Astm stp 1046 v2 1990

Contents Overview 1 PRESSURE VESSEL STEELS DOSE RATE EFFECTS Effects of 50°C Surveillance and Test Reactor Irradiations on Ferritic Pressure Vessel Steel Embrittlement—RANDY K.. "Dose R

Effects of Radiation

on Materials volume n FOURTEENTH INTERNATIONAL SYMPOSIUM

STP 1046

STP 1046

Effects of Radiation

on Materials:

14th International Symposium (Volume II)

N H Packan, R E Stoller, and A S Kumar, editors

#

ASTM

1916 Race Street Philadelphia, PA 19103

Library of Congress Cataloging-in-Publication Data

Effects of radiation on materials: 14th international symposium /

N H Packan, R E Stoller, and A S Kumar, editors

(STP; 1046)

Papers from the Fourteenth International Symposium on Effects of Radiation on Materials, held June 27-30, 1988 in Andover, Mass and sponsored by ASTM

Committee E-10 on Nuclear Technology and Applications

Includes bibliographies and index

ISBN 0-8031-1266-1

1 Materials—Effect of radiation on—Congresses I Packan, N H (Nicholas H.) II Stoller, R E (Roger E.), 1951- III Kumar, A S (Arvind S.) IV International

Symposium on Effects of Radiation on Materials (14th: 1988: Andover, Mass.) V

ASTM Committee E-10 on Nuclear Technology and Applications VI Series: ASTM special technical publication; 1046

TA418.6.E334 1990 89-18449 620.1'1228—dc20 CIP

Copyright e by AMERICAN SOCIETY FOR TESTING AND MATERIALS 1990

NOTE The Society is not responsible, as a body, for the statements and opinions advanced in this publication

Peer Review Policy Each paper published in this volume was evaluated by three peer reviewers The authors addressed all of the reviewers' comments to the satisfaction of both the technical editor(s) and the ASTM Committee on Publications

The quality of the papers in this publication reflects not only the obvious efforts of the authors and the technical editor(s), but also the work of these peer reviewers The ASTM Committee on Publications acknowledges with appreciation their dedication and contribution

of time and effort on behalf of ASTM

Printed in Baltimore MD May 1W0

Contents

Overview 1

PRESSURE VESSEL STEELS DOSE RATE EFFECTS

Effects of 50°C Surveillance and Test Reactor Irradiations on Ferritic Pressure

Vessel Steel Embrittlement—RANDY K NANSTAD, SHAFIK K ISKANDER,

ARTHUR F ROWCLIFFE, WILLIAM R CORWIN, AND G ROBERT ODETTE 5

Application of a United Kingdom Magnox Steel Irradiation Model to the HFIR

Tensile Properties of Neutron Irradiated A212B Pressure Vessel Steel—

MARGARET L HAMILTON AND HOWARD L HEINISCH 45

Experimental Assessments of Gundremmingen RPV Archival Material for Fluence

Rate Effects Studies—J RUSSELL HAWTHORNE AND ALLEN L HISER 55 Investigation of Materials from a Decommissioned Reactor Pressure Vessel—

A Contribution to the Understanding of Irradiation Embrittlement—

KARL KUSSMAUL, JURGEN FOHL, AND THOMAS WEISSENBERG 80

PRESSURE VESSEL STEELS FINE PRECIPITATES AND DEFECT CLUSTERS

Fine-Scale Microstructural Characterization of Pressure Vessel Steels and Related

Materials Using APFIM—MICHAEL K MILLER AND MARY GRACE BURKE 107

An Analysis of Small Clusters Formed in Thermally Aged and Irradiated FeCu

and FeCuNi Model Alloys—JOHN T BUSWELL, COLIN A ENGLISH,

MARK G HETHERINGTON, WILLIAM J PHYTHIAN, GEORGE D W SMITH,

AND GEOFFREY M WORRALL 127

SANS and DENS Study of Irradiation Damage in a Reactor Pressure Vessel

Material with a Systematic Variation of Irradiation Dose and Heat

Treatments—GEORGES SOLT, FRIEDRICH FRISIUS, WALDEMAR B WAEBER,

AND WILLI BUHRER 154

Magnetoacoustic and Barkhausen Emission Studies of Neutron Irradiated Iron and

Iron-Copper Alloys—EDWARD A LITTLE, DAVID J BUTTLE,

AND CHRISTOPHER B SCRUBY 165

Precipitation in an Aged Fe-0.85 at%Cu Alloy Observed by Muon Spin Rotation

Spectroscopy: A Prospective Method of Studying Irradiation Hardening—

GEORGES SOLT, WALDEMAR B WAEBER, ULR1CH ZIMMERMANN,

PHILIP TIPPING, FREDY N GYGAX, BASSAM HITTI, ALEXANDER SCHENCK.,

AND PETER A BEAVEN 180

PRESSURE VESSEL STEELS

WELDS

An Evaluation of Linde 80 Submerged-Arc Weld Metal Charpy Data Irradiated in

the HSST Program— ARTHUR L LOWE, JR 201 Effects of Radiation on Klc Curves for High Copper Welds— RANDY K NANSTAD,

DONALD E MCCABE, BLAINE H MENKE, SHAFIK K ISKANDER, AND

FAHMY M HAGGAG 214

Improved Correlations for Predicting the Effects of Neutron Radiation on

Linde 80 Submerged-Arc Weld Metals— ARTHUR L LOWE, JR

AND JAMES W PEGRAM 234

PRESSURE VESSEL STEELS FRACTURE TOUGHNESS

Fracture Toughness Shifts and Bounds for Irradiated Reactor Pressure Vessel

Materials— WILLIAM L SERVER AND TIMOTHY J GRIESBACH 253 Comparison of Experimental 41J Shifts with the Predictions of German

KTA 3203 and U.S NRC Regulatory Guide 1.99— DIETER BELLMANN AND

JURGEN AHLF 265

The Nil-Ductility Temperature Shift Arising from Irradiation as Predicted

Through the French Test Reactor Experiments— DOMINIQUE MIANNAY,

DANIEL DUSSARTE, AND PIERRE SOULAT 284

The Effects of Mechanical Stress Gradients on Irradiation-Induced Embrittlement

of Pressure Vessel Steel— JAMES F STUBBINS, ABDERRAFI M OUGOUAG,

AND JOHN G WILLIAMS 305

The Effect of Nickel in Irradiation Hardening of Pressure Vessel Steels—

G ROBERT ODETTE AND G E LUCAS 323

Fracture Resistance of Irradiated Stainless Steel Clad Vessels— DONALD E MCCABE 348 Tensile and Charpy Impact Behavior of an Irradiated Three-Wire Series-Arc

Stainless Steel Cladding— FAHMY M HAGGAG, WILLIAM R CORWIN,

DAVID J ALEXANDER, AND RANDY K NANSTAD 361

Correlation of C, and Drop-Weight Transition Temperature Increase Caused by

Irradiation— FRANZ J SCHMITT 373

MECHANICAL PROPERTIES

Residual Tensile Properties at Low and High Strain Rates of A1SI 316H

Predamaged by Creep, Low Cycle Fatigue, and Irradiation to 2 dpa—

CARLO ALBERTINI, KUNIHIRO IIDA, ANGELO DEL GRANDE, MARIO FORLANI,

ALBERTO PACHERA, AND MARIO MONTAGNANI 387

Response of Ferritic/Martensitic Steels to Neutrons at Irradiation Temperatures

from 20 to 823 K— AKIRA KOHYAMA, KAZUSI HAMADA, KENTARO ASAKURA,

Serrated Flow in Irradiated and Partially Denitrided Mild Steel— K LINGA MURTY

Effects of High Thermal and High Fast Fluences on the Mechanical Properties of

Type 6061 Aluminum on the HFBR— JOHN R WEEKS, CARL J CZAJKOWSKI,

AND PAUL R TICHLER 441

Charpy Impact Test Results of Ferritic Alloys Irradiated to 10 dpa at 55°C—

WAN-LIANG HU AND DAVID S GELLES 453

Fracture Behavior of Ferritic Steels Irradiated at 50°C in the High-Flux Isotope

Reactor (HFIR)— FAN-HSIUNG HUANG 459 Evaluation of the Fracture Toughness of Irradiated Stainless Steel Using Short

Rod Specimens—w L CLARKE, M A WHITE, AND S RANGANATH 470 Effect of Specimen Size on the Upper Shelf Energy of Ferritic Steels—

ARVIND S KUMAR, FRANK A GARNER, AND MARGARET L HAMILTON 487

Evaluation of Irradiation Embrittlement by Instrumented Impact Testing—

SUBRATA CHATTERJEE, S ANANTHARAMAN, U K VISWANATHAN,

AND K S SIVARAMAKRISHNAN 496

Evaluation of Ring Tensile Test Results—A Semi-empirical Approach—

SUBRATA CHATTERJEE, S ANANTHARAMAN, K.S BALAKRISHNAN, AND

K.S SIVARAMAKRISHNAN 515

IRRADIATION CREEP AND SWELLING

Irradiation Creep Behavior of the Fusion Heats of HT9 and Modified 9Cr-lMo

Steels—RAYMOND J PUIGH AND FRANK A GARNER 527

Irradiation Creep in Austenitic Stainless Steels at 60 to 400°C with a Fusion

Reactor Helium to dpa Ratio— MARTIN L GROSSBECK, LOUIS K MANSUR, AND

Influence of Thermomechanical Treatment and Environmental History on Creep,

Swelling, and Embrittlement of AISI 316 at 400°C and 130 dpa—

DOUGLAS L PORTER, ELON L WOOD, AND FRANK A GARNER 551

Swelling and In-Pile Creep Behavior of Some 15Crl5NiTi Stainless Steels in the

Temperature Range 400 to 600°C— KORNELIUS HERSCHBACH,

WALTER SCHNEIDER, AND HANS-J BERGMANN 570

The Cantilever Beam Method for Simulating Irradiation Creep and Growth—

JOHN R PARSONS AND CARL W HOELKE 588

Effects of Neutron Irradiation to 98 dpa on the Swelling of Various Copper

Alloys—HOWARD R BRAGER AND FRANK A GARNER 599

DAMAGE FACILITIES AND DOS1METRY

EUR AC: A Concept for a European Accelerator Neutron Source—

WALTER KLEY, GEORGE R BISHOP, AMAR S1NHA, AND JOSE MANUEL PERLADO 607

Analysis of Transmutation Nuclide Production in a Spallation Neutron Source—

AMAR SINHA AND MAHADEVA SRINIVASAN 623

Radiation Damage Calculations for Compound Materials—

LAWRENCE R GREENWOOD 633

Neutron Spectra Calculations for Ex-Core Irradiation Experiments at the Buffalo

Reactor— G PRILLINGER, EMMERT D MCGARRY,

AND J RUSSELL HAWTHORNE 642

Consideration of Calculated and Experimental Neutron Dosimetry for

Out-of-Core Metallurgy Irradiation Experiments at the Buffalo Reactor—

EMMERT D MCGARRY, J ROGERS, G PRILLINGER, AND

J RUSSELL HAWTHORNE 657

BREEDER CORE MATERIALS

Neutron-Induced Swelling of Commercial Alloys at Very High Exposures—

FRANK A GARNER AND DAVID S GELLES 673

Effect of Neutron Radiation on Mechanical Properties of Permanent Near Core

Structures— ALI ASGHAR TAVASSOLI 684 Influence of Swelling on Irradiated CW Titanium Modified 316 Embrittlement—

ANTOINE FISSOLO, RICHARD CAUVIN, JEAN-PIERRE HUGOT, AND VIVIANE LEVY 700

The Influence of Creep Damage and Neutron Irradiation on the Tensile Properties

of Type 304 Stainless Steel— BOB VAN DER SCHAAF 714 Simulated Transient Behavior of HT9 Cladding— N SCOTT CANNON,

FAN-HSIUNG HUANG, AND MARGARET L HAMILTON 729

The Swelling Behavior of Titanium-Stabilized Austenitic Steels Used as Structural

Materials of Fissile Subassemblies in Phenix— JEAN-LOUIS SERAN, H TOURON,

A MAILLARD, P DUBUISSON, J P HUGOT, E LE BOULBIN, P BLANCHARD,

Irradiation Creep and Swelling of OX16H15M3B Steel and Its Modification

OX16H15M3BP Steel— VALENTIN K SHAMARDIN, VICTOR N GOLOVANOV,

ALEXANDER V POVSTYANKO, VICTOR S NEUSTROEV, YURY K BIBILASHVILI,

IGOR S GOLOVNIN, GALINA V KALASHNIK, AND VALERY V ROMANEEV 753

FUELS AND CERAMICS

Operational Behavior of FBR Mixed Oxide Fuel Pins— WILLI K BIERMANN,

KARL EHRLICH, AND GUNTER MUHLING 769

In Situ High Voltage Electron Microscopy Investigation of Catastrophic Swelling in

Uranium Intermetallic Fuels— ROBERT C BIRTCHER, CHARLES W ALLEN,

GERARD L HOFMAN, AND LYNN E REHN 782 The Effect of Crystal Structure Stability on Swelling in Intermetallic Uranium

Compounds— JEFFREY REST, GERARD L HOFMAN, AND ROBERT C BIRTCHER 789 Index 813

Overview

The Fourteenth International Symposium on Effects of Radiation on Materials was held

on 27-30 June 1988 in Andover, MA This biennial symposium series commenced in 1956 and has served as a major international forum for the exchange and discussion of both the fundamental and technological aspects of behavioral changes in materials exposed to radia- tion environments The high level of participation at the latest symposium required four full days of conference sessions, and the peer-reviewed proceedings are being published in three volumes

The papers from the first three days of the symposium appear in the two volumes of this ASTM Special Technical Publication (STP) 1046 Volume I encompasses radiation damage- induced microstructures; point defect, solute, and gas atom effects; atomic-level measure- ment techniques; and applications of theory Volume II includes mechanical behavior, all papers dealing with pressure-vessel steels, breeder reactor components, dosimetry, and nuclear fuels The fourth day of the symposium was devoted to the single topic of reduced- activation materials, including austenitic, ferritic, and vanadium alloys, for future fusion reactors; these papers are being published in a companion volume: ASTM STP 1047, Re- duced-Activation Materials for Fusion Reactors

The first portion of the second volume of this STP is of vital importance to a critical component of today's nuclear reactors: Pressure Vessel Steels This subject has figured prominently in the previous symposia of this series over the years, and yet, far from being exhausted, new issues continually appear Two such new and "hot" topics were the targets

of five papers each at the Fourteenth Symposium "Dose Rate Effects" contains papers (one of them followed by a conference-presented "Discussion") that cover recent findings

of unexpected embrittlement in pressure vessel steels subjected to long-term neutron ex- posures in two reactors: the High Flux Isotope Reactor (HFIR) in Oak Ridge and the decommissioned Gundremmingen boiling water reactor in Germany The new studies reveal that there is a dependence of embrittlement (for example, raised ductile-to-brittle transition temperature) not only on the total neutron fluence but on the neutron flux (fluence rate)

as well The problem has important implications for current guidelines on long-term irra- diation service of pressure vessel materials

"Fine Precipitates and Defect Clusters" includes detailed studies of the fine copper-rich precipitates that have been found in irradiation-embrittled pressure vessel steels that contain residual levels of copper The papers present the applications of a variety of state-of-the- art probes, including atom probe field ion microscopy, small angle neutron scattering, high resolution transmission electron microscopy, muon spin rotation spectroscopy, and magneto- acoustic and Barkhausen emission techniques These efforts are resulting in a fuller under- standing of the microstructure and chemistry of these ultrafine features that apparently

STP1046-EB/May 1990

Copyright © 1990 by ASTM International www.astm.org

2 EFFECTS OF RADIATION ON MATERIALS

control the mechanical behavior The Pressure Vessel Steels section also includes new studies treating the critical subjects of "Welds" (3 papers) and "Fracture Toughness" (7 papers) The papers in the Mechanical Properties section deal with various materials, particularly ferritic and austenitic steels Tension and Charpy impact tests are the major investigative methods The last four papers focus on new or improved techniques for deriving mechanical behavior information from small amounts of irradiated material

A number of papers in Irradiation Creep and Swelling explore the critical interaction between stress and swelling There are also new irradiation creep results at high doses (130 dpa), lower temperatures (323 to 673 K, 60 to 400°C) and in mixed-spectrum reactors for

a fusion-like helium/damage ratio A novel simulation experiment for studying irradiation creep and growth using ion irradiation of thin cantilever beam specimens is also described The section on Damage Facilities and Dosimetry features two papers treating spallation neutron sources that have been proposed for advanced development and testing of fusion first wall materials There are also two papers dealing with neutron dosimetry of the Uni- versity of Buffalo research reactor that has been used extensively for irradiation embrittle- ment studies Finally there is a noteworthy extension of displacement-damage computer calculations into the area of compound materials, particularly ceramics, some of which exhibit significant differences from the prior estimates

The substantial, world-wide investment of time and capital into materials development for fast breeder reactors is apparent in the section on Breeder Core Materials Although this group overlaps substantially some of the earlier topics, such as "Irradiation Creep and Swelling," these papers have in common the attributes of high dose, proof-test investigations

of the breeder environment

The last section of Volume II, Fuels and Ceramics, contains papers dealing with radiation effects in certain nuclear fuel materials Mixed (uranium/plutonium) oxide and uranium intermetallic fuels are treated The latter materials have exhibited micrometer-sized voids, giving rise to anomalous swelling, which is here linked to early amorphization and the subsequent instability of the amorphized materials as irradiation proceeds

Nicolas H Packan

Oak Ridge National Laboratory, Oak Ridge,

TN 37831; symposium chairman and coed- itor

Roger E Stoller

Oak Ridge National Laboratory, Oak Ridge,

TN 37831; symposium vice-chairman and coeditor

Arvind S Kumar

University of Missouri-Rolla, Rolla, MO 65401; symposium vice-chairman and coeditor

Pressure Vessel Steels Dose Rate Effects

Randy K Nanstad,1 Shafik K Iskander,1 Arthur F Rowcliffe,1

Effects of 50°C Surveillance and Test

Reactor Irradiations on Ferritic Pressure

Vessel Steel Embrittlement

REFERENCE: Nanstad, R K., Iskander, S K., Rowcliffe, A R, Corwin, W R., and Odette,

G R., "Effects of 50°C Surveillance and Test Reactor Irradiations on Ferritic Pressure Vessel Steel Embrittlement," Effects of Radiation on Materials: 14th International Symposium (Vol- ume II), ASTM STP 1046, N H Packan, R E Stoller, and A S Kumar, Eds., American Society for Testing and Materials, Philadelphia, 1990, pp 5-29

ABSTRACT: The results of surveillance tests on the High Flux Isotope Reactor (HFIR) pressure vessel at the Oak Ridge National Laboratory (ORNL) revealed that a greater than expected embrittlement had taken place after about 17.5 effective full power years of operation, and an operational assessment program was undertaken to fully evaluate the vessel condition and recommend conditions under which operation could be resumed A research program was undertaken that included irradiating specimens in the Oak Ridge research reactor Specimens

of the A212 grade B vessel shell material were included, along with specimens from a nozzle qualification weld and a submerged arc weld fabricated at ORNL to reproduce the vessel seam weld The results of the surveillance program and the materials research program performed

in support of the evaluation of the HFIR pressure vessel are presented and show the welds

to be more radiation resistant than the A212B An important result that may have implications for power reactors was a higher than expected Charpy V-notch shift for the surveillance materials at relatively low neutron fluences Moreover, to obtain embrittlement (as measured

by either Charpy shift or increase in yield strength) equal to that from the low flux HFIR surveillance program, the neutron fluence (>1 MeV) required in the high flux Oak Ridge Research Reactor is about ten times that experienced by the HFIR surveillance specimens Results of irradiated tensile and annealing experiments are described as well as a discussion

of mechanisms which may be responsible for enhanced hardening at low damage rates

KEY WORDS: surveillance, reactor pressure vessel (RPV), high flux isotope reactor (HFIR), irradiation, Charpy V-notch (CVN), dose rate, neutron flux, neutron fluence, nil-ductility transition temperature (NDT), copper, nozzles, A212 grade B, A105 grade II, A350 grade LF3, Oak Ridge research reactor (ORR), welds, neutron spectrum, displacements per atom, thermal annealing

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory began operation in 1965; it is a high flux research reactor with a maximum neutron flux of 5 x 1015 neutrons/(cm 2 • s) and a power level of 100 MW(t) The core is surrounded by a beryllium

' Leader of Fracture Mechanics Group, research engineer, and leader of Structural Materials Group, respectively, Metals and Ceramics Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831

: Manager of Heavy Section Steel Irradiation Program, Metals and Ceramics, Oak Ridge National Laboratory, Oak Ridge, TN 37831

' Professor, Department of Chemical and Nuclear Engineering, University of Santa Barbara, Santa Barbara, CA

STP1046-EB/May 1990

Copyright © 1990 by ASTM International www.astm.org

6 EFFECTS OF RADIATION ON MATERIALS

reflector and is light water cooled and moderated The entire vessel is submerged in a pool

of water, with 5.2 m (17 ft) of water above the top of the vessel, as shown in Fig 1 In addition to its use as a research reactor, the HFIR is an important source for production

of transplutonium elements such as 2"Cf and 2MEs used in medical and other applications The reactor is housed in a pressure vessel fabricated from 73-mm thick (2.875-in.) A212 grade B (A212B) carbon steel plate with a welded hemispherical bottom head The shell was rolled from a single plate and thus contains a single longitudinal seam weld The vessel

is approximately 2.44 m (8 ft) in diameter and the cylindrical part of the vessel is 3.05-m (10-ft) high The cylindrical part of the shell has a Type 304L (UNS S30403) stainless steel roll bond cladding on the inner surface and a Type 347 (UNS S34700) stainless steel weld deposit cladding on the outer surface It was designed and constructed in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section VIII, and Nuclear Code Cases 1270N, 1271N, and 1273N for an internal pressure

of 6.9 MPa (1000 psi), a temperature of 93°C (200°F), and for 20 effective full power years (EFPY) The vessel was given a postweld heat treatment (PWHT) of 51 h at 510°C (950°F) With the reactor at full power (100 MW), the normal operating pressure and temperature were about 5.2 MPa (750 psi) and 50°C (122°F), respectively

IRRADIATION HYDRAULIC 7

Ml CONTROL DRIVI

FIG 1 — Vertical section of HFIR vessel and core

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 7

There are a large number of nozzles to allow access and for experimental facilities A cross-sectional view is shown in Fig 2 Of interest to this paper are the four beam tube nozzles designated HB-1, -2, -3, or -4, at the core midplane The beam tube nozzles were forged of either A105 grade II (A105II) (UNS K03504) or A350 grade LF3 (A350LF3) (UNS K32025) steels, chosen for their relatively low nil-ductility transition (NDT) temper- atures The nozzles were welded in place using the shielded metal arc process, and the exposed nozzle surface and welds were stainless steel clad by weld deposition These beltline nozzles support aluminum beam tubes reaching near the core through the beryllium reflector The beam tubes allow for neutron streaming to experimental facilities Because they displace water and beryllium, these tubes reduce the primary shielding for the vessel and cause an asymmetric neutron flux distribution These tubes are responsible for the higher embrittle- ment of relatively small areas in the nozzle regions

A surveillance program [/] was incorporated in the vessel operation to monitor potential embrittlement of shell and nozzle base metals The results of recent surveillance tests in- dicated that the vessel operating criterion was exceeded due to radiation embrittlement, and

an operational assessment program [2] was undertaken to recommend conditions under which operation could be resumed There were, however, no weld metal specimens included

in the surveillance program and irradiation data were required for the welds as well as for the A212B in orientations other than those used in the surveillance program A research program was undertaken that included the irradiation of specimens in the Oak Ridge Re- search Reactor (ORR) This paper will discuss the results of the HFIR surveillance program and the materials research program performed in support of the assessment of the HFIR pressure vessel In order to understand the mechanisms of such damage, irradiation and testing of tensile specimens, transmission electron microscopy, small angle neutron scatter- ing, and field ion atom probe examinations were performed to identify the defects responsible for hardening The results from this work will also be described

REACTOR PRESSURE VESSEL

NOZZLE WELD

FIG 2—Cross section of HFIR vessel showing locations of surveillance capsules at key locations 1 through 7

8 EFFECTS OF RADIATION ON MATERIALS

Surveillance Program

The structural components of the reactor pressure vessel were fabricated of ferritic ma- terials, and the design included provisions to minimize neutron exposure Even so, a material surveillance program was established at the time of reactor startup to monitor the extent

of embrittlement [7] Helium filled and welded capsules were prepared each containing three Charpy V-notch (CVN) specimens of base metal from nozzle prolongations and from drop- outs of shell material removed in order to install the nozzles These capsules, each of which also contained a flux monitor, were then placed at various locations known to be subjected

to high fluences, typically near and around the nozzles, as shown in Fig 2 The general locations of the capsules in the HFIR vessel are referred to by key numbers (Keys 1 through 7), and are indicated in Fig 2 The nozzle specimens are located around each beam tube Key locations 1 and 4 are in symmetric locations, as are those from Keys 6 and 7, and the results from these locations have been combined There is, however, considerable variation

in fluences at each of these locations Surveillance for the HFIR consisted of testing spec- imens of each material from two capsules; specimens from one capsule were used to bracket the irradiated NDT while specimens from the second were tested at the estimated NDT The chemical composition and tensile properties of materials used in the HFIR pressure vessel are shown in Tables 1 and 2, respectively (weld chemistries are discussed later) The two nozzle forgings designated HB1 and HB4 in Fig 2 were fabricated from the same heat

of A105II to the same dimensions Forgings HB2 and HB3 were fabricated from one heat but to different dimensions (that is, different amounts of mechanical work) and were eval- uated separately Differences in chemical composition between HB2 and HB3 are considered

to be due to analytical procedures All the materials have generally normal compositions relative to specifications With regard to those elements considered to increase radiation sensitivity under light water reactor conditions, that is, an irradiation temperature of 288°C,

TABLE 1—Chemical composition (wt%) of base metals for HFIR pressure vessel study

A212B A105II A350LF3 A350LF3 A212B Element Shell HB1, HB4 HB2 HB3 EGCR"

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS

TABLE 2— -Vendor-supplied tensile properties of HFIR pressure vessel materials at room temperature

Strength, ksi" Elongation

in 2 in 0 , Reduction of area,

* ORNL tension test results at room temperature, average of two tests

' Broke in base metal

d Requalification test

the A212B is relatively high in copper but low in nickel and phosphorous The A350LF3 is very high, of course, in nickel but fairly low in copper and phosphorous The A105II is quite low in all these elements

Less is known about compositional effects at this low temperature range than at higher temperatures representative of commercial power reactors The HFIR vessel materials are irradiated at about 50°C and researchers have postulated [3,4] that dissolved interstitials such as nitrogen and carbon may be important in low temperature irradiations Nitrogen and oxygen appear to be present in typical concentrations although what concentration levels

of those elements might increase radiation sensitivity are not known Nickel and copper have also been reported as enhancing low temperature irradiation [5] as has manganese in

an interactive role with carbon [6]

The surveillance specimens were removed in the LT orientation (longitudinal axis in the plate rolling direction and crack propagation transverse to the rolling direction) in accordance with the ASME code procedures in effect at the time A sulfur content of 0.040% noted for the A212B in the current evaluation gave rise to concerns about significant orientation effects in the plate, and optical metallography was performed on unirradiated specimens of all base materials Figures 3 and 4 are of unetched and etched A212B micrographs, re- spectively, and reveal the sulfide stringers that lie in the rolling plane These stringers amplified the concern that the toughness in the rolling direction might be significantly lower than that in the transverse direction The microstructure consists generally of ferrite and pearlite consistent with the heat treatment and plate thickness Orientation effects are discussed in more detail in the section on test reactor irradiations The A350LF3 forgings have a ferrite microstructure and Charpy testing revealed no evidence of orientation dif- ferences The A105II forgings have some pearlite in a predominantly ferrite matrix and also show no orientation differences; the grain size is smaller than that of the A212B plate and the amount of pearlite is less

10 EFFECTS OF RADIATION ON MATERIALS

FIG 3—Micrograph of unelched A212B HFIR surveillance material

Indexing of the NDT with Charpy Energy Levels

As described in the original surveillance program [/] the NDT temperature is indexed to

a specific Charpy V-notch energy for a particular material The source of correlations used

in Ref / is not referenced but they were most likely based on work at the Naval Research Laboratory and are defined in a U.S Department of Commerce document [7] Thus, when the shift in NDT is mentioned, it is the shift in the particular energy level as determined from Charpy tests; 20 J (15 ft • lb) for A105II and A212B, 27 J (20 ft • lb) for welds, and

41 J (30 ft • lb) for A350LF3

Surveillance Test Results

Various surveillance capsules were removed for testing after 2.3, 6.4, 15, and 17.5 EFPY The results of the tests for each key location (see Fig 2) are shown in Figs 5 through 9 A summary of the test results is given in Table 3 As mentioned previously, Keys 6 and 7 (A212B) have been combined due to the generally symmetric neutron flux distribution at these locations The listed NDT values were determined by constructing hand drawn curves fit to the mean energy level for each test temperature and corresponding to the appropriate CVN energy level Also shown are the changes in NDT (ANDT) from the unirradiated condition The estimated error of the mean NDT temperature is ±6°C (10°F)

, r>f

FIG 4—Micrograph of etched A212B HFIR surveillance material

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 11

IRRADIATED (11/15/86)(high flux) 4.0xl0 17 n/cm 2 (>1MeV)

-125 -100 -75 -50 -25 0

TEMPERATURE (°C) FIG 5—Surveillance results of Charpy V-notch impact tests for A105 grade II, components HB1 and HB4 (key numbers 1 and 4)

12 EFFECTS OF RADIATION ON MATERIALS

200

-50 -25 0 TEMPERATURE CO

FIG 7—Surveillance results of Charpy V-notch impact tests for A350 grade LF3, component HB3 (key number 3)

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 13

Relatively few specimens were available Although details of earlier tests are unclear, specimens removed after 15 and 17.5 EFPY were tested with a machine calibrated in ac- cordance with ASTM Standard Method for Notched Bar Impact Testing of Metallic Materials (E 23-86) The temperature measuring system was calibrated within 1°C over the test tem- perature range reported in this study The internal consistency of the results is demonstrated

by the NDT shifts of Keys 2 and 3, which, for approximately equal fluences, were 66 and 64°C (118 and 115T), respectively

The surveillance results have also indicated that the slope of the CVN energy versus test temperature curve has not changed significantly Tfns result is consistent with many other research results for pressure vessel steels at such low fluences and was of particular impor- tance to the assessment task because of the scarcity of archival materials for irradiation Figure 10 shows the change in NDT versus EFPY for the nozzles and shell The changes for the nozzles fit well with a linear expression starting from about the 2-EFPY point The shell material only has two data points, and a curve fit is more uncertain The straight line shown is considered to be conservative The data appear to provide a basis on which relatively sort-term extrapolations can be confidently based Figure 11 shows the changes in NDT of the surveillance results relative to fast neutron fluence (>1 MeV) and suggests that the A212B steel exhibits a greater sensitivity to irradiation than the nozzle materials, which all responded similarly This observation is discussed in a later section The figure also shows the HFIR surveillance results compared to low temperature material test reactor (MTR) irradiations with similar materials [8] The HFIR surveillance results with neutron fluxes ranging from ~2 x 108 to 1 x 109 neutrons/(cm2 • s) show an NDT shift for the A212B of 42°C (75°F) at a neutron fluence more than one order of magnitude less than the test reactor irradiations at neutron fluxes in the range 1012 to 10u neutrons/(cm2 • s) This observation implies a significant dose rate (fluence rate) effect and will be discussed in greater detail later

14 EFFECTS OF RADIATION ON MATERIALS

TABLE 3—Summary of NDT temperatures and bNDTfor HFIR surveillance tests

Component,

location Material Number Key

Nil-ductility Transition Temperature", (ANDT, K), for various EFPY °C

0 2.34 6.45 15.01 17.53 HB1, HB4 A105II 1,4 -62 -52 -46 -29 -27s HB4 (low

flux)

HB2 A350LF3 A105II 4 2 -79

(10) -64

(17) -50

(33) -10

(35) -29 (33) -13 HB3 A350LF3 3 -62 -43 (14) -29 (29) (55) -8 (66) 1.7 Shell (IC3) A212B 5 -21 (19) (33) (54) -9.4 (64) Shell (HB1A,

HB4A) A212B 6,7 -21

(11) 8.3 (29) (42) 21

" Estimated error of mean NDT temperatures is ±6°C, ANDTs shown in parentheses represent change from unirradiated condition

* Rough estimate based on severe data scatter

100

\o 15 EFFECTIVE FULL POWER YEARS

FIG 10—Change in NDT versus effective full power years for HFIR pressure vessel surveillance program

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 15

FIG 11—Change in NDT versus neutron fluence (>1 MeV) for HFIR pressure vessel surveillance program compared with similar materials irradiated in material test reactors

shift of the A212B base metal, namely 42°C (75°F), and the customary scatter of CVN data limited the use of data available in the literature A further limitation on the use of previous data was the lack of information on the chemistry of the materials used For example, the deleterious effect of copper on the irradiated behavior of steels was not discovered until about 1970, and most of the published data on A212B base metals and welds predates this discovery Also, other elements such as nickel and phosphorous are now known to affect the irradiated behavior Copper and nickel were not reported unless they were alloying elements Other elements, particularly interstitials such as nitrogen, have been suspected

of enhancing radiation-induced embrittlement at low temperatures [70], while the role of substitutional elements may be important, mainly because they may modify the action of interstitials [3] Also, low temperature and low fluence irradiations have been of lesser interest to those concerned with irradiation effects in light-water reactor (LWR) pressure vessels

The bulk of available literature deals with materials and conditions of relevance to LWRs,

in particular for an irradiation temperature of 288°C (550°F) and fluences of about 1 x 10" neutrons/cm2 (>1 MeV) The HFIR reactor pressure vessel wall operates at about 50°C, and the fluences of interest are on the order of 10" and 1018 neutrons/cm2 (>1 MeV) There was, nevertheless, a significant body of literature for low temperature (less than 150°C) irradiation, but most of it was for higher fluences than those of relevance to the HFIR Possible dose rate effects precluded the direct utilization of such data The culmination of all such considerations was that there was no alternative to actual testing on HFIR-specific materials to answer the outstanding questions

Archival material for welds and base metal were either nonexistent or available in very limited amounts The scarcity of original archival material also precluded the fabrication of

Klc or Ku specimens of a size necessary to yield useful data The primary mechanism chosen

16 EFFECTS OF RADIATION ON MATERIALS

to obtain Klc and KItl data was the use of ASME Code curves in conjunction with CVN data from this test program

To complete the assessment, data were required regarding irradiation effects on the vessel welds as well as for specimen orientations in the A212B shell other than that represented

in the surveillance program (LT, meaning a specimen with its longitudinal axis in the plate rolling direction and crack propagation transverse to the rolling direction) Specimens of A212B were tested to evaluate the propensity for crack propagation in the circumferential and axial orientations along the surface of the vessel (LT and TL, respectively), as well as into the thickness of the vessel (LS and TS, respectively)

The most important base metal was the A212B used for the shell because calculations had indicated that relatively small areas near the nozzles would limit the life extension for the reactor pressure vessel A small qualification weldment was available for evaluating the nozzle welds None was available for the seam weld, however, and a reproduction of the weld was fabricated

Description of Weld Metals

All four nozzles were welded to the vessel shell using 4.76-mm diameter (3/i6-in.) shielded metal arc carbon steel class E7018A1 electrodes An archival weld qualification block pro- duced by the vessel fabricator was available and used for this evaluation The block was about 75-mm thick, 225-mm wide, and 225-mm long (in the direction of welding) and clad

on both sides in the same manner as the vessel, and it was composed of A212B welded to A350LF3 with a single-J geometry and with the angled side lying in the A212B Following welding, the block was given a PWHT of 17 h at 510°C All test specimens were removed with the axis of the specimen perpendicular to the welding direction and crack propagation

in the welding direction The root region of the weld was not used to obtain test specimens for the irradiation program Charpy tests showed an NDT of - 18°C, which agreed very well with the previous results reported by the vendor

As stated earlier, no archive weldments were available which represented the vessel longitudinal seam and circumferential girth welds Drillings were removed from the vessel seam weld and the chemical composition was determined [2], with the results shown in Table 4 Linde 40 [American Welding Society (AWS) Type EA3] 4.76-mm diameter (Vie- in.) copper-coated weld wire and Linde 80 flux were used to reproduce the vessel weld as closely as possible using information from original weld documentation A weld pad was prepared and showed the copper content to be much lower than that of the vessel weld A double-J weld groove was prepared in an A212B dropout from the HFIR bottom head and 6.35-mm diameter ('/i-in.) E7018A1 shielded metal arc rod was used in two layers in the root region Copper was added to the weld by tack welding a length of thin 0.229-mm (~0.009-in.) copper wire in the weld groove prior to each weld pass The weldment was completed with 30 passes of submerged arc weld and given an initial PWHT of 17 h at 510°C Chemical analyses (Table 4) showed a copper content very close to those of the vessel seam weld Microprobe analyses across the weld confirmed that violent mixing action in the weld puddle uniformly distributed melted copper throughout the weld beads

Charpy specimens were removed from the bulk area of the weld in the same manner as described previously, and test results showed excellent agreement with the results reported for the original qualification weld The remainder of the weld was given an additional PWHT

of 34 h at 950°F to reproduce the conditions of the actual seam weld in the vessel and CVN tests showed an unirradiated NDT of - 21°C

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 17 TABLE 4—Weld metal chemical compositions (wt%)

HFIR Seam Weld Drillings Seam

Weld Qualification Nozzle

Charpy Impact Results

With regard to the effect of specimen orientation on the CVN energy, Fig 12 compares the results of tests performed on the various orientations of A212B shell material Particularly noticeable are the differences between the TL and LS orientations The former orientation barely attains 54 J (40 ft • lb) at 49°C (120°F), while the latter exceeded 272 J (200 ft • lb)

at 10°C (50°F) As shown previously in Figs 3 and 4, large sulfide stringers exist in the microstructure and provide easy paths for crack propagation in the rolling (longitudinal) direction

The materials included in the ORR irradiation program were the A212B plate in the LT,

TL, and TS orientations, the nozzle qualification weld metal, and the reproduction seam weld fabricated at ORNL The A212B (LT) specimens were selected from a group of spare unirradiated surveillance specimens, while the A212B (TL) and A212B (TS) were machined from a remaining portion of a nozzle dropout All A212B specimens were removed from the same depth in the plate thickness as were the original surveillance specimens

Specimens were irradiated in core edge position A9 of the ORR Reactor coolant water flowed directly over the specimens contained in open capsules Each capsule held 20 CVN specimens and included flux monitor gradient wires To minimize corrosion, specimens were coated with a stable black oxide through a caustic anodizing treatment; tests of unirradiated coated specimens showed no effects of the oxide coating on CVN toughness Irradiation temperature for the specimens was determined to be about 49.5°C [2]

The neutron flux in the core of the ORR, ~1 x 1013 neutrons/(cm2 • s) is about four to five orders of magnitude greater than exists at the HFIR pressure vessel wall A dosimetry capsule containing flux monitors and eight CVN specimens of a different heat of A212B

18 EFFECTS OF RADIATION ON MATERIALS

(Ref 77), (see Table 1 for chemical composition) received a neutron fluence of 1.54 X 10" neutrons/cm2 (>1 MeV) in a 4.1-h exposure From the CVN tests on that particular A212B material, which exhibited a ANDT of ~10°C, a fluence of about 1 x 1018 neutrons/cm2 was predicted to produce the same degree of embrittlement exhibited by the HFIR surveillance specimens at an exposure of only 1.3 x 10" neutrons/cm2 (>1 MeV) The irradiation time for the remaining three capsules was selected to result in a slightly greater shift than de- termined for the HFIR A212B surveillance specimens to allow for extrapolation for the extended service predictions

Results of the ORR irradiations of A212B in three orientations are shown in Figs 13 through 15, while Figs 16 and 17 show results for the seam weld reproduction and the nozzle qualification weld, respectively Noting that all the ORR irradiations were performed

to the same fluence, one can see that the NDT shifts for welds are less than or equal to that for base metals

The curves shown were generally fit to the mean energy values at each test temperature

In the case of the A212B (TS), however, the three high energy values for the irradiated condition were ignored For the seam weld data, which show two test results far outside the mean values for their respective groups, the high point was ignored, but the low point, even though determined to be an outlier, was included in mean energy determinations Those procedures were adopted as part of the conservative approach to the overall assess- ment of vessel integrity Examination of the seam weld outlier specimens revealed no plausible explanation for the results obtained For the A212B (TS), the large scatter was concluded to be the result of secondary cracking near sulfide stringers that were observed

on the fracture surfaces transverse to the direction of primary crack propagation

Table 5 summarizes the results of the ORR CVN irradiation program and shows that under the same irradiation conditions and at the same exposure level, the HFIR vessel welds exhibit lesser shifts in NDT (CVN shift) than does the A212B steel Regarding orientation

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 19

20 EFFECTS OF RADIATION ON MATERIALS

in the ORR

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 21

6.92 X 10"° 3.44 X IO' 6 5.12 x IO" 5 6.45 17 4.89 X 10" 7.26 X IO" 13 9.9 X IO' 6 1.48 x IO" 4 15.01 33 4.89 X 10 s 7.26 X io-' 3

2.31 X 10" 3.44 x IO 4 17.53 33 3.35 X 10* 4.89 X io- l3

1.85 X 10" 2.70 x IO 4 17.53 35 7.27 X 10* 1.08 X IO" 12 4.01 X 10" 5.97 x IO' 4

79 2.34 14 1.1 X 10" 1.58 X io- 12

8.20 X IO 16 1.16 x IO' 4 6.45 29 1.1 X 10' 1.58 X io- 12

2.26 X 10" 3.21 x IO- 4 15.01 56 1.1 X 10" 1.58 X IO 12 5.26 X 10" 7.47 x IO- 4 17.53 66 1.1 X 10" 1.58 X IO 12 6.14 X 10" 8.73 x IO' 4

62 2.34 19 1.29 X 10" 1.85 X io- 12

9.55 X IO' 6 1.36 x IO- 4 6.45 33 1.40 X 10" 2.01 X io-' 2

2.84 X 10" 4.09 x IO 4 15.01 54 1.03 X 10" 1.48 X IO" 12 4.88 X 10" 7.01 x IO" 4 17.53 64 1.29 X 10' 1.85 X IO"' 2 7.12 X 10" 1.02 x IO" 3

" Not irradiated ANDT assumed equal to A212B (LT)

'' Material (item 157) from previous ORNL study on an experimental gas cooled reactor, Ref //, dpa/s estimated from other experiments conducted in poolside facility of ORR

' Dosimetry information for these specimens not available, flux is estimated

22 EFFECTS OF RADIATION ON MATERIALS

effects in the vessel shell, the TS- and TL-oriented specimens showed equal or less ANDT than did the LT-oriented specimens used in the HFIR surveillance program Table 5 also summarizes, for both the ORR irradiations and HFIR surveillance, exposure conditions in terms of neutron flux (>1 MeV) and displacements per atom rate (dpa/s) as well as total exposures in terms of fast fluence (>1 MeV) and dpa (>0.1 MeV) Appropriate ASTM standards were used to determine dosimeter activities, and the first 38 groups of the ELXSIR cross section library were used to calculate dpa as described in Ref 2

The results from the ORR irradiations tend to confirm the suspected rate effect mentioned earlier Figure 18 compares the NDT increases of the HFIR A212B surveillance specimens with those from the A212B irradiations in the ORR The specimens irradiated in HFIR at

a low flux exhibited similar embrittlement at a fluence level at least one order of magnitude smaller than that required in the ORR at a much higher flux The effect is further substan- tiated in Fig 11, which shows that the HFIR A212B steel irradiated in the ORR behaved similar to other materials irradiated in MTRs under similar conditions

If the results are examined relative to displacements per atom, Fig 19 shows the rate effect is very similar to that shown relative to fast fluence A rigorous analysis [2] of the HFIR and ORR spectra (>0.1 MeV) in the vicinity of the A212B surveillance specimens revealed that while the HFIR spectrum contains a somewhat larger fraction of high energy neutrons (above about 2 to 3 MeV) than the ORR spectrum, the displacements per atom per average fast (>1 MeV) neutron are about equal at both reactor test locations Thus, any significant effects of spectrum must be associated with the softer regions of the spectrum (that is, <0.1 MeV) The potential role of low energy neutrons under these conditions is discussed in Ref 9 and in the following section on mechanisms

This apparent dose rate effect is discussed in terms of possible radiation damage mech- anisms in the next section Its implications for various components of power reactor plants subjected to low temperature and low fluxes, such as reactor pressure vessel supports, could

be important in terms of unexpectedly greater irradiation embrittlement [9,72]

io ,e z » io' r l D io ' ' io"

NEUTRON FLUENCE, >MeV In/tan 2 )

FIG 18—Comparison of HFIR A212B surveillance data with ORR irradiations of the HFIR A212B and a similar A212B steel (Ref 11)

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 23

' MTR DATA OF STEELE and HAWTHORNE^

'(ESTIMATED, IRRADIATED AT<95°C) A212B (HFIR ARCHIVES)

DISPLACEMENTS PER ATOM (E>0.1MeV)

FIG 19—Change in NDT vs displacements per atom (>0.1 MeV) for HFIR pressure vessel materials irradiated in the HFIR surveillance program at low fluxes compared with the HFIR A212B irradiated in the ORR at high fluxes and MTR trend curve of Steele and Hawthorne (dpa estimated for MTR data using ORR spectrum)

Investigation of Radiation Damage Mechanisms

The HFIR pressure vessel represents a relatively unexplored regime of radiation effects

in which a low irradiation temperature (~50°C) is combined with a low neutron flux [—2.4 x 108

neutrons/(cm2 • s) (>1 MeV)] A preliminary assessment of the possible mechanisms that could contribute to the unexpected level of embrittlement of the HFIR A212B material was carried out The assessment included yield stress measurements, postirradiation annealing studies, and attempts to identify the defects responsible for the hardening Further, an additional irradiation experiment was carried out in the ORR to determine the hardening behavior of A212B in a relatively high flux environment

Experimental Procedure

Flat tensile specimens, 25-mm long by 0.75-mm thick with a gage width of 1.5 mm and

a gage length of 7.5 mm, were electrical discharge-machined from tested CVN surveillance specimens The long axis of the tensile specimen was parallel to the long axis of the CVN specimen, and the width of the gage section was parallel to the notch direction Similar unirradiated specimens were prepared from archive CVN specimens and used in a series of ORR irradiation experiments These specimens were given a caustic anodizing treatment

to prevent corrosion during irradiation in the reactor pool water The specimens were suspended at position P8, 10 mm from the core poolside face plate, opposite core lattice position A8 and approximately 30 mm below the core horizontal midplane From dosimetry sensors placed alongside the tensile specimens, flux [neutrons/(cm2 s)] values were deter- mined to be 1.3 x 10'3 (>1 MeV), 2.7 x 1013 (>0.1 MeV), and 9.8 x 1013 (thermal) The displacement rate calculated for energies down to 0.1 MeV was 1.9 x 10~8 dpa/s Following

24 EFFECTS OF RADIATION ON MATERIALS

irradiation, tensile testing was carried out in air at a strain rate of 4 x 10"4/s For the annealing studies, small coupons were cut, using a slow speed diamond saw, from a broken A212B surveillance CVN specimen Anneals were carried out for 1 h in air at temperatures ranging from —200 to 500°C Microhardness measurements were made using a Vickers diamond pyramid indenter with a 500-g load

Results and Discussion

The yield stress data for the A212B irradiated in the HFIR and in the ORR are plotted versus the square root of neutron fluence in Fig 20 For the ORR irradiations, there is an initial incubation or threshold fluence during which very little change in strength occurs This stage is followed by a regime in which yield stress increases in proportion to the square root of fluence, the increase eventually saturating at ~6 x 1018 neutrons/cm2 Although the data for the HFIR environment are limited to two fluences, Fig 20 clearly shows that hardening occurs more rapidly in the HFIR pressure vessel environment than it does in a typical high flux MTR environment In Fig 21, shifts in the ductile-brittle transition tem- perature (DBTT) data are plotted in a similar fashion The higher fluence ORR data point

is taken from the CVN irradiation experiments with HFIR vessel archive material conducted

in the ORR A9 position as described earlier The lower fluence point is from the preliminary ORR irradiation experiment, discussed earlier, which utilized A212B (EGCR) Despite the paucity of data, DBTT clearly increases more rapidly with fluence in the relatively low flux HFIR environment Scanning electron microscope (SEM) fractography of broken CVN specimens from both environments showed similar characteristics, that is, transgranular cleavage in the lower shelf region with a transition to dimpled rupture in the upper shelf region

4.0 WO 9 ) (<t>1, n/cm 2 ) 1/4 (E>1.0MeV)

FIG 20—Increase in yield strength vs the square root of fluence (>1 MeV) for HFIR A212B irradiated

in the HFIR surveillance program at a low flux and in the ORR at a high flux

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 25

(xio 9 ) (<£t, n/cm 2 ) Vz (E>1.0MeV)

FIG 21—Change in ductile-to-brittle transition temperature versus the square root of fluence (>1 MeV) for HFIR A212B irradiated in the HFIR surveillance program at a low flux and in the ORR at a high flux

The literature on low temperature embrittlement often implies a direct relationship be- tween radiation hardening and the CVN transition temperature shift For example, low temperature data reported by Nichols [13] for A302B steels suggested a proportionality between temperature shift (AT) and yield stress increase (ACT,,) as

AT= CACTV with C ~ 0.5°C/MPa

A recent analysis [14] of the elevated temperature data base showed that the propor- tionality factor C depends upon the initial CVN properties and on the magnitude of the yield stress increase, with an average value of about 0.65(±0.15)°C/MPa The values of C obtained from Figs 20 and 21 are 0.67 and 0.57 for HFIR and for ORR irradiations, respectively These values agree well with the literature values and indicate a direct rela- tionship between radiation hardening and DBTT shift in both reactor environments The results of the annealing study conducted on the A212B surveillance material are shown in Fig 22 Each point represents an average of five microhardness measurements The large degree of scatter arises from the duplex nature of the microstructure; varying proportions of ferrite and pearlite are sampled by successive indentations The hardness of the A212B was only increased by —6% during irradiation in the HFIR to a fluence of 1.3 x 1017 neutrons/cm2 Some recovery of the damage is evident after a 1-h anneal at 300°C and appears to be complete after 1 h at 375°C

Postirradiation annealing experiments on ferritic steels reported in the literature give variable results for recovery times and temperatures Pachur [15], for example, has reported four recovery stages with the lowest temperature at ~250°C and the highest at 400°C In

26 EFFECTS OF RADIATION ON MATERIALS

A 24h 1

*-UNIRRADIATED AVERAGE

Transmission electron microscopy of the HFIR A212B material irradiated to ~1.3 x 1017

neutrons/cm2 was carried out using a Philips EM430T analytical microscope The precipitate morphology and dislocation structure were virtually unchanged by the HFIR irradiation Careful imaging of the matrix failed to reveal any defects that might account for the radiation hardening Similar results were also observed in a preliminary atom probe field ion micro- scope study and by small angle neutron scattering measurements carried out using the ORR facilities These studies imply that the defects responsible for hardening the HFIR pressure vessel are less than 2 nm in diameter and are present at a concentration level of less than

1017/cm3 Those observations are consistent with predictions from simple hardening models and the low hardening observed in this study

The results from these experimental studies support the hypothesis that the changes in fracture properties of the A212B HFIR material are the result of radiation hardening and that the hardening is unexpectedly rapid because of the unusual combination of low tem- perature and low flux involved

Investigations of flux effects on low temperature embrittlement have been reported in the literature Barton et al [76] evaluated hardening in mild steels irradiated to a fluence of

4 x 10" neutrons/cm2 at fluxes ranging from 10" to 1013 neutrons/(cm2 • s) Hinkle et al [17] evaluated hardening in nominally pure iron over a similar flux range at 90°C Neither study revealed any effect of flux More recently, Priest et al [18] compared the fracture properties of mild steels irradiated at 190°C in a test reactor with the properties of comparable steels irradiated at 180°C in a surveillance program The fluxes were 2.5 x 10" neutrons/ (cm2 • s) and 4 x 10s neutrons/(cm2 s), respectively, similar to those involved in our study Irradiation to a fluence of 1.5 x 1017 neutrons/cm2 in the high flux environment was found

to produce virtually no changes in fracture behavior However, similar exposure in the low flux environment produced a substantial upward shift in the fracture toughness versus tern-

NANSTAD ET AL ON SURVEILLANCE AND IRRADIATIONS 27

perature curve These observations are in qualitative agreement with our own, although the

~140°C difference in irradiation temperature and the compositional differences make it questionable whether the same types of defect structures are involved in the two studies Priest et al suggest that increasing the time to accumulate a given fluence level allows an increased number of copper clusters to form

Druce [79] has analyzed the HFIR data using the model developed by Fisher et al [20] for the United Kingdom Magnox pressure vessel steels The model is based on hardening arising from both radiation damage clusters and radiation-enhanced precipitation of copper Only the copper precipitation contribution is dependent on neutron flux according to the model while the total hardening is dependent on chemical composition, irradiation tem- perature, neutron flux, and total fluence The analysis concluded that the HFIR surveillance data, in which embrittlement increased with increasing copper content, are consistent with the copper-enhanced irradiation hardening model of Fisher, but that the predictions are extremely sensitive to the activation energy assumed for copper diffusion Although the experimental data imply a rate effect, the model does not predict a neutron flux effect over the range examined in the HFIR study, 107 to 1013 neutrons/(cm2 s), and furthermore, the hardening observed in the surveillance specimens is similar to mild steel irradiations at the same temperature in a United Kingdom MTR Druce notes the importance of neutron spectrum to the model predictions and points out that any differences between the HFIR and ORR spectra were ignored in his analysis

Clusters of solutes such as copper could conceivably be a factor in hardening the HFIR pressure vessel materials because of the extremely long irradiation times involved (—20 years) However, the principal hardening defects are more likely to be small interstitial clusters in the form of loops or small vacancy clusters in the form of depleted zones or microvoids, perhaps in association with impurity atoms The formation of such clusters is expected to be enhanced at very low fluxes The fates of point defects escaping collision cascades are widely appreciated as including mutual recombination, absorption at micro- structural sinks, and clustering with self-defects and impurities The fraction of defects annihilated by mutual recombination is proportional to the product of self-interstitial and vacancy concentrations When the flux, and hence the displacement rate, is lowered, the concentrations of both types of defects decrease, allowing more point defects per unit fluence

to avoid annihilation and to diffuse to existing sinks and cluster

As mentioned previously, the neutron spectrum below 0.1 MeV may be important to the embrittlement observed Nanstad et al [9] have postulated a model that incorporates point defect production and recombination effects associated with a low flux and the suspected efficiency of low energy neutrons in creating free point defects We postulate that cascades formed by lower energy neutrons have a more open structure and a larger fraction may avoid in-cascade recombination, the effect being more favorable at low fluxes Investigations are planned to perform detailed transport calculations of the spectra below 0.1 MeV for both the ORR and the HFIR surveillance sites, the results of which will address the spectrum uncertainties

Summary and Conclusions

1 Surveillance testing of ferritic steels from the HFIR pressure vessel revealed that embrittlement occurred at a significantly faster rate than predicted from test reactor irra- diations

2 Irradiations of HFIR archive A212B steel in the ORR required about ten times greater fluence to achieve similar embrittlement to the surveillance case and substantiated the postulated dose rate effect

28 EFFECTS OF RADIATION ON MATERIALS

3 Tensile and annealing experiments confirmed that classical radiation hardening oc- curred

4 Preliminary microstructural characterization efforts failed to reveal defects that might account for the radiation hardening, implying that defects are less than 2 nm in diameter and present at a concentration level below 10 I7 /cm\

5 Uncertainties regarding the role of the neutron spectrum below 0.1 MeV must be addressed before making a definitive conclusion regarding the rate effect

6 The results reported here have direct implications for predicted irradiation effects on structures such as light water reactor vessel supports that operate at low temperatures and low fluxes

References

[7] McWherter, J R., Schappel, R E., and McGuffey, J R., "HFIR Pressure Vessel and Structural Components Material Surveillance Program, ORNL-TM-1372, Oak Ridge National Lab., Oak Ridge, TN, January 1966

[2] "Evaluation of HFIR Pressure-Vessel Integrity Considering Radiation Embrittlement," R D Cheverton, J G Merkle, and R K Nanstad, Eds., ORNL/TM-10444, Oak Ridge National Lab., Oak Ridge, TN, April 1988

[3] Little, E A., "Neutron-Irradiation Hardening in Irons and Ferritic Steels," International Metals Reviews, The Metals Society and the American Society for Metals, March 1976, pp 25-60 [4] Weschler M S., "The Influence of Impurity-Defect Interactions on Radiation Hardening and Embrittlement," Proceedings of the ASMEICSME Pressure Vessel and Piping Conference, Mon- treal, Canada June 1978

[5] Nikolaev, V A and Badanin, V I., Atomnaya Energiya, Vol 37-6, No 491, 1974

[6] Brumovsky, M in Effects of Irradiation on the Substructure and Mechanical Properties of Metals and Alloys, ASTM STP 529, American Society for Testing and Materials, Philadelphia, 1973, p

46

[7] Tentative Structural Design Basis for Reactor Pressure Vessels and Directly Associated Components (Pressurized, Water Cooled Systems), PB 151987, Office of Technical Services, U.S Department

of Commerce, Washington, DC, December 1, 1958

[8] Steele, L E and Hawthorne, J R., "Effect of Irradiation Temperature on Neutron-Induced Changes in Notch Ductility of Pressure-Vessel Steels, NRL Report 5629, U.S Naval Research Laboratory, Washington, DC, June 28, 1961

[9] Nanstad, R K., Farrell, K., Braski, D N., and Corwin, W R., "Accelerated Neutron Embrit- tlement of Ferritic Steels at LowFluence: Flux and Spectrum Effects," Journal of Nuclear Materials, Vol 158, 1988, pp 1-6

[70] Steele, L E., "Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels," Technical Report Series No 163, International Atomic Energy Agency, Vienna, Austria, 1975

[77] Wechsler, M S., Berggren, R G., Hinkle, N E., and Stelzman, W J., "Radiation Hardening and Embrittlement in a Low-Carbon Pressure Vessel Steel," Irradiation Effects in Structural Alloys for Thermal and Fast Reactors, ASTM STP 457, American Society for Testing and Materials, Philadelphia, 1969, pp 242-260

[12] Cheverton, R D., Pennell, W E., Robinson, G C, and Nanstad, R K., "Impact of Radiation Embrittlement on Integrity of Pressure Vessel Supports for Two PWR Plants," NUREG/CR-

5320, ORNL/TM-10966, Oak Ridge National Laboratory, Oak Ridge, TN

[75] Nichols, F A., Philosophical Magazine, Vol 14, No 355, 1966

[14] Odette, G R., Lombrozo, P M., and Wullaert, R A., "Relationship Between Irradiation Hard- ening and Embrittlement of Pressure Vessel Steels," Effects of Radiation on Materials, ASTM STP 870, American Society for Testing and Materials, Philadelphia, 1985, pp 840-860

[75] Pachur, D., Nuclear Technology, Vol 59, No 463, 1982

[76] Barton, P J., Harries, D R., and Mogford, I L., "Effects of Neutron Dose Rate and Irradiation Temperature on Radiation Hardening in Mild Steels," Journal Iron Steel Institute, Vol 203, May