discussed. Besides that it is pointed out in

what fields additional activities have been started and what kind of Standards are needed to

achieve international acceptance of test procedures developed in individual laboratories.

It should be made clear, that the information provided in this report is not necessarily

complete since it is only based on the answers to the questionnaire. The report does not

replace any of the national procedures in evaluating the safety of a plant.

2

2 Summary

The change in material properties due to neutron irradiation is monitored by means of

surveillance programmes. Layout of surveillance programmes is standardized in most

countries either in national Standards and Regulatory Requirements or by adopting the

American Standard ASTM. To perform testing and evaluation of the results Standards for

− selection of materials and specimen sampling

− mechanical testing

− neutron exposure calculation and measurement

− temperature monitoring

are in use.

On the basis of a questionnaire which was sent to AMES participants and returned from

most members, it could clearly be seen that ASTM Standards are dominating in this field.

Parallel to this, there exist a few European EN Standards and activities in the European

Structural Integrity Society are under way. A review of the ASTM Standards and other

Standards or recommendations has shown, that only slight differences exist, which are not

supposed to affect the general safety strategy. However, it is desirable from the author`s

point of view to harmonize the test procedures also in the very details and to implement the

present practice in the EN Standards. In those activities it has strictly to be distinguished

between the level of Standards and Regulatory Requirements.

It could be recognized that some areas exist in which more technical background has to be

accumulated and where research activities are presently being performed which may finally

lead to Standards in the future. Those areas are

− use of small scale notch impact specimens, testing, evaluation and

transferability of results

− reconstitution techniques beyond the level of the existing ASTM Standard in

order to apply this technique to smaller remnants

3

It is recommended that these activities be supported because of their future importance.

With regard to neutron field determination and temperature monitoring there exist ASTM

and national Standards for performing specific tests. However, the strategy regarding the

kind of actions that have to be performed to obtain reliable data for the safety analysis is

rather on the level of Regulatory Requirements and plant specific expert opinions. In this

respect, the presently available Standards seem to be adequate and sufficient.

Harmonization on a European level, however, is desirable.

4

3 Identification of Existing, Planned and Required Standards

The material state of the reactor pressure vessel in the core belt line region at any time can

be characterized by the

- initial properties

- change in properties under service conditions depending on

- chemical composition- neutron exposure- temperature- time.

Data are necessary to evaluate conservative bounds to be used in the safety analysis with

respect to brittle and ductile failure.

In the following, the present Standards are reviewed which are in use to perform the

adequate tests and to generate the necessary data.

The review focusses on

- mechanical testing

- calculation of neutron exposure

- experimental determination of neutron exposure

The review has generally shown that in the different countries either the ASTM Standards

are directly used or that the technical content of the national Standards is quite close to the

ASTM Standards. Therefore the general technical issues of the ASTM Standards will be

described in the following text indicating in which countries significant deviations have to be

considered.

3.1 Determination of Mechanical Properties

In the unirradiated condition the material behaviour is usually described on a much broader

basis than in the irradiated condition. This is due to the fact that in irradiation channels of

power reactors there is only limited space and therefore only a limited number of specimens

as large as urgently necessary is irradiated in the frame of surveillance programmes.

5

In the report

"Survey of National Regulatory Requirements, Part I " [L1],

it is stated which material properties are needed to evaluate the material bounds for the

safety assessment and how fracture toughness curves can be derived for the irradiated

state based on unirradiated data adjusted on the basis of test results of irradiated Charpy

specimens according to the “reference temperature concept“. In general the following tests

have to be performed:

- tensile test

- Charpy-V-notch test

- drop-weight test

- fracture mechanics test

In order to obtain representative and conservative data for the component, requirements

exist for specimen sampling and specimen orientation.

3.1.1 Specimen Sampling

Representativity of Materials

Specimens to describe the material in the unirradiated state and specimens for the

surveillance program shall be taken from the actual materials used in fabricating the beltline

of the reactor pressure vessel. The fabrication history shall be fully representative for the

materials in the beltline region of the vessel with regard to austenitizing, quench and

tempering and post weld heat treatment. This requirement has to be fulfilled according to

ASTM E 185 [S1] to obtain credible date from the surveillance programme in agreement

with U.S. Regulatory Commission Regulatory Guide 1.99 [R1] and e.g. according to KTA

3203 [R2].

Location of Specimens

Specimens used to determine representative properties of the base material shall generally

be taken from a depth of at least ¼ T (T = thickness of the forging or plate) from the

quenched surface and at least 1 T according to ASME NB-2000 [R3] and ½ T according to

KTA 3201.1 [R4], respectively, from quenched edges.

6

For the surveillance program, ASTM E 185 [S1] requires to locate the specimens near the

¼ T plane. It is not explicitly mentioned if ¼ T is measured from the inner or the outer

surface. For plates it is not substantial, however, for forgings it might cause differences due

to segregations. KTA 3203 [R2] allows specimens to be taken from the ¼ to ¾ T location.

This range of material is also used in some cases - because of lack of material - even when

ASTM Standard is applied. In case of the weld, 10 mm (KTA 3203) and 12.7 mm (ASTM E

185), respectively, shall be removed from the surface and the root, the remaining part is

considered to be representative across the entire thickness. The same applies for the heat

affected zone (HAZ).

In Finland [R5] and Russia, where ASTM Standards are not applied, similar requirements

exist as shown in Tab. 3.1.

Specimen Orientation

In general, Charpy specimens are taken from the base material in the transverse direction

(T-L). Examples for designation of direction and the nomenclature of specimens are given

in Fig. 3.1 and 3.2. Specimens from the weld have also to be taken as transverse

specimens (T-L; L-direction for weld = welding direction), Fig. 3.3. In the past the notch

orientation for weld specimens was frequently used to have the crack running in the

thickness direction (T-S) and in some cases this was also applied for base material. For

testing the HAZ, transverse specimens (T-L) are applied. They are in some cases

individually adjusted to the course of the fusion line. Fig. 3.4 compares the different

orientation of specimens from the HAZ as required in ASTM E 185 and KTA 3203.

7

L = circumferential direction of forging,main forging direction

T = axial direction of forging

S = thickness direction

L = rolling direction

T = long transverse

S = thickness direction(short transverse)

Fig. 3.1: Commonly used designation of directions in forgings and plates with regard to themain working direction of the material

8

L = main rolling and forging direction, respectively

T = long transverse direction

S = thickness direction

Fig. 3.2: Example of specimen orientation and designation according to internationally used nomenclature

9

ASTM E 185

Fig. 3.3: Type, location and orientation of specimens from weld and HAZfor the surveillance programme

10

Tensile specimens are used in transverse direction (T) from the base material and in

longitudinal direction (L) from the weld (longitudinal = welding direction). Transverse

specimens can be used according to ASTM E 185, if the gauge length consists entirely of

weld metal.

3.1.2 Specimens and Test Procedures

3.1.2.1 Tensile Test

Tensile properties are generally determined with round bars in a quasi static test. Size of

specimens may differ according to the available space in the surveillance capsules. Size

and shape may also be different for testing unirradiated and irradiated specimens.

However, strength properties obtained with specimens of different size are generally

considered to be adequate. Countries working on the basis of ASTM perform testing

according to ASTM E 8 [S2] and at elevated temperature according to ASTM E 21 [S3].

For tensile testing, there exists a European Standard e.g in Germany DIN EN 10 002

[S4] or in the U.K. BS EN 10 002 which has replaced national Standards.

Differences exist in the results of elongation at fracture achieved by ASTM E 8 and the

European Standard. The gauge length according to ASTM E 8 is four times the diameter,

in EN 10 002 it is five times the diameter. The ASTM results are therefore slightly higher.

There exists, however, a metric version of the ASTM Standard - ASTM E 8 M [S5] - in

which the gauge length corresponds to the European short proportional tensile specimen

(l = 5 · do).

3.1.2.2 Charpy-V-Notch Test

To obtain toughness properties, Charpy-V-notch specimens (10 x 10 x 55 mm3) are used

according to the U.S. Standard ASTM E 23 [S6]. Most European countries have adopted

the ASTM Standard. In Germany DIN 50115 [S7] includes the same specimen geometry

as in ASTM, however, the geometry of the impact machine is slightly different from that of

the ASTM machine, mainly with regard to the striking edge.

11

The Standards for the Charpy impact test on an European level are e.g. the DIN-EN

10 045, part 1 [S8] and BS-EN 10 045-1 which are essentially the same as DIN 50115 and

still differ from ASTM E 23. Investigations comparing ASTM and EN-Standard have

indicated that the results may differ mainly in the upper shelf energy regime. The standard

practice for the qualification of the Charpy impact machine is provided in ASTM E 1236

[S9].

Tests are commonly performed at different temperatures. For ferritic steels a S-shape

curve (low shelf, transition, upper shelf) is obtained for the energy as a function of

temperature, from which the transition temperature at a given energy level and the upper

shelf energy can be determined, Fig. 3.5. Other criteria to be derived from the Charpy test

are lateral expansion and portion of cleavage fracture appearance.

For the initial material state the reference temperature RTNDT is determined on the basis of

the Charpy results. According to ASME Section III NB 2300 [R6] the energy and the lateral

expansion have to be measured with three specimens at each temperature. In some

cases, when the initial values are determined for the surveillance program, not always

three specimens are used at a temperature but the RTNDT is derived from a full Charpy

energy/temperature curve. For this case ASME III, NB 2300 gives some guidance how to

proceed, however, this guidance was found to be interpreted differently with regard to the

construction of a lower bound curve.

One of the major problems is the fitting of the data for the energy-temperature curve to

obtain a mean curve from which the shift in transition temperature due to neutron

irradiation can be evaluated. In use are eye-ball fitted curves, tangent hyperbolicus or

other fits based on statistical distributions. In all cases the number of specimens and the

distribution of data along the temperature axis are of importance and can affect the criteria

(index temperatures, like T41J) derived from that curves.

Charpy impact tests are performed with load time or load deflection monitoring

(instrumented Charpy test). The load time traces as shown in Fig. 3.5 for typical lower

shelf, transition and upper shelf behaviour can be used to determine the

12

Fig. 3.5: Characteristic behavior of Charpy-V notch energy as a functionof temperature and load deflection curves typical lower shelf,

transition and upper shelf behavior

13

energy in a redundant way and can help to better understand the fracture process. A

Standard procedure on European level for performing instrumented Charpy tests and

requirements for the measurement technique has been proposed by the European

Structural Integrity Society (ESIS) in "Proposed Standard Method for the Instrumented

Charpy-V Impact Test on Metallic Materials" [S10].

3.1.2.3 Drop-Weight Test

The result from the drop-weight test is one of the criteria to determine the Reference

Temperature RTNDT of the unirradiated initial material state. In all countries, except

Russia, drop-weight tests are performed. The basic Standard is ASTM E 208 [S11] from

which in Germany the Stahl-Eisen-Prüfblatt SEP 1325 [S12] was derived. The

requirements in both Standards are corresponding. The specimen used for this test is

mainly the specimen P2, Fig. 3.6.

The technique to apply the crack starter weld and the welding parameters itself may have

strong influence on the results. In 1984, ASTM has changed the requirement to apply the

crack starter weld from a “two-pass“ to a “single-pass“ procedure. It is pointed out in the

ASTM Standard that the results according to the “new“ procedure may not agree with

previous results. It is explicitly stated that the NDT-temperature derived from drop-weight

tests does not depend on specimen orientation. However, it is recommended to use the

same orientation for the whole set of specimens.

3.1.2.4 Fracture Mechanics Test

Linear Elastic Fracture Mechanics (LEFM)

Fracture toughness data are mainly determined in the unirradiated state. The Standards

for surveillance testing (ASTM E 185 and KTA 3203) do not require fracture mechanics

specimens for irradiated materials, however, in some surveillance programs, ½ T CT

specimens (specimen thickness B = ½“ = 12.7 mm) and WOL specimens are used. Shape

and dimensions of specimens are shown in Fig. 3.7 and 3.8.

14

Fig. 3.7: Geometry of fracture mechanics specimen (CT specimen)according to ASTM E 399;CT 1T : B = 1“, CT 25 : B = 25 mm

15

Fig. 3.8: Geometry of the WOL type fracture mechanics specimen, WOL-100

16

In the unirradiated state experimentally determined fracture toughness data of the material

in question may be used instead of the Reference Curve KIR or KIc presented in ASME III

App. G [R7], ASME XI A 4000 [R8] and KTA 3201.2 [R4]. Tests in the linear elastic regime

(LEFM) are performed according to ASTM E 399 [S13]. To obtain valid fracture toughness

data KIc at elevated temperature, this Standard requires large specimens. Therefore great

efforts were undertaken in the past to generate lower bound fracture toughness data from

tests with small specimens like compact tension (CT 10 and ½ T CT) and pre-cracked

Charpy specimens, which are performed in the elastic-plastic regime using the J-Integral

method [L2, L3].

Elastic-Plastic Fracture Mechanics

The J-Integral was introduced to generate quantitative fracture toughness data in the

upper shelf Charpy regime. The Code requirement of minimum toughness of 68 J (50 ft-lb)

in the upper shelf is not adequate to demonstrate quantitatively the safety margin e.g. in

case of pressurized thermal shock. ASTM E 813 [S14] provides guidelines for conducting

the test, measuring crack opening displacement and for evaluation of results. From those

tests a crack resitance curve J-R-curve (plot of J versus stable crack extension ∆a) can be

obtained from which crack initiation values can be derived. Additional guidlines in

connection with fracture mechanics tests in the elastic-plastic regime exist for Crack-Tip

Opening Displacement (CTOD) measurement, ASTM E 1290 [S15], R-Curve

determination, ASTM E 561 [S16] and J-R curve determination, ASTM E 1152 [S17].

Specimens are usually 20 % side grooved, Fig. 3.9, to achieve a higher constraint and

more uniform stable crack growth.

Comparison of results from testing different specimen geometries and sizes showed that

the KIc-values according to ASTM E 813 are geometry and size dependent and do not

provide lower bound data in all respects [L2, L3, L4]. On the basis of this Standard, the

European Structural Integrity Society (ESIS) has developed "Recommendations for

Determining the Fracture Resistance of Ductile Materials"

17

(ESIS PI) [S18]. It includes a procedure to determine the stretch zone width (SZW) which -

applied to the J-R-curve - leads to lower bound crack initiation values as a material

characteristic parameter, independent on size and geometry of the specimen. The SZW is

a very sensitive parameter to be determined with the scanning electron microscope

(SEM) and diffcult to quantify. ESIS PI is applied preferably in the upper shelf regime and

in the upper transition region. Results from large scale specimen testing have

demonstrated, that this method can basically also be applied in the low shelf regime

where it provides lower bound fracture toughness values which can be used instead of KIc-

data obtained from large scale specimens [L3, L4].

ESIS P 2 [S19] is a comprehensive recommendation for determining fracture toughness

parameters in the entire temperature range. It covers the linear elastic and elastic-plastic

regimes as well. The determination of KIc is according to ASTM E 399.

3.1.2.5 Small Scale Specimen Testing

Investigations of irradiated material - mainly in connection with boat samples taken from

actual components or testing of cladding - can often not be performed with standard size

specimens because of the limited amount of available material. In the past, several

institutions have evaluated material properties using small scale specimens. Most

commonly used are small scale tensile and small scale (subsize) Charpy specimens.

Small Scale Tensile Test

Small tensile specimens were often used when testing the heat affected zone (HAZ) of a

weld joint. Using electric discharge machining (EDM) specimens were prepared as shown

in Fig. 3.10. There is no doubt, that strength data obtained with these specimens are

adequate when compared with results from standard size specimens. The elongation at

fracture can also be considered to be in agreement with normal

18

Fig. 3.10: Small scale tensile specimens for HAZ testing

19

Fig. 3.11: Example of test results from HAZ testing with the use of small scaletensile specimens

20

size specimens if the gauge length corresponds to a proportional standard specimen. The

reduction in area at fracture is difficult to determine at these small cross sections (e.g.

0.5 x 2 mm²) and is usually not performed. Testing a sequence of specimens provides

information for base material, HAZ and weld as shown in Fig. 3.11. The type of specimen

used for HAZ testing is not yet standardized.

Small Scale Charpy Test

In order to determine notch impact properties, subsize Charpy specimens were sometimes

used in the past. Those specimens are not proportionally reduced normal size Charpy

specimens, but they are rather different in geometry, particularly with regard to those

parameters (thickness, notch angle and radius) affecting the constraint of the specimen. A

comparison of both types of specimens is shown in Fig. 3.12. It is well known that

geometry and impact velocity have an influence on the absorbed energy, Fig. 3.13, and

also on the height of the transition temperature, Fig. 3.14 [L5, L6].

The testing practice varies in different laboratories. Two main test procedures are in use:

• A 300 J impact machine is used with a low elevation of the hammer to reduce the

stored energy. The energy is then determined from the load time trace. The velocity is

reduced drastically compared with the normal impact machine (v ≈ 6 m/s) and therefore

affects the transition temperature.

• A small impact machine with maximum energy of 20 or 25 J is used which provides a

velocity of about 3 to 4 m/s.

The test results obtained from different test procedures do not agree. Moreover, the main

problem exists in the transferability of results to normal size Charpy specimens. This

conversion is necessary because only toughness properties achieved with those

specimens are specified in the Code.

21

Fig. 3.12: Comparison of specimen geometry of full size Charpy-V and subsize notch impact specimen

22

Fig. 3.13: Comparison of specific impact energy determined with full sizeCharpy-V and subsize impact specimens

23

Fig. 3.14: Parameters affecting the transition temperature in notch impact testing

24

Fig. 3.15: Correlation between the upper shelf energy of full size Charpy-V (USE) andsubsize notch impact specimens (USE*)

25

The upper shelf energy (USE) is not just proportional to the ratio of the ligament areas see

Fig. 3.13. In general, the upper shelf of the small specimens (USE*) can be plotted as a

function of the upper shelf energy of normal size specimens (USE) as shown in Fig. 3.15

[L6]. More difficult is the determination of index temperatures like T41J, T68J and T0.9mm as

they are used for normal size Charpy specimens. There are some correlation functions

[L5, L6], but activities in this field are not yet well coordinated and need urgently some

support.

3.1.2.6 Reconstituted Specimens

The reconstitution technique has become necessary for different reasons. It has been in

use for a long period of time on unirradiated materials to account for a lack of material.

The electron beam (EB) welding technique was applied to produce Charpy, tensile and

several types of fracture mechanics specimens up to 100 mm thick CT-specimens [L6,

L7]. Experience exists with inserts for Charpy specimens as small as 15 mm length.

Recently it became necessary to reconstitute also irradiated material. It was applied for

Charpy and subsize Charpy specimens. The greatest challenge is the requirement to

produce Charpy specimens from broken halves of irradiated surveillance specimens on

which the fracture energy has to be determined for another specimen orientation than

previously designed, example Fig. 3.16. Stud welding and EB welding technique were

both applied [L6, L7, L8, L9]. ASTM E 1253 [S 20] provides some guidance for the

application of the welding technique and specimen quality. From experience the following

factors play a decisive role:

• The heat input must be minimized to assure that any kind of annealing is avoided

during welding.

• The weld joint must have a high quality to avoid fracture in the weld during impact

testing, particularly in the lower transition region

• The weld joint incorporates residual stresses and gradients in strength which may

influence the formation of the plastic zone around the notch. For inserts of the

dimensions 10x10x10 mm³ a significant reduction in fracture energy may occur, caused

by the limitation in growth of the plastic zone, Fig. 3.17.

26

Fig. 3.16: Reconstituted Charpy-V notch specimen with T-L orientation from a brokenhalf of a L-T oriented Charpy specimen

27

Fig. 3.17: Energy losses in Charpy impact testing due to application ofreconstitution technique

Misalignement between insert and weld-on pieces may cause additional stresses in the

specimen.

With regard to the expected more frequent use of the reconstitution technique, more

experience should be build up within cooperative programmes in Europe.

28

3.2 Determination of Neutron Exposure

The neutron exposure has to be determined for the surveillance specimens and the

reactor pressure vessel (RPV). The calculation is focussed on the determination of the

lead factor, which represents the ratio of exposure of the surveillance specimens and the

highest anticipated exposure at the RPV wall. Furthermore absolute exposure numbers

have to be determined in order to apply the trend curves for material property changes

and to extrapolate or interpolate the results from the surveillance program to design life

time (DLT) conditions. A combination of neutron field calculation and neutron dosimetry is

in use [L10]. ASTM E 706 (0) [S21] gives a comprehensive view of a variety of actions

necessary to determine reliably the material state of the RPV on the basis of surveillance

results, Tab. 3.2. This master matrix comprises main tasks like

• guidelines for analysis and interpretation of nuclear reactor surveillance results

• guidelines for application of calculation and neutron dosimetry methods

• sensor set design

• supplementary actions and verification methods.

The verification of neuton exposure data and the transferability of that data to the RPV

was a big issue in the past. International and national programmes were performed in test

reactors to demonstrate the capability of the existing tools used in this field. Experiments

in the Pool Critical Assembly (PCA) for low flux and the Pool Side Facility (PSF) for high

flux in conjunction with the "blind test" performed in the Oak Ridge National Laboratory

(ORNL), USA, were some of the outstanding activities with international participation.

Other investigations with regard to dosimetry and material property changes on mockups

to simulate RPV conditions were performed in the Melusine Reactor in France, VENUS

PWR Core Source in Belgium and the NESDIP PWR cavity in the U.K. A large number of

references on neutron exposure determination and correlation to material damage is

presented in ASTM E 706 (0).

29

Table 3.2: Standard Matrix to Determine Neutron Exposure Accordingto ASTM E 706

• Selection of Critical Areas

− Surveillance Test for Nuclear Reactors

− Practice for Determining Radiation Exposures for Nuclear Reactor Support Structures

• Analysis and Interpretation of Nuclear Reactor Surveillance Results

− Effect of High Energy Neuton Radiation on the Mechanical Properties of Metallic Materials

− Surveillance Test Result Extrapolation

− Displaced Atoms (dpa) Exposure Unit

− Damage Correlation for Reactor Vessel Surveillance

− Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials

• Guidelines for Application of Calculation and Neutron Dosimetry Methods

− Application of Neutron Transport Methods for Reactor Vessel Surveillance

− Application of Neutron Spectrum Adjustment Methods

− Application of ENDF/A Cross Sections and Uncertainty Files

− Application and Analysis of Damage Monitors for Reactor Vessel Surveillance

− Application and Analysis of Temperature Monitors for Reactor Vessel Surveillance

• Sensor Set Design and Irradiation

− Application and Analysis of Radiometric Monitors

− Application and Analysis of Solid State Track Recorder Monitors

− Application and Analysis of Helium Accumulation Fluence

• Supplamentary Actions and Verification

− Supplemental Test Methods for Reactor Vessel Surveillance

− Analysis and Interpretation of Physics Dosimetry Results for Test Reactors

− Benchmark Testing of Reactor Vessel Dosimetry

− Predicting Neutron Radiation Damage to Reactor Vessel Materials

30

3.2.1 Neutron field Calculation

From the answers to the questionnaire, it became obvious, that there exists no Standard

for the use of a specific computer code and a specific nuclear cross section data set. This

reflects generally the spirit of ASTM E 706 which allows the use of different methods as

far as they are verified. Usually, transport codes are applied for the neuton field calculation

as listed in Tab. 3.3. Although no specific indications were given in the answers to the

questionnaire it can be concluded that

• core power distribution, considering burn-up and changes in core loading configuration

• reactor operating history

• complex geometry in the horizontal section with adjustments for determining axial flux

distribution

are taken into account. Different neutron cross section libraries are in use, however,

activities in the past may have led to an upgraded data set in each country due to the

international exchange of nuclear data in this field.

Usually the spectrum is calculated in energy groups ranging from thermal neutrons up to

fast neutrons. Calculations are performed in different degree of detail with regard to the

number of energy groups. It is quite common to use around 50 energy groups. For more

detailed studies, also a higher number of groups is applied. Considering the detailed

modelling of the geometry, reliable nuclear cross section data and a reasonably reduced

number of energy groups, an accuracy in fluence determination of ± 10 % (1 standard

deviation) can be achieved. Essential higher efforts in calculation is probably not adequate

to significantly reduce the error band mentioned above. Those effects have to be covered

by a margin in determining the transition temperature shift ∆T41J as required e. g. in the

U.S. Reg. Guide 1.99 or by upper bound trend curves for ∆T41J determination as

implemented in the German KTA 3203, when best estimate values of the fluence are

used.

31

3.2.2 Exposure Units

Although the calculation of the neutron field is performed over the entire energy range, for

the correlation of neutron exposure with material damage only neutrons above a certain

energy level are taken into account. Differences exist in the choice of the energy threshold

value. Obviously influenced by the U.S. Standards, most European countries use the

threshold of E > 1 MeV whereas for Russian reactors E > 0.5 MeV is used. The role of the

exposure unit “displacement of atoms“ (dpa) is not clearly expressed in the Codes and

Standards. The lead factor is usually determined from the fluence values E > 1 MeV,

however, dpa and E > 0.5 MeV are also used in some cases, compare Tab. 3.3.

The U.S. Reg.Guide 1.99 uses dpa to determine the attenuation of damage through the

vessel wall. Equation 3 in that document

Φx = Φsurface ⋅ e-0.24 ⋅ x,

which describes the fluence Φ at the location x (in inches) depending on the fluence at the

surface Φsurface was derived from a dpa based calculation, although Φ is the fluence

considering neutron energies E > 1 MeV. In parallel, individual dpa calculations may be

used according to the U.S. Code to determine the gradient of damage through the vessel

wall.

According to the German KTA 3203 a dpa calculation s h o u l d be performed, however,

there is no requirement to use dpa in any procedure to evaluate the material state of the

RPV.

The different exposure units - different energy thresholds and dpa - make it difficult to

compare data on an international basis. Due to the influence of the spectrum of the

neutron source and geometric conditions, data cannot simply be converted by a certain

factor. National experience has grown in correlating neutron exposure with material

property changes on an individual basis of neutron energy threshold and surveillance

data. A re-evaluation of surveillance data from the point of view of an uniform exposure

unit would require a tremendous amount of work and would not be

32

feasible in detail for many of the older reactor surveillance data due to the lack of

information on the neutron spectrum. For the future it would be desirable to have both

exposure data reported, fluence E > 1 MeV and dpa, as well.

3.2.3 Neutron Dosimetry

The neutron dosimetry is based on the spectrum calculation, the nuclear cross sections of

monitor materials and the measurement of the absolute activity. Activation and decay

phases have to be thoroughly implemented in the evaluation of data. Depending on the

half life and the reaction energy, a variety of monitor materials are in use. Several

Standards give guidance on the use and evaluation of monitor materials [S22-S35].

As from the answers of the questionnaire, the main isotopes in use are listed in Tab. 3. 4.

In addition to the activation monitors, fission monitors like U-238 and Np-237 are in use.

However, due to the requirements in handling those fission monitors they are not widely

used.

In general, the application of monitor materials is not standardized but depends on the

experience of the individual laboratory. Nuclear data and procedures to evaluate the

activity measurements, accounting for activation and decay time periods, are summarized

in several Standards.

In the past, practices were developed to determine neutron exposure from other materials

than the previously designed neutron monitors. Reasons for that were the need to verify

the axial and azimuthal distribution of fluence or additional measurements to check the

effect of mitigation methods due to changes in the core loading configuration. Scratch

samples were taken from the austenitic cladding from which, after chemical separation,

the activity of Nb or other elements could be determined, see Tab. 3.4. The importance of

measurements from scratch samples could grow in the future, since, after regular

withdrawal of the surveillance capsules, this is reliable way to survey the fluence

accumulation for the proceeding time. Another successfully applied method is the cavity

dosimetry. Dosimeters are easy to install and to retrieve and allow a cycle by cycle follow-

up of the fluence.

33

3.2.4 Evaluation of Surveillance Results

From surveillance testing, data sets of mechanical properties and corresponding neutron

exposure become available. The neutron exposure can be derived from experimental data

as they result from neutron monitors or can be calculated. In the case of ASTM there is no

clear evidence what data have to be used for extrapolating surveillance results to the

RPV. In the German KTA 3203 it is clearly indicated that the calculated data have to be

used, since for one specific reactor the error can be minimized in using the same

procedure for both, the surveillance position and the RPV wall. Systematic errors are

eliminated in calculating the ratio of exposure (lead factor). The dosimetry data are mainly

used to verify the calculation by another independent method.

In some cases it was found that the fluence dependence of change in properties did not

follow the expected trend. It has to be assumed, that the irradiation time may have an

influence on the equilibrium of the precipitation state due to irradiation. This time effect

does not necessarily correspond with the mechanisms discussed in conjunction with the

neutron flux ("dose rate") effect. Therefore, in extrapolating the surveillance results to the

RPV the irradiation time should also be considered in future.

34

3.3 Determination of Irradiation Temperature

The change in material properties due to service conditions is mainly a result of

accumulation of damage and annealing occuring at the same time. The irradiation

temperature is a decisive parameter that controls the equiblibrium of the two processes.

Depending on capsule design features, gamma heating can cause an increase in

specimen temperature higher than that of the RPV wall. To monitor the maximum

temperature, low melting materials are used in the surveillance capsule which give

information about the peak temperature during the radiation cycles. A more representative

information is given by the time average of the coolant temperature considering calculated

values of gamma heating.

ASTM E 185 [S1] gives guidance about temperature monitoring and allows a deviation of

14 K from the expected capsule exposure temperature. With regard to the application of

surveillance data to the RPV wall. U.S. Reg. Guide 1.99 [R1] requires a matching in

temperature of ± 14 K. A more restrictive guidance for evaluating the irradiation

temperature is contained in KTA 3203 [R2]. Deviations in temperature from the average

coolant temperature of more than 5 K have to be considered in the determination of the

shift in transition temperature.

From experience, the melting monitors do not give a reliable information about the long

term specimen temperature since short time overheating due to plant specific measures

during start-up or shut-down can cause melting of the temperature monitors without

having influence on the long range material behaviour. Therefore melting monitors are

only considered to give limits on the upper temperature bound. Temperature monitors are

widely used. ASTM E 1214 [S36] and KTA 3203 [R2] give examples for alloys and the

corresponding melting temperature.

For Russian reactors a method was developed to determine the specimen temperature

from changes in properties of diamond which is used as monitor material in irradiation

capsules.

To determine the temperature of the RPV wall, one usually relies on the coolant

temperature in combination with an estimated temperature increase on the basis of

calculated values for gamma heating. Additionally thermocouples are installed in some

plants to measure the wall temperature directly.

35

Due to different capsule design and capsule locations in the reactor, there are great

differences in the methodologies to evaluate the irradiation temperature. The existing

Standards give only recommendations for certain methods. The evaluation of adequate

temperature of surveillance specimens and RPV wall with regard to the transferability of

the surveillance results is part of the safety analysis.

36

4 Conclusion

In order to asses irradiation effects of RPV steels and welds of light water reactors there

are three areas of information:

− determination of mechanical properties

− determination of neutron exposure

− determination of irradiation temperature.

The data necessary for the safety analysis are determined by means of surveillance

programmes which differ in many respects as stated in AMES Report No 4 [L1]. For this

report on Standards, the American and the German Standards were studied in detail.

Other information was taken from answers to a questionnaire which were provided by

correspondents from different European countries, members of AMES. It became obvious

that the ASTM Standards are dominating in Europe and are mainly used directly or that

the national Standards in use correspond quite well with the ASTM Standards. ASTM

Standards cover all major areas of surveillance testing whereas in Germany there exist

two different levels. The requirements to perform surveillance tests are on a level of

Regulatory Requirements, the testing itself is on the level of Standards. In Europe, actions

to harmonize test Standards are under way. They will lead to EN-Standards. For some of

the tests mentioned in Chapter 3, Standards already exist, e.g. for tensile testing EN 10

002 [S4] and Charpy testing EN 10 045-1 [S8]. In the field of fracture mechanics, the

“European Structural Integrity Society“ (ESIS) generates guidelines e.g. for elastic-plastic

fracture mechanics testing and evaluation (ESIS PI and P2) [S18, S19] and instrumented

Charpy test (ESIS draft) [S10] which are also supposed to become EN Standards.

Table 4.1 gives on overview of Standards that are in use for mechanical testing and

neutron dosimetry. In the field of mechanical testing differences between ASTM and other

Standards exist in

− Charpy impact test with regard to geometric faetures of the impact

machine which have influence on the test result

− elastic-plastic fracture mechanics (EPFM) with regard to evaluation of

fracture toughness, while the test procedure is identical.

37

Table 4.1: Use of Standards for surveillance testing according to the evaluationof the questionnaire

Test(Action)

Country1) Standard

B, E, S, GB ASTM E 185

specimen samplingsurveillance programme D KTA 3203

FIN YVL 3.9

B, R, E, S ASTM E 8, E 8 (M), E 21

tensile test FIN, D, GB EN 10 002

R GOST

Charpy-V-notch test B, E, S, ASTM E 23

FIN, D, GB EN 10 045 / DIN 50 115

instrumented Charpy test B, D, R ESIS draft No 10

drop-weight test B, FIN, E, S, GB ASTM E 208

D SEP 1325

B, FIN, D, R, E ASTM E 399

fracture mechanics test(LEFM)

FIN, GB BS 5447

R GOST 25.506

B, FIN, E ASTM E 813

fracture mechanics test(EPFM) B, FIN, D ESIS P1, P2

FIN, R GOST 25.506

38

Table 4.1: Use of Standards for surveillance testing according to the evaluationof the questionnaire (continued)

Test(Action)

Country Standard

small scale tensile test -

small scale Charpy test B, FIN, D DIN 50 115

reconstituted specimens B, FIN, R, E,GB ASTM E 1253

neutron field calculation E ASTM E 706, E 482

neutron dosimetry E, S ASTM E 482, E 560

D DIN 25456

temperature monitors E ASTM E 1214

D KTA 3203

1) B = Belgium E = SpainFIN = Finland S = SwedenD = Germany GB = United KingdomR = Russia

With regard to reconstituted Charpy specimen ASTM E 1253 limits the applicability of this

technique to inserts l ≥ 15 mm to avoid annealing of irradiation damage and the influence

of constraint by the weld joint. This Standard does not cover present needs of inserts with

a minimum lenght of l = 10 mm. Activities are under way in individual laboratories to

39

develop more advanced procedures to overcome the problems pointed out in the ASTM

Standard.

In the area of neutron field calculation and dosimetry measurements with regard to

transferability of surveillance results to the RPV, there exists mainly fundamental

experience on a national basis, benchmarked on an international level. Standards in this

field have the character of recommending procedures and pointing to specific problem

areas rather than requiring and describing techniques for calculations, tests and

measurements. The spirit of ASTM E 706 is, that individual precedures may by applied as

long as they are successfully validated on international experiments and round robin tests.

None of the correspondants has indicated that additional standardization in this field would

be generally necessary. Only with regard to fluence determination from scratch samples,

recommendations were given to standardize this method.

5 Future work

This report has been seen by the AMES Steering Group and will be approved.

In our opinion further work should be undertaken in Europe in the following areas:

− curve fitting of Charpy impact results including optimization of selection of

test temperatures and number of specimens to determine transition

temperature and upper shelf energy

− more use of fracture energy results to determine transition temperature

shift

− development of specifications for the geometry of small scale tensile

specimens with regard to HAZ properties

40

− further work on the specification of the geometry of small scale notch

impact specimens, adequate test apparatus including temperature control

of the specimens and evaluation of the results in comparison and

correlation with full size Charpy specimens

− a consideration of the reconstitution of irradiated specimens having an

insert length of 10 mm for the specific consideration of annealing effects

and factors reducing the upper shelf energy.

41

REFERENCES

STANDARDS (S)

[S1] ASTM E 185-82Standard Practice for Conducting Surveillance Tests for Light-WaterCooled Nuclear Power Reactor Vessels

[S2] ASTM E 8-93Standard Test Methods for Tension Testing of Metallic Materials

[S3] ASTM E 21-92Standard Test Methods for Elevated Temperature Tension Tests ofMetallic Materials

[S4] DIN EN 10 002 part 1Zugversuch, Prüfverfahren bei Raumtemperatur(Tension Test, Test Procedures at Room Temperature)

[S5] ASTM E 8 M-93Standard Test Methods for Tension Testing of Metallic Materials[Metric]

[S6] ASTM E 23-94bStandard Test Method for Notched Bar Impact Testing of MetallicMaterials

[S7] DIN 50 115Kerbschlagbiegeversuch(Notch Impact Test)

[S8] DIN EN 10 045 Teil 1Kerbschlagbiegeversuch nach Charpy(Notch Impact Test acc. to Charpy)

[S9] ASTM E 1236-91Standard Practice for Qualifying Charpy Impact Machines asReference Machines

42

[S10] ESIS Draft 10, 1994Proposed Standard Method for the Instrumented Charpy-V impactTest on Metallic Materials

[S11] ASTM E 208Standard Test Method for Conducting Drop-Weight Test to DetermineNil-Ductility Transition Temperature of Ferritic Steels

[S12] Stahl-Eisen-Prüfblatt 1325 (1982)Fallgewichtsversuch nach W.S. Pellini(Drop-Weight Test acc. to W.S. Pellini)

[S13] ASTM E 399Standard Test Method for Plane-Strain Fracture Toughness of MetallicMaterials

[S14] ASTM E 813-89Standard Test Method for JIC, a Measure of Fracture Toughness

[S15] ASTM E 1290-93Standard Test Method for Crack-Tip Opening Displacement (CTOD)Fracture Toughness Measurement

[S16] ASTM E 561-92aStandard Practice for R-Curve Determination

[S17] ASTM E 1152-87Standard Test Method for Determining J-R Curves

[S18] ESIS PI-92ESIS Recommendations for Determining the Fracture Resistance ofDuctile Materials

[S19] ESIS P2-92ESIS Procedure for Determining the Fracture Behavior of Materials

[S20] ASTM E 1253-88Standard Guide for Reconstitution of Irradiated Charpy Specimens

43

[S21] ASTM E 706-87 (94)Standard Master Matrix for Light-Water Reactor Pressure VesselSurveillance Standards, E 706 (0)

[S22] ASTM E 1005-84Standard Test Method for Application and Analysis of RadiometricMonitors for Reactor Vessel Surveillance, E 706 (IIIA)

[S23] ASTM E 261-90Standard Practice for Determining Neutron Fluence Rate, Fluenceand Spectra by Radioactivation Techniques

[S24] ASTM E 944-89Standard Guide for Application of Neutron Spectrum AdjustmentMethods in Reactor Surveillance, (IIA)

[S25] ASTM E 482-89Standard Guide for Application of Neutron Transport Methods forReactor Vessel Surveillance, E 706 (IID)

[S26] ASTM E 636-83Standard Practice for Conducting Supplemental Surveillance Tests forNuclear Power Reactor Vessels, E 706 (IH)

[S27] ASTM E 560-84Standard Practice for Extrapolating Reactor Vessel SurveillanceDosimetry Results, E 706 (IC)

[S28] ASTM E 261-70Standard Method for Measuring Neutron Flux by RadioactivationTechniques

[S29] DIN 25 456 (Teil 1)NeutronenfluenzmessungBestimmung der Fluenz schneller Neutronen mit Aktivierungs- undSpaltdetektoren(Neutron Fluence Measurement; Fast Neutron Fluence Determinationwith Radioactivation and Fission Detectors)

[S30] DIN 25 456 (Teil 2)

44

NeutronenfluenzmessungBestimmung der Fluenz schneller Neutronen mit Eisen-Aktivierungsdetektoren(Neutron Fluence Measurement; Fast Neutron Fluence Determinationwith Iron Radioactivation Detectors)

[S31] DIN 25 456 (Teil 3)NeutronenfluenzmessungBestimmung der Fluenz schneller Neutronen mit Nickel-Aktivierungsdetektoren(Neutron Fluence Measurement; Fast Neutron Fluence Determinationwith Nickel Radioactivation Detectors)

[S32] DIN 25 456 (Teil 4)NeutronenfluenzmessungBestimmung der Fluenz schneller Neutronen mit Niob-Aktivierungsdetektoren(Neutron Fluence Measurement; Fast Neutron Fluence Determinationwith Niobium Radioactivation Detectors)

[S33] DIN 25 456 (Teil 5)NeutronenfluenzmessungBestimmung der Fluenz schneller Neutronen mit Kupfer-Aktivierungsdetektoren(Neutron Fluence Measurement; Fast Neutron Fluence Determinationwith Copper Radioactivation Detectors)

[S34] DIN 25 456 (Teil 6)NeutronenfluenzmessungBestimmung der Fluenz schneller Neutronen mit Thorium-Spaltdetektoren(Neutron Fluence Measurement; Fast Neutron FluenceDetermination with Thorium Fission Detectors)

[S35] DIN 25 456 (Teil 7)NeutronenfluenzmessungBehandlung von Unsicherheiten bei der Bestimmung der Fluenzschneller Neutronen(Neutron Fluence Measurement; Treatment of Uncertainties with theDetermination of Fast Neutron Fluence)

[S36] ASTM E 1214-87Standard Guide for Use of Melt Wire Temperature Monitors forReactor Vessel Surveillance, E 706 (IIIE)

45

REGULATORY REQUIREMENTS (R)

[R1] United States Regulatory Commission,Regulatory Guide 1.99 Rev. 2 (1988), Radiation Embrittlement ofReactor Vessel Materials

[R2] Sicherheitstechnische Regeln des KTA,KTA 3203, Überwachung der Strahlenversprödung von Werkstoffen desReaktordruckbehälters von Leichtwasserreaktoren (1984)(Surveillance of Irradiation Embrittlement of Light Water Reactors)

[R3] ASME Boiler and Pressure Vessel Code, Section III, Article NB 2000,Material

[R4] Sicherheitstechnische Regeln des KTA,KTA 3201.2, Komponenten des Primärkreises vonLeichtwasserreaktoren, Teil 2: Auslegung, Konstruktion und Berechnung(1984)(Components of the Primary Circuit of Light Water Reactors, Part 2:Design, Construction and Calculation)

[R5] Nuclear Power Plant Pressure Vessels, Construction and Welding FillerMaterials, YVL 3.9 Finnish Center for Radiation and Nuclear Safety

[R6] ASME Boiler and Pressure Vessel Code, Section III Article NB 2300,Fracture Toughness Requirements for Material

[R7] ASME Boiler and Pressure Vessel Code, Section III, App. G,Protection Against Nonductile Failure

[R8] ASME Boiler and Pressure Vessel Code Section XI Article A-4000,Material Properties

LITERATURE (L)

[L1] R. GerardSurvey of National Regulatory Requirements and Identification ofExisting, Planned and Required Standards at European Level Relevantwith Irradiation Damage and Mitigation Methods, Part IAMES Report No. 4 (1995)

46

[L2] U. Eisele, E. RoosEvaluation of Different Fracture-Mechanical J-Integral Initiation Valueswith Regard to their Usability in the Safety Analysis of ComponentsNED 130 (1991) pp. 237-247

[L3] G. Sun, E. Roos, U. EiseleErmittlung zähbruchmechanischer Werkstoffkennwerte an Drei-Punkt-Biegeproben im Übergangsbereich der WerkstoffzähigkeitMat.-wiss. u. Werkstofftechnik 23 (1992), S. 250-259

[L4] U. Eisele, E. Ross, M. Seidenfuss, H. SilcherDetermination of J-Integral-Based Crack Resistance Curves andInitiation Values for the Assessment of Cracked Large-Scale Specimens,ASTM STP 1131 (1982) pp. 37-59

[L5] E. Klausnitzer, H. Kristof, R. LeistnerAssessment of Toughness Behavior of Low Alloy Steels by SubsizeImpact Specimens, 8th Intern. SMiRT Conference, Brussels, Belgium,Aug. 1985

[L6] K. Kußmaul, J. Föhl, T. WeißenbergInvestigation of Materials from the Decommissioned Reactor PressureVessel of Gundremmingen Unit A Power Plant, 11th Intern. SMiRTConference, Tokyo, Japan 18-23. Aug. 1991, Volume F, pp. 237-242

[L7] E. KlausnitzerElektronenstrahlgeschweißte Verbundproben für die mechanischeWerkstoffprüfungWerkstofftechnik J. of Material Technology 7 (1976), S. 357-364

[L8] E. Van Walle, A. Fabry, Th. Van Ransbeek, J.L. Puzzolante,W. Vandermeulen, J. Van de VeldeThe Reconstitution of Small Remnant Parts ofCharpy-V-Specimens11th Intern. SMiRT Conference, Post SMiRT Seminar No. 2, Taipei,Taiwan, August 1991

[L9] W. Vandermeulen, F. Fabry, J.L. Puzzolante, J. Van de Velde,Th. Van Ransbeek, E. Van WalleCharpy Specimen Reconstitution as a Means of Providing Data forLicencing PurposesInt. J. Pres. Ves. & Piping 54 (1993) pp. 89-98

47

[L10] J. Bros, A. BallesterosDosimetry and Neutron Transport Methods for Reactor VesselSurveillance, Techn. Report COSU-CT 94-0073-ES