NUREG/CR-6897

 ANL-04/28

Assessment of Void Swelling in

Austenitic Stainless Steel 

Core Internals

Argonne National Laboratory

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Washington, DC 20555-0001
 

 NUREG/CR-6897

 ANL-04/28

Assessment of Void Swelling in

Austenitic Stainless Steel

Core Internals

Manuscript Completed: December 2004

Date Published: January 2006

Prepared by

H. M. Chung

Energy Technology Division

Argonne National Laboratory

9700 South Cass Avenue

Argonne, IL 60439

W. H. Cullen, Jr., NRC Project Manager

Prepared for

Division of Engineering Technology 

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

NRC Job Code Y6388
ii
Abstract 

As many pressurized water reactors (PWRs) age and life extension of the aged plants is

considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the 

subject of increasing attention. In this report, the available database on void swelling and density 

change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and

microstructural characteristics were carefully examined, and key factors that are important to 

determine the relevance of the database to PWR conditions were evaluated. Most swelling data

were obtained from steels irradiated in fast breeder reactors at temperatures >385°C and at dose 

rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation 

temperature and given steel, the integral effects of dose and dose rate on void swelling should not

be separated. It is incorrect to extrapolate swelling data on the basis of “progressive compounded 

multiplication” of separate effects of factors such as dose, dose rate, temperature, steel composition, 

and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine 

credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. 

Although the void swelling data extracted from fast reactor studies is extensive and 

conclusive, only limited amounts of swelling data and information have been obtained on 

microstructural characteristics from discharged PWR internals or steels irradiated at temperatures

and at dose rates comparable to those of a PWR. Based on this relatively small amount of

information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. 

As additional data and relevant research becomes available, the newer results should be integrated 

with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR 

baffle reentrant corners are the most likely location to experience high swelling rates, and hence,

high swelling at EOL, especially in internal regions of small volume where irradiation temperature 

is high. However, it is considered unlikely that void swelling in a reentrant corner will exceed the 

threshold level of ≈4% beyond which the swelling rate reaches the steady state rate of 1%/dpa. 

However, this estimation is only preliminary, and a more accurate quantification of maximum 

temperature of reentrant corners at EOL and life-extension situations would be useful. 

iii 

Foreword

v

This contractor-prepared NUREG-series report on void swelling is the first in a series of reports

addressing irradiation effects on the internal structures of pressurized-water reactor vessels. Forthcoming

NUREG-series reports will address irradiation embrittlement in cast and wrought stainless steels (SSs),

as well as the development and application of models for predicting the effects of irradiation.

Void swelling results from the agglomeration of microscopically small voids (or vacancies) produced

during neutron irradiation. The primary variables that control the degree of void swelling are temperature

and neutron dose.

Research on irradiation degradation in liquid metal, fast breeder reactor (LMFBR) systems in the early 1980s

produced an enormous quantity of data describing the kinetics and consequences of void swelling in SSs

and nickel-base alloys proposed for use in those systems. The operational parameters of breeder reactor

systems feature operating temperatures and neutron flux rates that are considerably higher than those that

are characteristic of light-water reactors (LWRs). Extrapolation of the LMFBR data to LWR operational

regimes is extremely tenuous, given the differences in energy spectra and temperature, leading to

predictions of LWR materials degradation that range from “possibly significant” to “little, if any, effect.”

Primarily through the Generic Aging Lessons Learned (GALL) program, the U.S. Nuclear Regulatory

Commission established guidance that the issue of void swelling should be made a part of the vessel

internals aging management program (AMP) for each licensee. As a result, the safety evaluation reports

(SERs) issued by the agency in response to license renewal applications generally state that each licensee

will make void swelling, and the other irradiation degradation issues, part of their AMP. The SERs also

state that licensees should keep the agency informed “on a periodic basis” of progress in the evaluation

of irradiation degradation issues.

This report contains critical reviews of several publications containing data that can be directly applied,

or extrapolated with some confidence, to the operational characteristics of LWRs. As such, this review

includes several relatively recent reports emanating from industry-funded programs. The conclusion is

that void swelling is not likely to be an issue for either the initial term of a license or its extension period.

Much additional research is currently underway, and the results should be carefully monitored for any

indication to the contrary.

 

Carl J. Paperiello, Director

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission
vi
vii

Contents

Abstract........................................................................................................................................................ iii

Foreword ..................................................................................................................................................... v

Executive Summary.................................................................................................................................... xi

Acknowledgements..................................................................................................................................... xiii

1 Introduction .......................................................................................................................................... 1

2 Void Swelling Data from PWR Components..................................................................................... 3

2.1 Void Swelling in Ringhals PWR Flux Thimble Tube.......................................................... 3

2.2 Void Swelling in PWR Lock Bars and Baffle Bolts ............................................................ 4

2.3 Void Swelling in a Tihange PWR Baffle Bolt...................................................................... 5

2.4 Void Swelling in Japanese PWR Flux Thimble Tubes........................................................ 6

2.5 Void Swelling in a Westinghouse PWR Flux Thimble Tube .............................................. 6

2.6 Summary of Void Swelling Data from PWR Internals........................................................ 7

3 Important Factors Used to Extrapolate Swelling Data from Fast-Breeder to PWR Conditions..... 9

3.1 Component Most Susceptible to High Swelling................................................................... 9

3.2 Maximum Irradiation Temperature ....................................................................................... 10

3.3 Swelling in EBR-II Components Irradiated at ≤380°C at Dose Rates Comparable to

PWR Dose Rates.................................................................................................................... 11

3.4 Swelling in Steels Irradiated at 305-335°C in BN-350 at Dose Rates Comparable to

PWR Dose Rates.................................................................................................................... 13

3.5 Effect of Irradiation-Induced Precipitation ........................................................................... 14

3.5.1 Low Swelling of CW Type 316 SS and High-Density Precipitation of γ’ and

an Unidentified Phase in PWR Baffle Bolts........................................................... 16

3.5.2 Low Swelling in 348 SS and Dense Precipitation of NbC in a PWR ................... 16

3.5.3 Low Swelling and Dense Precipitates in SA Type 304 Irradiated to 50 dpa

at 370°C in EBR-II at Dose Rate Relevant to PWR Reentrant Corner................. 17

3.5.4 Low Density Change and High-Density Precipitation in CW Type 316 SS

Irradiated to 23-51 dpa at 375-430°C in EBR-II at Dose Rate Relevant to

PWR Reentrant Corner............................................................................................. 17

3.5.5 Low Swelling and High-Density Precipitation of G Phase and TiC in

X18H10T SS Irradiated in BN-350 at 305-355°C at Dose Rates Relevant to

PWRs......................................................................................................................... 17
viii

3.6 Swelling Behavior of Type 304L SS..................................................................................... 17

3.7 Steady-State Swelling Rate of 1%/dpa.................................................................................. 17

3.8 Threshold Swelling to Enter the Regime of Steady-State Swelling Rate ........................... 18

4 Assessment of the Potential for Void Swelling for PWR Internals at EOL..................................... 19

4.1 Thin-Walled Flux Thimble and Instrument Tubes............................................................... 19

4.2 Baffle Bolts............................................................................................................................. 19

4.3 Baffle Plate Reentrant Corners .............................................................................................. 19

5 Summary and Conclusions.................................................................................................................. 21

References................................................................................................................................................... 23
ix

Figures

1. Range of irradiation temperature and dose for which void swelling data have been reported

for PWR core internals. ....................................................................................................................... 7

2. Void swelling of PWR internals plotted as a function of dose. ........................................................ 8

3. Example of estimation of EOL void swelling reported by Garner for solution-annealed

Type 304 SS reentrant corners. ........................................................................................................... 10

4. Density change in CW Type 316 SS fuel hex can irradiated in reflector region of EBR-II at

376-460°C to 5-80 dpa. ....................................................................................................................... 12

5. Microstructure of SA Type 304 SS fuel subassembly hex can irradiated in EBR-II at 370°C

to 50 dpa. .............................................................................................................................................. 12

6. Void swelling in Ti-containing Russian steel after 5.3 y in BN-350 breeder at 311-313°C........... 13

7. Schematic illustration of characteristics of irradiation-induced precipitation sensitive to dose

rate and irradiation temperature. ......................................................................................................... 14

8. Schematic illustration of three types of relationship between irradiation-induced

precipitation and void swelling as a function of irradiation dose. .................................................... 15

9. Low void swelling and high-density precipitation of γ’ phase and unknown phase in the

Tihange PWR baffle bolt fabricated from CW Type 316 SS............................................................ 16
x

Tables

1. Void swelling in Ringhals-II PWR flux thimble tube after 23 years in service. ............................ 4

2. Void swelling data from industry PWR Baffle Bolt Project, quoted in MRP-50............................ 5

3. Void swelling data obtained from Japanese PWR flux thimble tubes. ............................................ 6

4. Void swelling data obtained from Westinghouse PWR flux thimble tubes..................................... 7
xi

Executive Summary

As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered,

void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of

increasing attention. In this report, the available database on void swelling and density change of

austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural

characteristics were carefully examined, and key factors that are important to determine the relevance of

the database to PWR conditions were evaluated.

Most swelling data that are available were obtained from steels irradiated in fast breeder reactors at

temperatures >385°C and at dose rates that are orders of magnitude higher than PWR dose rates. Extreme

care must be exercised when interpreting and extrapolating such data. These data cannot be extrapolated

to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions.

Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void

swelling cannot be separated. It is incorrect to extrapolate swelling data on the basis of “progressive

compounded multiplication” of separate effects of factors such as dose, dose rate, temperature, and

material composition.

Limited swelling data are available for cold-worked (CW) Type 316 SS irradiated to 53 dpa at 376-

386°C and for solution-annealed (SA) Type 304 SS irradiated to 50 dpa at ≈370°C in EBR-II reflector

positions at dose rates comparable to those of PWR reentrant corners. As such, these data are relevant to

the conditions of PWR reentrant corners. Swelling in these materials was less than 1%.

The low void swelling observed in PWR components and in EBR-II steels under PWR-relevant

dose rates appears to be associated with irradiation-induced formation of very fine precipitates (such as G

phase, carbides, and γ’ phase) in high number density. Such irradiation-induced precipitation at low

temperatures (<370°C) creates an extremely large internal surface, i.e., the interface between the steel

matrix and the precipitates. This interface acts as an efficient sink to irradiation-induced vacancies,

thereby suppressing the agglomeration of the vacancies. Irradiation-induced precipitation is sensitive to

minor alloying and impurity elements, irradiation temperature, and dose rate.

In thin-walled flux thimbles and instrument tubes, the effect of gamma heating is insignificant. The

currently available database is sufficient to conclude that void swelling in this type of reactor internal,

mostly fabricated from CW Type 316 SS, is not an issue.

Most PWR baffle bolts are fabricated from CW Type 316 SS. The data obtained from the industry

baffle bolt program show that swelling is insignificant (<0.25%) for dose levels up to ≈20 dpa and

irradiation temperatures of up to ≈340°C. Data obtained on EBR-II components irradiated at

temperatures <380°C and at comparable dose rates are consistent with the data from the industry bolt

program. Microstructural characteristics of the two groups of materials are also consistent. Thus, it is not

likely that void swelling in this type of reactor internal will exceed the threshold level (i.e., ≈4%) that is

necessary to enter the regime of the steady-state swelling rate of 1%/dpa.

Most baffle reentrant corners are fabricated from SA Type 304 SS and are most susceptible to high

swelling rates, and hence, high swelling at EOL. The maximum irradiation temperature in some regions

of the reentrant corners has been estimated to be in the range of ≈380-420°C. Only one investigation has

reported void swelling for this class of steel after irradiation in EBR-II at 370°C to ≈50 dpa at a dose rate 

comparable to that of reentrant corners. Void swelling in this material was only 0.54%. The low swelling
 

 xii

 

appears to be related to high-density irradiation-induced precipitation of very fine carbides. Therefore, it 

is considered unlikely that void swelling in reentrant corners will exceed the threshold level of ≈4%. As a 

consequence, the potential impact of void swelling on core flow and the structural functions of the PWR 

internals is believed to be negligible. 

However, this estimation is only preliminary. More relevant swelling data, including the effects of 

dose rate and temperature would be extremely helpful. Analyses to compute more accurate quantification 

of maximum temperature at EOL and life-extension situations are needed for SA Type 304 and Type 316 

SSs. A better mechanistic understanding of the roles of irradiation-induced microstructural evolution on 

void swelling is also needed. Several experimental programs to address these issues are underway, and 

are expected to produce useful data in the next few years. 
xiii

Acknowledgements

The author thanks W. J. Shack, W. H. Cullen, Jr., and A. D. Lee for helpful comments.
xiv
 

 1

 

1 Introduction 

As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, 

void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of 

increasing attention. Excessive void swelling can lead to dimensional instability of the component and 

significant degradation of fracture toughness. It could also influence or contribute to the susceptibility of 

the component to irradiation-assisted stress corrosion cracking (IASCC), stress relaxation, or irradiation 

embrittlement. 

Because of experimental problems largely related to the salvaging and handling of large, highly 

irradiated reactor components, it is either very difficult or impractical to obtain directly applicable void 

swelling data from actual PWR internals that have reached an end-of-life (EOL) condition. Therefore, a 

method commonly used to infer such information is to extrapolate more abundant data obtained under 

liquid metal, fast breeder reactor (LMFBR) conditions to PWR EOL or life-extension conditions. Many 

uncertainties are involved in this process, and the outcome of such extrapolation is often considered either 

unconvincing or questionable. 

Compared to LMFBRs, the neutrons resulting from the fission process in light water reactors 

(LWRs) are moderated and slowed, or “thermalized”, resulting in a higher probability of their capture by 

surrounding materials. As these thermal neutrons bombard the structural materials that support the fuel 

and reactor core components, the atoms of the material, are displaced, often much more than just once, 

resulting in the production of large numbers of vacancies and interstitials, with important consequences 

on strength and resistance to failure of these materials. Simultaneous with the bombardment and 

consequent damage, the moderately high temperatures of the core internal materials [~338°C (~640°F)] 

produce an offsetting process of annealing and consolidation of the radiation damage into voids, 

dislocation loops and other microscopically faulted structures.*

 In addition, many of the point defects 

formed through neutron bombardment simply diffuse to sinks of various types, including external 

surfaces, grain boundaries, or interphase surfaces. If the temperatures are high (around 400° to 560°C) 

and the irradiation damage extensive (>1023 n/cm2, E>0.1 MeV), the voids that are formed could 

conceivably produce dimensional changes, resulting in misfitting of the components, increases in stress, 

and other unacceptable consequences. 

The data from investigations related to LMFBR performance shows the void swelling process is 

divided into two phases – an incubation period, followed by a period of void formation. Once voids have 

started to form, a general rule of thumb is that swelling proceeds at a rate of 1% per dpa accumulated. 

However, the incubation period is influenced by a number of important variables, including cold work, 

which tends to extend the period of incubation, and metal chemistry, which can either extend or 

foreshorten the incubation period. Because cold work extends the onset of swelling, core baffle bolts are 

frequently fabricated from 20% cold-worked (CW), Type 316 SS. There are many other considerations to 

the topic of void swelling, and the reader is referred to Ref. 1 – an excellent review of irradiation damage, 

albeit from a LMFBR point of view. 

 

*

 Neutrons can also cause transmutation of some of the lighter elements such as boron, generally found in trace amounts in structural materials. 

However, the amounts of helium and possibly hydrogen that are produced by transmutation of light elements is not likely to contribute 

significantly to void formation. Thermal neutron bombardment of nickel can produce a few parts per million amounts of helium during the 

course of a reactor lifetime, which could contribute to void production. 
2

The relatively high operating temperatures and much higher (but less thermalized) neutron fluxes

that characterize fast reactors enable relatively easy detection of void formation in many core component

materials, especially solution–annealed (SA) and CW 300–series SSs. At issue has been whether voids

would form and become problematic under the less favorable LWR operational characteristics. The

objective of this report is to critically review the applicable database and independently assess void

swelling in austenitic SS PWR core internals.
3

2 Void Swelling Data from PWR Components

In this section, void swelling data obtained from discharged PWR core internals are reviewed. At

this time, such data are rather limited except for results obtained on flux thimble tubes.

2.1 Void Swelling in Ringhals PWR Flux Thimble Tube (Ref. 2)

The Electric Power Research Institute (EPRI) project (Project Manager, H. T. Tang) for this

investigation2 was cosponsored by the industry International Irradiation-Assisted Stress Corrosion

Cracking Advisory Committee (Chairman, T. Mager, Westinghouse), and the report was authored by

Westinghouse (J. Conermann and R. Shogan) and Japanese PWR (JPWR) investigators (S. Yamaguchi,

H. Kanasaki, S. Nashida, K. Fujimoto, Y. Tamaguchi, and T. Yonezawa).

Several types of tests under PWR conditions were performed on small specimens obtained from a

flux thimble tube after 23 y of service in the Ringhals PWR (in Sweden), i.e., microstructural

characterization by transmission electron microscopy (TEM), stress corrosion crack initiation in O-ring

specimens, slow-strain-rate tensile tests in PWR water, and tensile tests in an inert environment.

The flux thimble tube was fabricated from Type 316 SS (composition in wt.%: C 0.045, Si 0.43,

Mn 2.69, P 0.026, S 0.01, Ni 13.3, Cr 17.4, and Mo 2.69). The Mn content of the heat is somewhat

higher than the normal AISI specification limit of 2.0 wt.%. The thimble was 10-12% CW after final

anneal, producing a yield strength of 480-620 MPa and an ultimate tensile strength of >690 MPa.

The position of the thimble tube was near the center of the core. Depending on the axial level, the

estimated irradiation temperature and neutron damage of the specimen blanks ranged, respectively, from

290 to 325°C and from 17 to 65 dpa. Because all specimen blanks were in service for the same period of

time of 23 y, the dose rate of each blank varied, that is, a lower dose rate for a 17-dpa blank and 3.8 times

higher dose rate for a 65-dpa blank.

Microstructural characterization was performed on three blanks at the Nuclear Development

Company (NDC) in Japan. Thin foils were characterized with a field-emission-gun transmission electron

microscope (FEG-TEM), Model JEOL JEM-2010F. Microcavities and voids were imaged in bright field

at magnifications of 500,000 and 1,000,000 times by the through-focus technique (over-focus, in-focus,

and under-focus). The size of most voids was in the range of 1-2 nm, but spherically shaped

microcavities as large as ≈6 nm in diameter were also observed.

It is customary to independently measure density change of irradiated specimens in parallel with

void characterization. However, the swelling data were not confirmed by this technique.

Determination of the degree of swelling was based on the average size and number density of the 

voids that were imaged in the bright field. In this process, the thickness of the imaged thin foil region

was accounted for. The results, obtained for three samples, are summarized in Table 1.

The void number was counted by a computerized method. Bright-field high-magnification images

of an examined region that contained thousands of voids were scanned into a computer, and the number

of voids was determined with an image-processing software. This method is fairly accurate.
4

Table 1. Void swelling in Ringhals-II PWR flux thimble tube (CW Type 316 SS) after 23 y in service

(from MRP-73, Ref. 2).

Sample

ID

Irradiation

Temp.

(°C)

Neutron

Damage

(dpa)

Average

Void Size 

(nm)

Total

Void

Number

Imaged

Volume

(106 nm3)

Swelling

(%)

A 320 65 1.20 2969 9.14 0.0294

B 295 61 1.24 2130 8.14 0.0270

E 325 17 1.34 829 7.57 0.0140

Uncertainty in the thickness of the examined region of the thin foil could lead to some degree of

uncertainty in the reported degree of void swelling. However, typical uncertainty in foil thickness is no

more than a factor of two, which means that void swelling should be no more than ≈0.06 %.

The electron-transparent region of the thin film was prepared by electrochemical jet polishing with

a methanol-water solution that contained 5 vol.% perchloric acid. In this type of method, artifact surface

hydrides and oxides of microscopic size could be produced, depending on material type and jet-polishing

temperature. If such artifacts were produced, the net effect would be somewhat less void swelling than

the measured value. However, such an effect is not significant for austenitic SSs.

Besides twins and the customary black-dot defects, the investigators did not characterize

irradiation-induced precipitates, which would require difficult dark-field and tilting analysis. Therefore, it

is not clear if voids and cavities were the only significant microstructural features.

However, the study provides a valuable database which shows that void swelling at 290-325°C at

≤65 dpa was insignificant in the CW Type 316 SS flux thimble tube after 23 y of service in the Ringhals

PWR.

2.2 Void Swelling in PWR Lock Bars and Baffle Bolts (Ref. 3)

The EPRI interim report (Project Manager, H. T. Tang) was authored by U.S. industry investigators

(S. Byrne and I. Wilson, Westinghouse and S. Fyfitch and H. Xu, Framatome) and F. A. Garner of Pacific

Northwest National Laboratory (PNNL).

3 The sections on void swelling in the report seem to heavily

reflect the view of the latter investigator who has extensive experience in void swelling in steels irradiated

in a fast breeder reactor or under fusion reactor conditions. The report provides a comprehensive

assessment of void swelling data obtained under such accelerating conditions, and of the effects of 

separate factors such as irradiation temperature, dose rate, dose, neutron spectrum, impurities, cold work,

annealing, and irradiation creep.

However, the report provides only very limited information on void swelling in PWR internals.

The data, summarized in Table 2, are based on two earlier reports.

4,5 Void swelling data in the latter

document appear to be from the results of a transmission electron microscopy (TEM) investigation

conducted by Thomas and Bruemmer and documented in Ref. 6. The results of this investigation,

performed on a Tihange PWR baffle bolt, will be reviewed separately below.

The data summarized in Table 1 (flux thimble) and Table 2 (baffle bolt) are essentially consistent,

in that void swelling in the examined PWR components was insignificant. Void swelling of 0.20-0.24%

was observed in the local regions of bolt 2K1R5, which had been exposed to somewhat higher
5

temperatures, i.e., 330-343°C vs. <329°C. For bolt 4522, the data indicate a slightly higher swelling

(0.029 vs. 0.0%) at higher temperature (329 vs. 290°C). However, a definite trend of temperature

dependence is not observed.

2.3 Void Swelling in a Tihange PWR Baffle Bolt (Refs. 3, 6, and 7)

The primary focus of the report in Ref. 3 (EPRI-1003422) is FEG-TEM characterization of grainboundary crack-tip microstructure and microchemistry of a field-cracked Oyster Creek boiling water

reactor (BWR) top guide and a Tihange PWR baffle bolt. Several high-quality photomicrographs shown

in the report provide information on void swelling in the baffle bolt material (Bolt ID 2K1R, Specimen ID

2K1R1, from the bolt shank, CW Type 316 SS, see Table 2). In Figs. 3-5 (a) and (b) of the report, two

types of spherical voids are shown, one 4-6 nm in diameter; the other, <1 nm. This observation is

consistent with the results of the NDC characterization of voids in the Ringhals flux thimble irradiated to

61-65 dpa (see Section 2.1 above).

As listed in Table 2, void swelling in the Tihange bolt 2K1R was ≈0.2% at 7.5–12.2 dpa.

Reference 7 by Garner et al. 2001, presents FEG-TEM photomicrographs that show the morphology and

distribution of voids at three locations of the same bolt: head (1 mm from the head surface, 320°C,

19.5 dpa, seven voids visible), shank (25 mm from the head surface, 343°C, 12.2 dpa, 0.20% swelling),

and near the bolt threads (55 mm from the head surface, 333°C, 7.5 dpa, 0.24% swelling).

Table 2. Void swelling data from industry PWR Baffle Bolt Project, quoted in MRP-50, (Ref. 3).

Bar or

Bolt ID

Specimen

Location

Steel

Type

Irradiation

Temp. (°C)

Neutron

Damage (dpa)

Swelling

(%)

G130 Lock bar SA 304 290 20.7 0.0

3324 Lock bar SA 304 290 25 0.0

2322 Lock bar SA 304 290 24 0.0

2K1R5 Bolt head CW 316 320 19.5 Visible voids

2K1R5 Bolt shank CW 316 343 12.2 0.20

2K1R5 Near bolt

threads

CW 316 330 7.5 0.24

G96B Near bolt

threads

CW 316 329 9.2 0.011

G96B Bolt head CW 316 290 18.4 0.0

4522 Near bolt

threads

SA 347 329 8.6 0.029

4522 Bolt head SA 347 290 16.4 0.0

4116 Bolt head SA 347 290 14.1 0.0

2322 Near bolt

threads

SA 347 329 11 0.004

2322 Bolt head SA 347 329 20 0.0
6

2.4 Void Swelling in Japanese PWR Flux Thimble Tubes (Ref. 8)

This joint project8 was conducted by investigators of the Japan Institute of Nuclear Safety

System, Inc., Nippon Nuclear Fuel Development Co., and Kyoto University. Microstructural

characterization was performed with a Hitachi HF-2000 FEG-TEM.

Two flux thimble tubes were irradiated in a JPWR during service for 9 and 13 effective full power

years (EFPY). The estimated neutron damage and irradiation temperature, respectively, are 35 dpa and

310°C for the former (9 EFPY) tube and 1-53 dpa at 290-320°C for the latter tube. Both tubes were

fabricated from a 15%-CW heat of Type 316 SS that contained (in wt.%) 0.040 C, 0.62 Si, 1.63 Mn,

0.22 P, 0.006 S, 12.6 Ni, 16.94 Cr, and 2.22 Mo.

The shape of the examined region of the thin foil and the distribution of voids were determined by

the technique of stereomicroscopy as a function of foil thickness. Stereomicroscopy of voids or

precipitates as small as ≈1 nm is very difficult and painstaking work. Size distribution of voids was also

determined. Most voids were <1.4 nm, and only a limited number of voids >2 nm was observed. The

overall quality of the FEG-TEM work appears to be excellent, and the information on the spatial and size 

distributions of the voids provides additional confidence in the quality of swelling data.

Table 3 summarizes the results of FEG-TEM characterization of void size, distribution, and

swelling percent. The results, consistent with the swelling behavior of the Ringhals flux thimble tube and

Tihange baffle bolt, showed that swelling was only <0.042% and that there was no clear dependence on

dose in the range of 1-53 dpa. The latter is probably because swelling was so low.

Table 3. Void swelling data obtained from Japanese PWR flux thimble tubes (CW Type 316 SS), from

Fujii et al. 2001 (Ref. 8).

Dose

(dpa)

Dose rate

(10-9 dpa/s)

Irradiation

Temperature

(°C)

Void Average

Diameter

(nm)

Void Number

Density

(1023 m-3)

Void

Swelling

(%)

1 2 290 (No void) 0.0 0.0

3 8 290 0.94 3.6 0.015

10 26 320 0.92 5.0 0.020

28 69 320 0.95 9.4 0.042

31 76 290 1.01 6.9 0.038

33 82 320 1.04 3.1 0.018

53 130 300 1.05 5.8 0.036

35 110 310 0.94 3.8 0.016

35 110 310 0.96 4.2 0.020

35 110 310 0.98 3.8 0.019

2.5 Void Swelling in a Westinghouse PWR Flux Thimble Tube (Ref. 9)

In this earlier work by Westinghouse investigators,

9 two flux thimble tubes were examined after

11 EFPY of service in a PWR. The tubes were fabricated from CW Type 316 SS that contained (in wt.%)

0.044 C, 0.75 Si, 1.75 Mn, 0.014 P, 0.011 S, 13.10 Ni, 17.50 Cr, 2.64 Mo, and 0.040 Co.
7

The irradiation at 305-315°C produced ≈35 dpa and 208 appm He (He/dpa ratio 5.9 appm/dpa).

The procedure for specimen preparation was similar to that of the other studies above (electrochemical jet

polishing at –35°C in a solution of 20% perchloric acid and 80% methanol). Microstructural

characterization was performed with a Phillips CM30 analytical TEM. Table 4 summarizes void swelling

data from this investigation. The result is similar to results from the Ringhals and the JPWR flux thimble

tubes.

Table 4. Void swelling data obtained from Westinghouse PWR flux thimble tubes (CW Type 316 SS),

from Foster et al. (Ref. 9).

Dose

(dpa)

Irradiation

Temp. (°C)

Void Average

Diameter (nm)

Void Number Density

(1023 m-3)

Swelling

(%)

3 310 (No void) 0 0

35 310 ≈1 6 0.03

2.6 Summary of Void Swelling Data from PWR Internals

The void swelling data listed in Tables 1–4 are plotted in Fig. 1 as a function of dpa and irradiation

temperature. Note that a higher level of swelling is denoted with a larger circle with a darker color.

Figure 2 is a plot of void swelling of PWR internals as function of dose (dpa).

The maximum observed void swelling was 0.020-0.24% at ≈333-343°C at 7.5-12.2 dpa for the flux

thimble tube and baffle bolts fabricated from CW Type 316 SS. No swelling larger than 0.25% was

observed. This level of swelling is insignificant.

280

290

300

310

320

330

340

350

360

0 10 20 30 40 50 60 70 80 90 100

0

0-0.01

0.01-0.02

0.02-0.03

0.03-0.04

0.042-0.1

0.20-0.25

Ir

r

a

dia

tio

n

T

e

m

p

e

r

a

t

u

r

e

(

°

C

)

Neutron Damage (dpa)

void swelling (%)

PWR Internals

SA304, CW316,

SA347 SSs

CW316

CW316

CW316

SA304

CW316

SA347

CW316

SA347

CW316

CW316

CW316

CW316

Approximate Line

for Occurrence of

Void Swelling

of <0.25%

0.0294%

Figure 1. Range of irradiation temperature and dose for which void swelling data

(in color code) have been reported for PWR core internals.
8

0

0.05

0.1

0.15

0.2

0.25

0.3

0.35

0.4

0 10 20 30 40 50 60 70 80

Ringhals Thimble, CW316

Lock Bar, SA304

Tihange Baffle Bolt, CW316

Baffle Bolt, SA347

JPWR thimble, CW316

WPWR thimble, CW316

V

oid

S

w

ellin

g

(

%

)

Neutron Damage (dpa)

PWR Internals

SA304, CW316,

SA347 SSs

Figure 2. Void swelling of PWR internals plotted as a function of dose (dpa).

To show, approximately, the range of dpa-temperature in which void swelling of <0.25% was

observed, a trend line was drawn in the figure. Because the void swelling was so low and because the

uncertainty limit associated with such low swelling is significantly large, it is not possible to extract a 

more quantitative correlation among swelling, dpa, and irradiation temperature.

One conclusion that can be made with reasonable confidence is that void swelling in PWR flux

thimble tubes and in baffle bolts is not a concern. For baffle bolts, the primary concern is the 

susceptibility to IASCC rather than void swelling. At this time, however, there is no conclusive evidence

that void swelling plays an important role in IASCC of PWR baffle bolts.
9

3 Important Factors Used to Extrapolate Swelling Data from FastBreeder to PWR Conditions

Void swelling refers to the volume expansion of materials under neutron irradiations. Microscopic

voids developed from vacancy coalescence give rise to this geometry instability, which is quantified by

percent change in volume of the material. Void swelling occurs as a result of microstructural changes

involving point defects migration to various sinks such as interface, grain boundaries, and dislocations.

The interaction between interstitials and dislocations is generally believed to be stronger due to a 

relatively large strain field that surrounds interstitials. Therefore, a preferential interstitial flux exists

towards dislocations. Because of this “bias” interstitial flow, the remaining vacancies cannot be

annihilated by recombining with interstitials and result in void nucleation and growth. Fission products

such as helium or hydrogen also play an important role in void swelling. By combining with these gas

atoms, void nucleation is facilitated through reduction of surface energy of vacancy cluster. However, the

fundamental driving force of void swelling remains to be the excess vacancy flux towards voids.

The most important parameters that can influence void swelling are temperature and accumulated

dose; other parameters that can influence void swelling include dose rate, material microstructure, and

stress. At a given temperature, an incubation period of void swelling is observed in the low dose regime.

Beyond a critical dose, a dramatic increase in void swelling takes place and a steady–state linear 

dependence of void swelling with radiation dose is observed. At constant radiation dose, the temperature

dependence of void swelling shows a peak at an intermediate temperature,

10 characteristic peak

temperature is typically around 0.3 Tm. For austenitic SSs, a CW structure is more resistant to void

swelling relative to a SA structure.

11 Cold work extends the incubation period for void swelling and

increases the critical dose for the steady–state linear void swelling.

The void swelling data from PWR components that were reviewed in Section 2 are limited in

irradiation temperature (≤343°C), dose (≤65 dpa), or both. For some components, these limits do not

pose a problem in evaluating void swelling at EOL, e.g., the Ringhals flux thimble tube which was in

service for 23 y at 295-325°C and exhibited <0.03% swelling at <65 dpa. This observation and similar

results of Fujii et al.,

8 are sufficient to conclude that void swelling in a flux thimble tube is not a concern.

However, some core internals are exposed to more severe conditions, i.e., significantly higher dose

at higher temperatures. These components (e.g., baffle plate reentrant corners) are thick-walled, and their

service temperature is somewhat elevated because of gamma heating. According to Garner,

* baffle

reentrant corners are exposed to the highest temperature and to the highest level of damage compared to

the other components (baffle plate, bolt, or former) (see Fig. 3). Assessment of void swelling for such a

component is possible only through extrapolation of data obtained under non-PWR conditions in which

dose rate and irradiation temperature differ significantly. Such extrapolation is a formidable task, and

requires great caution. This section describes our current understanding of the key factors and soundness

of such an extrapolation.

3.1 Component Most Susceptible to High Swelling

Garner attempted to evaluate the swelling behavior of reentrant corners, the component considered

most susceptible to high swelling, Fig. 3 shows a summary of his analysis. For 40-y operation, the

 

* Private communications Frank Garner to Bill Cullen, August 2005.
10

estimated swelling is as large as 55% (see Fig. 3d). Although this estimation is intended primarily to

illustrate an approach and is based on conservative assumptions, the large difference between Garner’s

estimation for reentrant corners and the maximum swelling actually observed for other PWR components,

i.e., 55% vs. ≤0.24%, is striking, and indicative of the difficulties encountered in the extrapolation of

LMFBR-related test parameters and methods to the LWR operating characteristics.

3.2 Maximum Irradiation Temperature

Irradiation temperature is probably the single most important factor that influences void swelling in

an austenitic SS at EOL. The irradiation temperature in the reentrant corners in Fig. 3 was estimated to be 

as high as 420°C, about 90-95°C higher than the PWR outlet coolant temperature. The increase is

(b)

(a)

(c) (d)

Figure 3. Example of estimation of EOL void swelling reported by Garner for solution-annealed

Type 304 SS reentrant corners (see footnote previous page for Ref.).
11

primarily due to gamma heating. The estimated value is probably too high. The maximum temperature

in reentrant corners probably does not exceed ≈380°C. For example, Allen et al. considered a maximum

temperature of ≈370°C for PWR internals at a 40-y EOL.

12

From the standpoint of void swelling that occurs over a long period of time (i.e., EOL or lifeextension situations), the difference in irradiation temperature (e.g., 420°C vs. 370°C) is a crucial factor.

Irradiation temperature of a reentrant corner is influenced by many factors such as: steel type; impurities

in the steel; the source, energy, and the quantity of gamma-emitting radio nuclides; decay rate of the

gamma source; thermal conductivity of the steel; oxide layer thickness on the surface of the reentrant

corner; coolant temperature; fuel linear power generation rate; fuel burnup; axial location of the reentrant

corner; and core axial power distribution. The linear power generation rate of the peripheral or corner

rods in the peripheral fuel assembly strongly influences the temperature of the adjacent reentrant corner.

These days, most PWRs operate with fuel burnup extended as high as ≈62 MWd/kg U. Because fuel liner 

power generation rate decreases significantly with increasing burnup, high-burnup operation is conducive

to a lower irradiation temperature in a reentrant corner (than medium-burnup operation). Satisfactory

quantification of the factors listed above, however, remains elusive. For a better understanding of void

swelling behavior of reentrant corners under EOL or life-extension condition, accuracy in irradiation

temperature is more important than any other factors, e.g., dose and dose rate.

3.3 Swelling in EBR-II Components Irradiated at ≤380°C at Dose Rates

Comparable to PWR Dose Rates

Most of the data from steels irradiated in EBR-II were obtained for temperatures >385°C.

However, some data have been reported by Allen et al.

12 for CW Type 316 SS at 376-386°C and by

Chung et al.

13 for SA Type 304 SS at 370°C. The data are shown in Fig. 4. The steel specimens were

obtained from fuel subassemblies (1-mm-thick hex can) located in the reflector region of the reactor. In

that location, the steels are irradiated at lower temperatures, and their dose rates are about one order of

magnitude lower than that in the fueled region. As such, the data are more relevant to the behavior of a

PWR reentrant corner than the data obtained from steels irradiated at higher temperatures in the core 

center.

The CW Type 316 SS hex cans were fabricated from two heats, and the data in Fig. 4 reflect the

effect of three dose rates that were a factor of ≈3 apart. Figure 4 reveals: (a) a low but appreciable level

of densification at low dpa, indicating the effect of irradiation–induced precipitation, (b) an insignificant

effect of dose rate, (c) swelling at 376-386°C no greater than 0.8%, and (d) the 1%/dpa steady-state

swelling rate not reached even at 80 dpa at 460°C.

The distribution of voids and the microstructure of the hex can reported in Ref. 13 (≈50 dpa at

≈370°C) are summarized in Fig. 5. Like a PWR reentrant corner, the hex can was fabricated from a SA

Type 304 SS. At ≈50 dpa, the void swelling in the SA Type 304 SS hex can (i.e., 0.54%) agrees well

with the density change in the CW Type 316 SS hex can irradiated at 376-386°C (i.e., ≈0.8%, Fig. 4).

Figure 3-3 of Ref. 3 shows a TEM photomicrograph of SA Type 304 SS irradiated in EBR-II to

21.7 dpa at 380°C. The figure, attributed to “unpublished micrograph of F. A. Garner,” shows a void

swelling of ≈2%. This much swelling is significantly higher than that for <386°C in Fig. 4. It is not

known if the steel was irradiated in the fueled region where the dose rate is orders of magnitude higher.

The accuracy of the irradiation temperature is not known either. Also, the morphology (polygonal shape)

and size (≈20-60 nm) of the voids in the micrograph differs somewhat from the morphology (spherical)
12

and the size (20-50 nm) of the voids in Fig. 5. The primary difference is the much higher void number

density in the former material than in the latter.

-1

0

1

2

3

4

5

0 10 20 30 40 50 60 70 80 90 100

CW316, Subassembly A

CW316, Subassembly B

CW316, Subassembly C

Density Change (%)

Dose (dpa)

SA 304 SS

!50 dpa at 370°C

void swelling 0.54%

Chung et al. 2001

EBR-II fuel subassemblies

(hex cans), 376-460°C

3 dose rates, factor of 3 difference

2 heats of 12% CW 316 SS

maximum

density change

for 376-386°C

1 %/dpa

steady-state

swelling line

Figure 4. Density change in CW Type 316 SS fuel hex can irradiated in reflector region of EBR-II at

376-460°C to 5-80 dpa (Ref. 8). Void swelling in a similar component fabricated from SA

Type 304SS is shown for comparison (from TEM analysis of Ref. 13; see Fig. 12).

(a) (b)

Figure 5. Microstructure of SA Type 304 SS fuel subassembly hex can irradiated in EBR-II at 370°C to

50 dpa (from Ref. 13); (a) voids near grain boundary, (b) high magnification, (c) voids and

twins in low magnification, and (d) dark-field image of dense carbide precipitates.
13

(c) (d)

Figure 5. (Contd.)

3.4 Swelling in Steels Irradiated at 305-335°C in BN-350 at Dose Rates

Comparable to PWR Dose Rates

BN-350 is a sodium-cooled fast breeder reactor (FBR) located in Kazakhstan. Void swelling

behavior of a Ti-containing steel similar to AISI 321 SS was investigated by Porollo et al. after irradiation

at 305-355°C at dose rates between 11.2 x 10-9 and 156 x 10-9 dpa/s.

14 The lower range of dose rate is

comparable to those of a flux thimble tube, core barrel, and baffle former in a western PWR (see Table 3).

The higher dose rate is comparable to the dose rate of a PWR reentrant corner. The experiment was

performed with a flow restricter of the reactor, which was fabricated from a Russian Type X18H10T steel

(composition in wt.%: 18.5Cr, 9.5Ni, 1.5Mn, 0.7Si, 0.6Ti, and ≤0.12C). The component was irradiated

continuously for 5.3 y without interruption.

Void swelling under all examined conditions was <0.98%. Swelling at ≤307°C was too low to

measure. Figure 6 shows void swelling for irradiation temperatures in the range of 311-313°C. Note that

in the constant-time irradiation, dose, dose rate, and specimen temperature vary, depending on the

0

0.1

0.2

0.3

0.4

0.5

0 10 20 30 40 50 60

V

oid

S

w

ellin

g (%)

Dose (dpa)

X18H10T SS

18.5Cr-9.5Ni-1.5Mn-0.7Si-0.6Ti-!0.12C

BN-350 FBR Flow Restrictor,

5.3 yr Continuous Irradiation

in an Out of Core Position

Prollo et al. 2001

10th Envi. Degr. Mater.

Irrad. Temp.

311-313°C

Figure 6.

Void swelling in Ti-containing

Russian steel after 5.3 y in BN-350

breeder at 311-313°C (3-53 dpa)

(from Ref. 14).
14

location of the specimen. Although the swelling was low, i.e., <0.5%, it exhibits the trend that swelling at

a lower dose rate (hence lower dose) is higher than at higher dose rate (hence higher dose).

This observation provides strong evidence that even for a given irradiation temperature for a 

particular steel, the integral effects of dose and dose rate cannot be separated. Therefore, it would be

incorrect to extrapolate swelling data on the basis of “progressive compounded multiplication” of the 

separate effects of dose, dose rate, and temperature.

3.5 Effect of Irradiation-Induced Precipitation

Irradiation-induced precipitation leads to profound modification of microstructure. Similar to

irradiation-induced segregation which is a precursor process, irradiation-induced precipitation is sensitive

to steel type, minor alloying and impurities, irradiation temperature, dose, and dose rate.

For a given material irradiated to the same damage level, the degree of irradiation-induced

segregation and irradiation-induced precipitation is sensitive to temperature and dose rate.

15,16 This

behavior is schematically illustrated in Fig. 7. Irradiation-induced precipitation is difficult to predict

theoretically for a given material under a given irradiation condition. However, it is well known that, as

irradiation temperature decreases, the size of the irradiation-induced precipitate decreases strongly and

number density increases strongly (see Fig. 7).

Increasing Dose Rate at Constant Temperature or

Increasing Temperature at Constant Dose Rate

PWR

EBR-II

Hex Can

Type A

precipitate Type B

Precipitate

Number Density of Irradiation-Induced

Precipitates

Figure 7. Schematic illustration of characteristics of irradiation-induced precipitation

sensitive to dose rate and irradiation temperature.

In austenitic SSs, several types of irradiation-induced precipitation occur under neutron irradiation,

e.g., G phase, M23C6, TiC, NbC, and γ’ precipitation. Sometimes, epsilon martensites have been reported

in a heavily swelled Type 304L steel. However, martensite precipitation is thought to be driven by the 

local deformation in the heavily swelled region rather than by an irradiation-induced process.

Irradiation-induced precipitates at the PWR-relevant temperatures of 290-380°C are extremely

small (i.e., a few to several tens of nanometers), and the number density is extremely high. Therefore,

irradiation-induced precipitation at low temperatures creates an extremely large internal surface, i.e., the

interface between the steel matrix and the precipitates. Such interface acts as a sink to irradiation-induced

vacancies, thereby suppressing the agglomeration of the vacancies. Because of this, nucleation and
15

growth of voids are greatly suppressed when irradiation-induced precipitates are produced in high number

density. This behavior has been known well for vanadium-based alloys, a group of alloys being

developed for fusion reactor structural materials. Although unalloyed vanadium exhibits large swelling, a

Ti-containing vanadium alloy, which usually contains a significant level of Si, O, N, and C as impurities,

is highly resistant to swelling. This behavior has been investigated extensively and attributed to highdensity precipitation of very fine Ti silicides.

17–19

Figure 8 (from Ref. 16) is a schematic representation of the relationship between void swelling and

the number density of irradiation-induced precipitates. For a given material at a given temperature, three

types of kinetics of irradiation-induced precipitation are shown in Fig. 8a: no precipitation (Case I), slow

(Case II), and fast kinetics (Case III). Void swelling behaviors expected for the three cases are illustrated

in Fig. 8b. This model predicts that swelling in a complex material such as an austenitic SS will be 

strongly influenced by the factors that control the kinetics of irradiation-induced precipitation,

i.e., alloying and impurity elements, temperature, dose rate, and dose.

0

10

20

30

40

50

0 10 20 30 40 50 60 70

Relative Number Density of 

Irradiation-Induced Precipitates

Dose (dpa)

Constant Irradiation Temp.

(e.g., 340°C)

Case III

Case II

Case I

(a)

0

10

20

30

40

50

0 10 20 30 40 50 60 70

Relative Void Swelling

Dose (dpa)

Constant Irradiation Temp.

(e.g., 340°C)

Case III

Case II

Case I

(b)

Figure 8. Schematic illustration of (a) three types of relationship between irradiation-induced

precipitation and (b) void swelling as a function of irradiation dose (From Ref. 19).
16

3.5.1 Low Swelling of CW Type 316 SS and High-Density Precipitation of γ’ and an Unidentified

Phase in PWR Baffle Bolts

A good correlation between low swelling and high-density precipitation is observed for CW

Type 316 SS from the behavior of Tihange baffle bolts (void swelling ≈0.24% after irradiation to 7.5 dpa,

see Section 2.4). The distribution of voids and high-density precipitation of γ’ phase is shown in Fig. 9

(from Ref. 5). In addition to γ’ phase, unknown precipitates were also observed.

(a) (b)

(c) (d)

Figure 9. Low void swelling (a and b) and high-density precipitation of γ’ phase (c) and unknown phase

(d) in the Tihange PWR baffle bolt fabricated from CW Type 316 SS (0.24 % swelling at

7.5 dpa, from Ref. 5).

3.5.2 Low Swelling in 348 SS and Dense Precipitation of NbC in a PWR

Good correlation between low void swelling and dense precipitation of very fine NbC phase is

observed from the microstructure of SA348 SS reported by Garzarolli et al.

20 The authors examined a

B4C-filled tube similar to a control rod cladding after irradiation to ≈10 dpa at ≈325°C near the core
17

center of a PWR. Void swelling shown in Fig. 3 (left) of their report is insignificant (i.e., <0.05%), and

NbC precipitates are observed in very high density.

3.5.3 Low Swelling and Dense Precipitates in SA Type 304 Irradiated to 50 dpa at 370°C in EBR-II

at Dose Rate Relevant to PWR Reentrant Corner

Good correlation between low void swelling and dense precipitation of an unidentified phase is

observed from the microstructure shown in Fig. 5 in Ref. 13. The material (SA Type 304 SS) and the 

irradiation temperature (≈370°C) in this study are similar to those of PWR reentrant corners.

3.5.4 Low Density Change and High-Density Precipitation in CW Type 316 SS Irradiated to 23-51

dpa at 375-430°C in EBR-II at Dose Rate Relevant to PWR Reentrant Corner

Various types of precipitates were reported by Cole et al.

21 for the CW 316 SS hex can that

exhibited low swelling (density change) after irradiation to 23–51 dpa at 376-430°C in EBR-II at dose

rates relevant to PWR reentrant corners (see Fig. 4, Section 3.3). The observed precipitates include Ti–

rich precipitates, carbides (M6C, M23C6, and M7C3), S-rich precipitates, and extremely small (<10 nm)

Fe–rich intermetallics. Direct correlation between the density change and the distribution of the 

precipitates (e.g., dark-field images of the various types of precipitates) was, however, not provided.

3.5.5 Low Swelling and High-Density Precipitation of G Phase and TiC in X18H10T SS Irradiated

in BN-350 at 305-355°C at Dose Rates Relevant to PWRs

Good correlation between the two phenomena is obvious in the microstructure shown in Ref. 14. A

similar behavior is expected for the western counterpart AISI 321 SS.

3.6 Swelling Behavior of Type 304L SS

Void swelling has been known to be higher in Type 304L SS than in Type 304 SS under similar

conditions. This behavior may be related to the trend that high-density irradiation-induced precipitation

of carbides is less likely to occur in Type 304L SS. However, most internals in a PWR are fabricated

from SA Type 304, CW Type 316, CW Type 347, or CW Type 348 SSs rather than Type 304L SS.

3.7 Steady-State Swelling Rate of 1%/dpa

The steady-state (or breakaway) swelling rate of 1%/dpa has been observed for austenitic SSs at

high dose if the swelling exceeds a threshold level; however, careful review of the data in Ref. 3 indicates

that such behavior was observed for western steels only at high irradiation temperatures, i.e., 427-650°C

for CW Type 316 irradiated in EBR-II, Fig. 3-6; 450, 500, and 550°C for SA316 irradiated in RAPSODIE

FBR, Figs. 3-7 and 3-27; 590-610°C for CW Type 316 irradiated in PHENIX FBR, Fig. 3-28; 538, 593,

and 650°C for Fe-15Cr-Ni ternary irradiated in EBR-II, Fig. 3-39; 400-538°C for Fe-Cr-Ni ternary

irradiated in EBR-II, Fig. 3-40; 425°C for 316 SS irradiated in the Dounreay FBR, Fig. 3-45; 450-538°C

for Type 304L SS irradiated in EBR-II, Fig. 3-51; and 510-560°C for Type 304 SS irradiated in EBR-II,

Fig. 3-52. There is no evidence that the steady-state swelling rate of 1%/dpa was observed for PWRrelevant temperatures of <380°C.
18

3.8 Threshold Swelling to Enter the Regime of Steady-State Swelling Rate

The threshold swelling to enter the regime of steady-state swelling rate (commonly referred to as

breakaway swelling rate) ranges from 3 to 5%. However, as indicated above, there is no evidence that

this threshold level of void swelling was reached in any experiment at PWR-relevant temperatures of

<380°C.
19

4 Assessment of the Potential for Void Swelling for PWR Internals

at EOL

In Section 2, currently available data from several types of PWR internals were reviewed in detail.

In all cases, void swelling was no greater than 0.42% for service temperatures <343°C and for dose levels

<65 dpa. In Section 3, void swelling and density change data from EBR-II components irradiated at

≤380°C at dose rates comparable to those of PWR internals were reviewed. Current understanding of

important factors necessary to extrapolate swelling data from fast-breeder to PWR conditions was also

reviewed. The review allows an initial assessment of EOL swelling for several types of PWR internals.

This section discusses the implications of the observations from those two sections, and applies the

recognized void swelling criteria to instrument tubes, baffle bolts and core formers - components that are

exposed to the most aggressive combinations of flux and temperature, and should be lead items in the 

development of void swelling issues for PWRs.

4.1 Thin-Walled Flux Thimble and Instrument Tubes

The effect of gamma heating is insignificant in these components (mostly fabricated from CW 316

SS), which are in contact with coolant. Therefore, irradiation temperature during service is not expected

to exceed ≈325°C. The currently available database appears to be adequate to reach a conclusion that

void swelling in this type of reactor internal is not a concern for the period of license extension.

4.2 Baffle Bolts

Most baffle bolts are fabricated from CW Type 316 SS. The database obtained from the industry

baffle bolt program shows that swelling is insignificant for dose levels up to ≈20 dpa and irradiation

temperatures up to ≈340°C. The database obtained on EBR-II components irradiated at <380°C and at

dose rates comparable to those of a baffle bolt is consistent with the data from the industry bolt program.

The low void swelling in this class of steel seems to be related to irradiation-induced formation of very

fine precipitates in very high number density. Given the data that is available at this time, it is not likely

that void swelling in this type of internal will exceed the threshold level (i.e., ≈4%) that is necessary to

enter the regime of the steady-state swelling rate of 1%/dpa, including the period of operation during

license extension.

4.3 Baffle Plate Reentrant Corners

This type of reactor internal is primarily fabricated from SA Type 304 SS and is most susceptible to

high swelling rates. The maximum irradiation temperature in some regions of a reentrant corner has been

estimated to be in the range of ≈380-420°C. Although further studies are needed to more accurately

calculate the temperature for EOL and life-extension conditions, ≈380°C is believed to be a more likely

upper limit.

No database is available for a comparable material irradiated under the PWR condition. Only one

investigation has been reported for void swelling in this class of material after irradiation in EBR-II at

370°C to ≈50 dpa at a dose rate comparable to that of a PWR reentrant corner. Void swelling in this

material was only 0.54%. The low swelling is believed to be related to high-density irradiation-induced

precipitation of very fine carbides.
20

It is not likely that void swelling in this type of reactor internal will exceed the threshold level of

swelling of ≈4% that is necessary to enter the regime of the steady-state swelling rate of 1%/dpa,

including the period of license extension. However, this estimation is only preliminary; more data from

SA Type 304 SS at PWR-relevant dose rates and a better mechanistic understanding are needed. More

accurate quantification of maximum irradiation temperature at EOL and life-extension situations is

recommended. Specimens of SS and nickel-base alloys are currently being irradiated at applicable

temperatures in an attempt to provide a more extensive database for void swelling effects. The results of

these programs are in the early stages of publication; more results are expected flow steadily for the next

few years.
5. Summary and Conclusions 

The study of void swelling for LWR conditions is very much a work in progress, and a considerable 

body of results is expected to emerge over the next several years. At this juncture, neither the reasonable 

extrapolation of results from higher temperatures and higher dose rates, nor the small but growing body of

examinations of salvaged light-water reactor core internal materials suggests that void swelling will be a 

significant problem during the first license extension period for the current fleet. The following 

observations and conclusions are based on this information: 

1. At this time, the database of void swelling for PWR internals that can be used to directly evaluate the 

behavior at end-of-life (EOL) and life-extension situations is very limited. However, significant data are 

available for thin-walled internals such as flux thimble tubes. 

2. Most swelling data that are available were obtained from steels irradiated in fast breeder reactors at 

temperatures >385°C and at dose rates that are orders of magnitude higher than PWR dose rates. Extreme 

care must be exercised when interpreting and extrapolating such data. These data should not be 

extrapolated to determine credible void swelling behavior for PWR EOL or life-extension conditions. 

3. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on 

void swelling cannot be separated. Unless it is demonstrated that interaction effects are small, it can be 

misleading to extrapolate swelling data on the basis of "progressive compounded multiplication" of

separate effects of factors such as dose, dose rate, temperature, and material composition. 

4. Limited swelling data are available for CW 316 SS irradiated to 53 dpa at 376-386°C and for 

solution-annealed Type 304 SS irradiated to 50 dpa at ≈370°C in EBR-II reflector positions at dose rates 

comparable to those of PWR reentrant corners. As such, these data are relevant to the conditions of PWR 

reentrant corners. Swelling in these materials was less than ≈1%. 

5. Low void swelling observed in the PWR components and in the EBR-II steels under PWR-relevant 

dose rates appears to be associated with irradiation-induced formation of very fine precipitates (such as G 

phase, carbides, and γ’ phase) in high number density. Such irradiation-induced precipitation at low 

temperatures (<370°C) creates an extremely large internal surface, i.e., the interface between the steel 

matrix and the precipitates. Such interface acts as an efficient sink to irradiation-induced vacancies, 

thereby suppressing the agglomeration of the vacancies. Irradiation-induced precipitation is sensitive to 

minor alloying and impurity elements, irradiation temperature, and dose rate. 

6. In thin-walled flux thimbles and instrument tubes, the effect of gamma heating is insignificant. The 

currently available database is sufficient to conclude that void swelling in this type of reactor internal, 

mostly fabricated from CW 316 SS, is not an issue. 

7. Most baffle bolts are fabricated from CW 316 SS. The data obtained from the industry baffle bolt 

program show that swelling is insignificant (<0.25%) for dose levels up to ≈20 dpa and irradiation 

temperatures of up to ≈340°C. Data obtained on EBR-II components irradiated at temperatures <380°C 

and at comparable dose rates are consistent with the data from the industry bolt program. Microstructural 

characteristics of the two groups of materials are also consistent. It is not likely that void swelling in this 

type of reactor internal will exceed the threshold level (i.e., ≈4%) that is necessary to enter the regime of the

steady-state swelling rate of 1%/dpa. 

8. Most baffle reentrant corners are fabricated from SA Type 304 SS and are most susceptible to high 

swelling rate, and hence, high swelling at EOL. The maximum irradiation temperature in some regions of 

21
the reentrant corners has been estimated to be in the range of ≈380-420°C. Only one investigation has 

reported void swelling for this class of steel after irradiation in EBR-II at 370°C to ≈50 dpa at a dose rate 

comparable to that of reentrant corners. Void swelling in this material was only 0.54%. The low swelling 

appears to be related to high-density irradiation-induced precipitation of very fine carbides. Therefore, as a

very preliminary conclusion, it is considered unlikely that void swelling in reentrant corners will exceed the 

threshold level of ≈4%. More relevant data, especially a more accurate quantification of the maximum

temperature throughout the operating life and life-extension situations of core support components, are 

needed for this class of steel. Also, needed is a better mechanistic understanding of the roles of

irradiation-induced microstructural evolution on void swelling. 

22
23

References

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3. MRP-50, “Materials Reliability Program Technical Basis Document Concerning IrradiationInduced Stress Relaxation and Void Swelling in Pressurized Water Reactor Vessel Internals

Components,” EPRI-1000970, Interim Report, October 2001

4. Byrne, S., F. A. Garner, S. Fyfitch, and I. L. Wilson, “Application of Void Swelling Data to

Evaluation of Pressurized Water Reactor Components,” Proc. 10th Intl. Conf. on Environmental

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Aug. 5-9, 2001, Lake Tahoe, NV.

5. Edwards, D. J., F. A. Garner, B. A. Oliver, and S. M. Bruemmer, “Microstructural Evaluation of a

Cold–Worked 316SS Baffle Bolty Irradiated in a Commercial PWR,” Proc. 10th Intl. Conf. on

Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, NACE/ANS/

TMS, Aug. 5-9, 2001, Lake Tahoe, NV.

6. Thomas, L., and S. Bruemmer, “Analytical Transmission Microscopy (ATEM) Characterization of 

Stress Corrosion Cracks in LWR-Irradiated Austenitic Stainless Steel Components,” EPRI1003422, Electric Power Research Institute, Palo Alto, CA, May 2002.

7. Garner, F. A., B. M. Oliver, L. B. Greenwood, D. J. Edwards, S. Bruemmer, and M. L. Grossbeck

“Generation and Retention of Helium and Hydrogen in Austenitic Steels Irradiated in a Variety of

LWR and Test-Reactor Spectral Environments,” Proc. 10th Intl. Conf. on Environmental

Degradation of Materials in Nuclear Power Systems - Water Reactors, NACE/ANS/TMS, Aug. 5-9,

2001, Lake Tahoe, NV.

8. Fujii, K., K. Fukuya, G. Furutani, T. Torimaru, A. Kohyama, and Y. Katoh, “Swelling in 316

Stainless Steel Irradiated to 53 dpa in a PWR,” Proc. 10th Intl. Conf. on Environmental

Degradation of Materials in Nuclear Power Systems - Water Reactors, NACE/ANS/TMS, Aug. 5-9,

2001, Lake Tahoe, NV.

9. Foster, J. P., D. L. Potter, D. L. Harrod, T. R. Mager, and M. G. Burke, “316 Stainless Steel Cavity

Swelling in a PWR,” J. Nucl. Mater. 224, p. 207 (1995).

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Alloys,” J. Nucl. Mater., 216 (1994) pp. 97–123.

11. Busboom, H. J., G. C. McClellan, W. L. Bell, and W. K. Appleby, “Swelling of Types 304 and 316

Stainless Steel Irradiated to 8 x 1022 n/cm2,” General Electric Co. Report GEAP–14062,

Sunnyvale, CA, 1975.
24

12. Allen, T. R., H. Tsai, R. S. Daum, D. L. Porter, J. I. Cole, T. Yoshitake, N. Akasaka, T. Donomae,

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of Materials in Nuclear Power Systems - Water Reactors, Aug. 10-14, 2003, Stevenson, WA.

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20. Garzarolli, F., P. Dewes, R. Hahn, and J. L. Nelson, "Deformability of High-Purity Stainless Steels

and Ni-Base Alloys in the Core of a PWR," in Proc. 6th Intl. Symp. on Environmental Degradation

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21. Cole, J. I., T. R. Allen, H. Tsai, S. Ukai, S. Mizuta, N. Akasaka, T. Donomae, and T. Yoshitake,

“Swelling and Microstructural Evolution in 316 Stainless Steel Hexagonal Ducts Following LongTerm Irradiation in EBR-II,” in Effects of Radiation on Materials: 20th Intl. Symp., ASTM STP

1405, S. T. Rosinski, M. L. Grossbeck, T. R. Allen, and A. S. Kumar, Eds., American Society for

Testing and Materials, West Conshohocken, PA 2001.
NRC FORM 335

(2–89)

NRCM 1102,

3201, 3202

U. S. NUCLEAR REGULATORY COMMISSION

BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse)

2. TITLE AND SUBTITLE

1. REPORT NUMBER

(Assigned by NRC. Add Vol., Supp., Rev.,

and Addendum Numbers, if any.)

NUREG/CR–6897

ANL–04/28

3. DATE REPORT PUBLISHED

MONTH YEAR

January 2006

4. FIN OR GRANT NUMBER

Assessment of Void Swelling in Austenitic Stainless Steel Core Internals

Y6388

5. AUTHOR(S) 6. TYPE OF REPORT

Technical

7. PERIOD COVERED (Inclusive Dates)

H. M. Chung

8. PERFORMING ORGANIZATION – NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if

contractor, provide name and mailing address.)

Argonne National Laboratory

9700 South Cass Avenue

Argonne, IL 60439

9. SPONSORING ORGANIZATION – NAME AND ADDRESS (If NRC, type “Same as above”: if contractor, provide NRC Division, Office or Region, U.S. Nuclear

Regulatory Commission, and mailing address.)

Division of Engineering Technology

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555–0001

10. SUPPLEMENTARY NOTES

William H. Cullen, Jr., NRC Project Manager

11. ABSTRACT (200 words or less)

As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior

of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available

database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test

procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine

the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated

in fast breeder reactors at temperatures >385°C and at dose rates that are orders of magnitude higher than PWR dose rates.

These data cannot be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or lifeextension conditions. Limited amount of swelling data and information on microstructural characteristics are available that

were obtained from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of

a PWR. Based on the information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern.

PWR baffle reentrant corners are most susceptible to high swelling rates, and hence, high swelling at EOL, especially in

limited regions where irradiation temperature is high. However, this estimation is only preliminary, and a more accurate

quantification of maximum temperature of reentrant corners at EOL and life-extension situations is needed.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating this report.) 13. AVAILABILITY STATEMENT

Unlimited

14. SECURITY CLASSIFICATION

(This Page)

Unclassified

(This Report)

Unclassified

15. NUMBER OF PAGES

Void swelling

Austenitic stainless steels

PWR Environment

16.ICE PR

NRC FORM 335 (2–89)