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Mechanical properties of irradiated beryllium
joumal of
Journal of Nuclear Materials 195 (1992) 102-10X North-Holland
Mechanical
properties
J.M. Beeston, G.R. Longhurst
llUCll!iW
maturlaki of irradiated
beryllium
and R.S. Wallace
Idaho National Engineering Laboratory, EC&G Idaho, Inc., P.O. Box 1625, Idaho Falls, ID 83415, USA
S.P. Abeln EC&G Rocky Flats, Inc., P.O. Box 464, Golden, CO 80402-0464,
USA
Received 24 February 1992; accepted 8 May 1992
Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATRJ to a neutron fluence of 2.6~ 10” n/m* (E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.
1. Introduction
appm. Tritium production is expected to be approximately 8% of the helium production.
Beryllium is being considered as a neutron multiplier in the International Thermonuclear Experimental Reactor (ITER). It may also be used as a plasma-facing material on the divertor and first wall structures. Particularly in the blanket application, densities less than 100% are considered for their reduced thermal conductivity while still performing the neutron multiplication function. Lower thermal conductivity allows the tritium breeding material to be maintained at a high temperature, thereby facilitating tritium recovery, while maintaining reasonable coolant temperature. The United States design for the ITER water-cooled solid-breeder blanket includes a temperature distribution from 90 to 500°C with a significant fraction of the beryllium at the low temperature of 100°C. Production of helium under those circumstances is estimated to be approximately 6000 appm/(1026 n/m21 [l]. Most of the beryllium will be at a temperature between 150 and 480°C with end-of-life helium production of 2000-3000 0022-3115/92/$05.00
2. Objective The objective of this study was to gain manufacturing experience with, and determine the mechanical behavior of, four densities of beryllium. Three beryllium densities (80, 85, and 97%) were obtained by cold isostatic pressing plus sintering (CIP-S). Some of the nominally 85% dense samples were actually 86.6% dense. Full density (100%) was reached by the addition of hot isostatic pressing to the CIP-S process. We also wanted to determine what changes to the ductility and strength are caused by neutron irradiation. We completed testing of two of the densities, 80 and 97%, after irradiation at 75°C. Specimens of 85% density irradiated at 420°C are awaiting decapsulation at Battelle Pacific Northwest Laboratory. The irradiation of additional specimens of the four densities over
0 1992 - Elsevier Science Publishers B.V. All rights reserved
103
J.M. Beeston et al. / Mechanical properties of irradiated Be
a wider range of temperature is in progress in the Fast Flux Test Facility (FFTF) Materials Open Test Assembly (MOTA) 2B experiment. Results from those tests will be reported separately.
Table 1
3. Procedures
Fe
Chemical analysis of beryllium powder
Element wt%
Element wt% Element wt%
Be Be0
C Al
99.1 (i:z Be 0.6 0) 0.09
Si
0.08 Mg 0.04 Other metallic 0.03
0.02 < 0.04 each
3.1. Chemistry 3.2. Processing
Manufacture of the four densities was accomplished at EG&G Rocky Flats. The beryllium powder used was Brush Wellman standard structural grade SP-200F. The powder was 99.9% - 325 mesh (- 44 km) with the chemistry (analysis conducted at Brush Wellman) shown in table 1.
Fig. 1. Photomicrographs
The manufacture of porous beryllium to different densities can be accomplished with powder metallurgy techniques. These porous Be samples were all CIP’ed identically and then sintered at different temperatures to achieve the desired density. CIP’ed billets were
showing porosity associated with (beginning at upper left and going clockwise) 100, 97, 85, and 80% dense beryllium.
104 Table 2 Production
J.M. Beeston et al. / Mechanical
temperatures
and times for the four densities
Sample density
CIP (pressure/time)
- 80% - 85% - 97% - 100%
413 413 413 413
MPa/lS MPa/lS MPa/lS MPa/lS
min min min min
of irradiated Be
achieved
Sinter (temperature/time/vacuum)
HIP tpressure/temp.l
105O”C/4 llOO”C/4 125O”C/4 125O”C/4
N/A N/A N/A 103 MPa/lOOO”C/3
made from SP-200-F powder using a 51-mm (2-inch) diameter by 457-mm (l&inch) long polyurethane bag in which the powder was pressed. The polyurethane had a Shore A hardness of 60-65. After vibratory packing, residual air was removed and the bag was sealed and CIP’ed at 414 MPa (60 ksi) for 15 min. The CIP process converts the loose pack powder, which is approximately 55% dense, to a compact that is approximately 80% dense. At this stage the CIP billet is rigid and can be handled and/or machined. The CIP’ed billets were sintered under vacuum (13 mPa) at different temperatures to achieve the desired density. The processing sequence for each density is given in table 2. The sintering step controls the resultant density by varying temperature or time. For these specimens, the temperature was varied between 1050 and 1250°C and the time was held constant at 4 h. Successful sintering requires a controlled heatup cycle. During heatup, residual gasses contained in the CIP billet are released
Fig. 2. Photomicrograph
properties
of 86.6% CIP/sinter
h/10m4 h/10m4 h/10-’ h/10m4
Torr Torr Torr Torr
h
to the vacuum system. If the heatup cycle is too fast, the parent metal will be oxidized. The CIP billets were placed in the vacuum furnace the day before temperature cycling and evacuated to 1.3 mPa. This vacuum was maintained for a minimum of 12 h. Approximately 2 h was required for heatup to the final sintering temperature. The vacuum was monitored closely during this initial heatup such that the temperature was stabilized or reduced at points of major gas evolution (i.e. when the vacuum gauge indicated pressure above 13 mPa1. Hot isostatic pressing was used as a final step in the 100% dense samples. Hot isostatic pressing closes all residual porosity remaining after sintering. The pressure, temperature, and time parameters were 103 MPa, lOOO”C,and 3 h, respectively. The production of individual specimen shapes by the CIP-S method was not pursued due to funding and time constraints. A fabrication cost assessment report
Be. Note that there were areas of higher and lower density.
J.M. Beeston et al. / Mechanical properties of irradiated Be
produced with the cooperation of Brush Wellman, Elmore, OH contains a discussion of approximate costs for block pieces [2]. 3.3. Metallography We conducted metallography on specimens taken from the blocks with the typical porosity distributions shown in fig. 1. The 86.6% dense material had some areas of higher density as shown in fig. 2. Variations in density in a block are probably due to powder segregation in the vibrated compact with agglomeration of fines. For example, calculation of the densities of some samples from weight and dimensions gave 82% instead of 80% and 99% instead of 97%. Fines were generally richer in Al and Si which are sintering aids. The over abundance of fines resulted in higher density at a given sintering temperature. During cold isostatic pressing, the powder particles were severely cold worked, and upon sintering at the lower temperatures ( u 1050°C) recrystallization occurred with a fine grain structure. 3.4. Mechanical
preparation
Specimens for tensile tests at EG&G Rocky Flats were machined under oil to better contain particulate dispersal during the machining process. Machining under oil should also reduce machining damage that could influence tensile testing, although some machin-
Table 3 Measured
mechanical
Sample density
Condition
100%
Control
91%
Control Irradiated
86.6% 85% 80%
Control Control Control Irradiated
properties
for beryllium
ing damage was observed on all the tensile test specimens. The density was determined after initial machining of the block. The block was measured to within 0.025 mm, weighed to the nearest tenth of a gram, and the density was determined by the ratio of the weight to volume. We calculated the fractional density of the samples by assuming a theoretical density of 1.854 g/cm3 corresponding to beryllium with 1% BeO. The densities of the specimens from the blocks varied slightly. The mechanical properties of the various samples are given in table 3. The specimens for use as control samples and for irradiation were dry machined and measured. From the four blocks of different density (8.5, 90, 97, 100%) 240 specimens were machined, all 7.6-mm diameter. Compression test specimens were 20-mm long while those for splitting tension tests or tritium analysis were 6-mm long. 3.5. Irradiation The irradiation of three sets of samples was initiated. The tests reported on here used samples irradiated in the Advanced Test Reactor (ATR) A-10 and A-l 1 positions from April 29, 1990 to August 5, 1991 at a temperature of 75°C. The set consisted of six compression samples and six splitting tension or tritium analysis samples with one half of the set of samples 80% dense and the other half 97% dense. The samples attained a fluence of 2.6 x 10z5 n/m* (E > 1 MeV).
samples Yield strength (0.2%, MPa)
Ultimate strength (MPa)
Ductility
type Tension Compression Tension Compression Splitting Compression Tension Tension Tension Compression Splitting Compression
245+10 230+ 11 234 + 24 216k 7 130* 4 928+ 4 205 + 32 211+19 162+ 13 140+ 3 88f 4 611 k58
318+ 5 458+44 b 302k 12 310+ 10 h 982+ 245 k 240 f 167k 204k
Test
105
a At end of test. h Calculation of ultimate strength at 1.5% strain to preclude ’ At drop in load or end of test, whichever occurred first.
11 42 20 14 5 h
633 f 57
barrelling
effects.
Linear swelling AL/L (%)
Number of tests
0.6&0.05 a 2.7 0.6kO.l a 3.1
_
4 2 4 5
0.6 f 0.05 ’ 0.93 f 0.05 a 0.37+0.06 a 0.25kO.l a 1.74
0.18~0.05 -
f%)
0.4 f 0.04 c
_
_ 0.18+0.04
3 4 3 5 5 2 3
1Oh
J.M. Beeston et ul. / Mechanical
Measurements of the helium and tritium contents of two of these samples by Baldwin [3] of Battelle Pacific Northwest Laboratory gave 4He at 733 and 872 appm and ‘H at 55.3 and 71.8 appm for the 80 and 97% dense specimens, respectively. These values were roughly two-thirds of the a priori calculated values of 1200 appm helium and 100 appm tritium, respectively.
properties of irradiated Be Table 4
Hardness and surface
roughness
Density (o/o)
Rockwell hardness
100 97 85 80
83.6 k 0.7 81.8+_0.3 89.1 k 1.4 29.3 f 0.5
B
data Surface roughness (rms km) 1.3 1.3 1.6 2.5
4. Test results The compression tests on the control and irradiated specimens from the ATR A-10 and A-11 positions were conducted at room temperature on an MTS Test Frame according to ASTM E-9 “Standard Test Methods of Compression Testing of Metallic Materials at Room Temperature”. The results of control specimen compression tests are given in table 3. The compression tests on the irradiated specimens were conducted at a strain rate of 0.004 m/m/min. Length measurements were made on the irradiated specimens before testing from which a swelling value, A L/L, was calculated. Since hot pressed beryllium exhibits generally isotropic swelling behavior, 3AL/L = AV/V. The results of compression tests on irradiated samples are given in table 3. The splitting tension tests on control and irradiated specimens were conducted according to ASTM D-3967, “Standard Test Method for Splitting Tensile Strength of Intact Rock Core Specimens”. In these tests, cylindrical specimens are placed between parallel platens, with the specimen axis parallel to the platen faces, and loaded in compression. This induces a zone of tensile stress perpendicular to the plane defined by the lines of contact of the test cylinder with the platens. If failure occurs in the tensile zone as the compression load is increased, then these tests can be regarded as an indirect measure of tensile strength. We recognized that the control (unirradiated) specimens would not qualify for a valid test because of their ductility (elongation of 0.25 to 0.6% - see table 3). However, because of the low ductility (0.25%) of the 80% density samples, we were interested in making this test. Two of the 80% density unirradiated samples failed in brittle fracture at a stress of 32 f 10 MPa while two more failed in the ductile mode as also did one of the 97% dense samples. The specimens that failed in the ductile mode did not exhibit the center diameter cracking characteristic of a valid splitting tension stress pattern. We give the results of the splitting tension tests by diametral line compression on the irradiated speci-
mens also in table 3. The irradiated specimens failed suddenly (nonductile) with the characteristic diametral line failure. One half of the cylinder exhibited secondary cracks, while one half appeared to have little secondary cracking. The applied strain rate was 0.004 m/m/min. 4.1. Hardness and surface roughness The hardness and surface roughness were measured on machined samples of each of the four densities. The samples were 7.6 mm in diameter by 6 mm long. The Rockwell B hardness, measured in four places, and the surface roughness data in rms micrometers are given in table 4. 4.2. Tritium release One of the concerns associated with structural failure of irradiated beryllium is the potential for tritium release. Analysis by Longhurst [4] of release measurement data by Baldwin [l] indicated a low solubility of tritium in the beryllium itself, but tritium produced in the beryllium, mainly from the reaction chain beginning with the ‘Be (n,a) 6He reaction, seems to reside mainly in microvoids and helium bubbles with some in oxide inclusions. The chemical state of the tritium released depends largely on which reservoir it comes from. Tritium coming from the voids and bubbles is mainly elemental gas while tritium coming from oxide inclusions tends to come out as tritiated water, HTO. When beryllium containing tritium experiences mechanical failure, we expect that some tritium will be released. To get some idea of the nature and magnitude of the release, the air in the vicinity of the beryllium specimen was sampled continuously with the sample stream routed through a dual-ion chamber tritium monitor and then through a water bubbler. The dualion chamber was for gamma compensation. The observed signal due to gamma effects in the ion chamber
J. M. Beeston et al. / Mechanical properties of irradiated Be
was of the same order as that from tritium. General observations were that for the high-density irradiated specimens where failure was generally by plastic deformation, although there was perceptible tritium release, it was very small, barely detectable above background. By comparison, when the low-density beryllium failed, there was massive brittle fracture, and a large number of fragments of various sizes were produced. Production of the free surface area was invariably associated with a substantial tritium release. It was not possible to get accurate quantitative information on the magnitude of the release because, as was apparent from smoke drift patterns, high air flows prevented thorough mixing of the tritium in the glovebox. Also, transient times for both the volume change in the ion chamber and the glovebox are nearly identical. That made it difficult to distinguish classical, fully-mixed dilution from a scenario where the tritium was released in a relative dense plume that was quickly transported to the vent. While the sampling point was located such as to be in the midst of the plume, only order of magnitude estimates of total release were possible. Our estimate was that a total of 217 t.t.Ci was released in our testing, of which 211 pCi came from two 80% dense sample failures. That corresponds to something on the order of 5 X 1O-4 of the tritium believed to be in the specimens. Tritium content in the water bubbler was only 68 k 14 nCi suggesting that less than 0.1% of the tritium coming from the samples was in the oxide form.
5. Discussion The ductility at the tensile yield strength of beryllium is reported by Martin and Ellis [5] to be low because of limited slip systems, so that plastic flow is mainly due to twinning. The ductility of beryllium in compression is notably higher than in tension tests. Upon irradiation at low exposures, the production of helium reduces the ductility due to the large amounts of helium produced and the pinning effects of the interstitial helium atoms. For this reason, and the added advantage of being able to use shorter specimen lengths in testing than would be required for tension tests (hence using less irradiation space per test specimen), we chose com.pression testing to investigate the mechanical properties of irradiated beryllium in this work. Although the ductility was higher in compression, the tensile yield strength at 0.2% offset was lower in compression than in tension for unirradiated material.
107
In comparing the tensile yield strength with the compression yield strength for the control specimens in table 3, one may see that the compression yield strength of unirradiated beryllium was only about 91% of its tensile yield strength while the compression ductility was four to six times greater (specimens did not buckle). The tensile strength of irradiated beryllium is difficult to measure at fluences greater than 1 X 10’” n/m* (E > 1 MeV) (helium contents of about 450 appm) [6,7]. However, the splitting tensile test, which produces diametral line compression allows a measurcmcnt to be made of the tensile strength in complex stress fields. Comparing data in table 3, the splitting tensile (yield) strength of the 80% dense irradiated specimens (helium content of about 730 appm) was found to be about 54% of the unirradiated specimen yield strength. For the 97% dense material the splitting tensile (yield) strength of the irradiated samples was 55% of the unirradiated specimen yield strength. The 80% initial density beryllium had lower strengths in both irradiated and unirradiated conditions than the 97% dense material. A reduction in the mechanical and physical property values with increased porosity and lower density was expected. In carlier work [8] it was shown that except for the fracture toughness and notched Charpy impact, the property value decrease with porosity was generally linear for unirradiated samples. With data for only two densities WC cannot determine such linearity for the irradiated material. Comparing the compression test data for the irradiated samples with data for the unirradiated specimens in table 3, one may see there was approximately a four-fold increase in the compression yield strength with a four-fold decrease in ductility. The bulk swelling, AV/V= 3A L/L of 0.54% was unexpectedly high. Comparison of the present measurements with other work [6,9] on beryllium at equivalent fluence and temperature revealed similar compression yield strength and ductility. The swelling, however, was a factor of four times higher than that given by the previously determined equation [9] AV/V(%)
= 0.549(@t)“““‘,
(li
where @t is the fluence in units of 10Zh n/m* (E > 1 MeV). We note that the swelling of the 80% dense beryllium was the same as that for the 97% dense beryllium. We rationalized that swelling must be due to operation of the gas laws in the closed pores or bubbles of both the 80%- and 97%-dense materials. Bubble behavior in irradiated beryllium has received considerable attention [lO,ll]. We expected some imprecision on the
10X
J.M. Beeston CI al. / Mrchund
swelling values since the accuracy of the measurement allowed some variation. However, the repeatability of the three tests for each density was very good. The evolution in the elemental form of most of the tritium that comes off when low-density specimens failed was consistent with the notion that it was resident in voids and bubbles, or closed porosities as opposed to being incorporated in hydroxides at oxide inclusions. When fracture occurred, it probably occurred along agglomeration surfaces for these voids and bubbles. That allowed tritium in the exposed dcfects to escape with effectively no activation energy. Tritium that may have been chemically bound to oxides would have required thermal energy to escape. The essential lack of tritium emission during plastic failures supports this hypothesis.
Acknowledgement This work was supported by the US Department of Energy, Office of Fusion Energy, Under DOE Idaho
Field Office Contract DE-A607-76ID01570. We also express appreciation for the excellent support services we received in the process of conducting the experiments.
properties
of irrudiatedBe
References [l] M.C. Billone, C.C. Lin and D.L. Baldwin, Fusion Technol. 19 (1991) 1707. [2] J.M. Beeston, G.R. Longhurst and T. Parsonage, Berylium Fabrication/Cost Assessment for ITER, Idaho National Engineering Laboratory Report, EGG-FSP-9120 (1990). [31 D.L. Baldwin, Second Workshop on Ceramic Breeder Interactions, Clearwater, Florida, November 22-23, 1991, PNL-SA-20198 S. [41 G.R. Longhurst, Behavior of Tritium in ITER Beryllium, Idaho National Engineering Laboratory Report, EGGFSP-9304 (1990). 151 A.J. Martin and G.C. Ellis, in The Metallurgy of Beryllium (The Institute of Metals, Chapman and Hall, Ltd., London, 1963) p. 3. [61 J.M. Beeston, L.G. Miller, E.L. Wood, Jr. and R.W. Moir, J. Nucl. Mater. 122 & 123 (1984) 862. 171 C.E. Ells and E.C.W. Perryman, J. Nucl. Mater. 1 (1959) 73. Nl D. Beasley and R.E. Cooper, Fourth International Conference on Beryllium, Metals Society, London, October 4-7, 1977, Paper 24. [91 J.M. Beeston, Properties of Irradiated Beryllium - Statistical Evaluation, Idaho National Engineering Laboratory Report, RE-E-76-031 (1976). 1101J.M. Beeston. G.R. Longhurst, L.G. Miller and R.A. Causey, Gas Retention in Irradiated Beryllium, EGGFSP-9125 (1990). [ill R.S. Barnes and R.S. Nelson, Theories of Swelling and Gas Retention in Reactor Material Proc., Metallurgical Society Conferences, Asheville, NC, September 1965, p. 22s.
Journal of Nuclear Materials 195 (1992) 102-10X North-Holland
Mechanical
properties
J.M. Beeston, G.R. Longhurst
llUCll!iW
maturlaki of irradiated
beryllium
and R.S. Wallace
Idaho National Engineering Laboratory, EC&G Idaho, Inc., P.O. Box 1625, Idaho Falls, ID 83415, USA
S.P. Abeln EC&G Rocky Flats, Inc., P.O. Box 464, Golden, CO 80402-0464,
USA
Received 24 February 1992; accepted 8 May 1992
Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATRJ to a neutron fluence of 2.6~ 10” n/m* (E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.
1. Introduction
appm. Tritium production is expected to be approximately 8% of the helium production.
Beryllium is being considered as a neutron multiplier in the International Thermonuclear Experimental Reactor (ITER). It may also be used as a plasma-facing material on the divertor and first wall structures. Particularly in the blanket application, densities less than 100% are considered for their reduced thermal conductivity while still performing the neutron multiplication function. Lower thermal conductivity allows the tritium breeding material to be maintained at a high temperature, thereby facilitating tritium recovery, while maintaining reasonable coolant temperature. The United States design for the ITER water-cooled solid-breeder blanket includes a temperature distribution from 90 to 500°C with a significant fraction of the beryllium at the low temperature of 100°C. Production of helium under those circumstances is estimated to be approximately 6000 appm/(1026 n/m21 [l]. Most of the beryllium will be at a temperature between 150 and 480°C with end-of-life helium production of 2000-3000 0022-3115/92/$05.00
2. Objective The objective of this study was to gain manufacturing experience with, and determine the mechanical behavior of, four densities of beryllium. Three beryllium densities (80, 85, and 97%) were obtained by cold isostatic pressing plus sintering (CIP-S). Some of the nominally 85% dense samples were actually 86.6% dense. Full density (100%) was reached by the addition of hot isostatic pressing to the CIP-S process. We also wanted to determine what changes to the ductility and strength are caused by neutron irradiation. We completed testing of two of the densities, 80 and 97%, after irradiation at 75°C. Specimens of 85% density irradiated at 420°C are awaiting decapsulation at Battelle Pacific Northwest Laboratory. The irradiation of additional specimens of the four densities over
0 1992 - Elsevier Science Publishers B.V. All rights reserved
103
J.M. Beeston et al. / Mechanical properties of irradiated Be
a wider range of temperature is in progress in the Fast Flux Test Facility (FFTF) Materials Open Test Assembly (MOTA) 2B experiment. Results from those tests will be reported separately.
Table 1
3. Procedures
Fe
Chemical analysis of beryllium powder
Element wt%
Element wt% Element wt%
Be Be0
C Al
99.1 (i:z Be 0.6 0) 0.09
Si
0.08 Mg 0.04 Other metallic 0.03
0.02 < 0.04 each
3.1. Chemistry 3.2. Processing
Manufacture of the four densities was accomplished at EG&G Rocky Flats. The beryllium powder used was Brush Wellman standard structural grade SP-200F. The powder was 99.9% - 325 mesh (- 44 km) with the chemistry (analysis conducted at Brush Wellman) shown in table 1.
Fig. 1. Photomicrographs
The manufacture of porous beryllium to different densities can be accomplished with powder metallurgy techniques. These porous Be samples were all CIP’ed identically and then sintered at different temperatures to achieve the desired density. CIP’ed billets were
showing porosity associated with (beginning at upper left and going clockwise) 100, 97, 85, and 80% dense beryllium.
104 Table 2 Production
J.M. Beeston et al. / Mechanical
temperatures
and times for the four densities
Sample density
CIP (pressure/time)
- 80% - 85% - 97% - 100%
413 413 413 413
MPa/lS MPa/lS MPa/lS MPa/lS
min min min min
of irradiated Be
achieved
Sinter (temperature/time/vacuum)
HIP tpressure/temp.l
105O”C/4 llOO”C/4 125O”C/4 125O”C/4
N/A N/A N/A 103 MPa/lOOO”C/3
made from SP-200-F powder using a 51-mm (2-inch) diameter by 457-mm (l&inch) long polyurethane bag in which the powder was pressed. The polyurethane had a Shore A hardness of 60-65. After vibratory packing, residual air was removed and the bag was sealed and CIP’ed at 414 MPa (60 ksi) for 15 min. The CIP process converts the loose pack powder, which is approximately 55% dense, to a compact that is approximately 80% dense. At this stage the CIP billet is rigid and can be handled and/or machined. The CIP’ed billets were sintered under vacuum (13 mPa) at different temperatures to achieve the desired density. The processing sequence for each density is given in table 2. The sintering step controls the resultant density by varying temperature or time. For these specimens, the temperature was varied between 1050 and 1250°C and the time was held constant at 4 h. Successful sintering requires a controlled heatup cycle. During heatup, residual gasses contained in the CIP billet are released
Fig. 2. Photomicrograph
properties
of 86.6% CIP/sinter
h/10m4 h/10m4 h/10-’ h/10m4
Torr Torr Torr Torr
h
to the vacuum system. If the heatup cycle is too fast, the parent metal will be oxidized. The CIP billets were placed in the vacuum furnace the day before temperature cycling and evacuated to 1.3 mPa. This vacuum was maintained for a minimum of 12 h. Approximately 2 h was required for heatup to the final sintering temperature. The vacuum was monitored closely during this initial heatup such that the temperature was stabilized or reduced at points of major gas evolution (i.e. when the vacuum gauge indicated pressure above 13 mPa1. Hot isostatic pressing was used as a final step in the 100% dense samples. Hot isostatic pressing closes all residual porosity remaining after sintering. The pressure, temperature, and time parameters were 103 MPa, lOOO”C,and 3 h, respectively. The production of individual specimen shapes by the CIP-S method was not pursued due to funding and time constraints. A fabrication cost assessment report
Be. Note that there were areas of higher and lower density.
J.M. Beeston et al. / Mechanical properties of irradiated Be
produced with the cooperation of Brush Wellman, Elmore, OH contains a discussion of approximate costs for block pieces [2]. 3.3. Metallography We conducted metallography on specimens taken from the blocks with the typical porosity distributions shown in fig. 1. The 86.6% dense material had some areas of higher density as shown in fig. 2. Variations in density in a block are probably due to powder segregation in the vibrated compact with agglomeration of fines. For example, calculation of the densities of some samples from weight and dimensions gave 82% instead of 80% and 99% instead of 97%. Fines were generally richer in Al and Si which are sintering aids. The over abundance of fines resulted in higher density at a given sintering temperature. During cold isostatic pressing, the powder particles were severely cold worked, and upon sintering at the lower temperatures ( u 1050°C) recrystallization occurred with a fine grain structure. 3.4. Mechanical
preparation
Specimens for tensile tests at EG&G Rocky Flats were machined under oil to better contain particulate dispersal during the machining process. Machining under oil should also reduce machining damage that could influence tensile testing, although some machin-
Table 3 Measured
mechanical
Sample density
Condition
100%
Control
91%
Control Irradiated
86.6% 85% 80%
Control Control Control Irradiated
properties
for beryllium
ing damage was observed on all the tensile test specimens. The density was determined after initial machining of the block. The block was measured to within 0.025 mm, weighed to the nearest tenth of a gram, and the density was determined by the ratio of the weight to volume. We calculated the fractional density of the samples by assuming a theoretical density of 1.854 g/cm3 corresponding to beryllium with 1% BeO. The densities of the specimens from the blocks varied slightly. The mechanical properties of the various samples are given in table 3. The specimens for use as control samples and for irradiation were dry machined and measured. From the four blocks of different density (8.5, 90, 97, 100%) 240 specimens were machined, all 7.6-mm diameter. Compression test specimens were 20-mm long while those for splitting tension tests or tritium analysis were 6-mm long. 3.5. Irradiation The irradiation of three sets of samples was initiated. The tests reported on here used samples irradiated in the Advanced Test Reactor (ATR) A-10 and A-l 1 positions from April 29, 1990 to August 5, 1991 at a temperature of 75°C. The set consisted of six compression samples and six splitting tension or tritium analysis samples with one half of the set of samples 80% dense and the other half 97% dense. The samples attained a fluence of 2.6 x 10z5 n/m* (E > 1 MeV).
samples Yield strength (0.2%, MPa)
Ultimate strength (MPa)
Ductility
type Tension Compression Tension Compression Splitting Compression Tension Tension Tension Compression Splitting Compression
245+10 230+ 11 234 + 24 216k 7 130* 4 928+ 4 205 + 32 211+19 162+ 13 140+ 3 88f 4 611 k58
318+ 5 458+44 b 302k 12 310+ 10 h 982+ 245 k 240 f 167k 204k
Test
105
a At end of test. h Calculation of ultimate strength at 1.5% strain to preclude ’ At drop in load or end of test, whichever occurred first.
11 42 20 14 5 h
633 f 57
barrelling
effects.
Linear swelling AL/L (%)
Number of tests
0.6&0.05 a 2.7 0.6kO.l a 3.1
_
4 2 4 5
0.6 f 0.05 ’ 0.93 f 0.05 a 0.37+0.06 a 0.25kO.l a 1.74
0.18~0.05 -
f%)
0.4 f 0.04 c
_
_ 0.18+0.04
3 4 3 5 5 2 3
1Oh
J.M. Beeston et ul. / Mechanical
Measurements of the helium and tritium contents of two of these samples by Baldwin [3] of Battelle Pacific Northwest Laboratory gave 4He at 733 and 872 appm and ‘H at 55.3 and 71.8 appm for the 80 and 97% dense specimens, respectively. These values were roughly two-thirds of the a priori calculated values of 1200 appm helium and 100 appm tritium, respectively.
properties of irradiated Be Table 4
Hardness and surface
roughness
Density (o/o)
Rockwell hardness
100 97 85 80
83.6 k 0.7 81.8+_0.3 89.1 k 1.4 29.3 f 0.5
B
data Surface roughness (rms km) 1.3 1.3 1.6 2.5
4. Test results The compression tests on the control and irradiated specimens from the ATR A-10 and A-11 positions were conducted at room temperature on an MTS Test Frame according to ASTM E-9 “Standard Test Methods of Compression Testing of Metallic Materials at Room Temperature”. The results of control specimen compression tests are given in table 3. The compression tests on the irradiated specimens were conducted at a strain rate of 0.004 m/m/min. Length measurements were made on the irradiated specimens before testing from which a swelling value, A L/L, was calculated. Since hot pressed beryllium exhibits generally isotropic swelling behavior, 3AL/L = AV/V. The results of compression tests on irradiated samples are given in table 3. The splitting tension tests on control and irradiated specimens were conducted according to ASTM D-3967, “Standard Test Method for Splitting Tensile Strength of Intact Rock Core Specimens”. In these tests, cylindrical specimens are placed between parallel platens, with the specimen axis parallel to the platen faces, and loaded in compression. This induces a zone of tensile stress perpendicular to the plane defined by the lines of contact of the test cylinder with the platens. If failure occurs in the tensile zone as the compression load is increased, then these tests can be regarded as an indirect measure of tensile strength. We recognized that the control (unirradiated) specimens would not qualify for a valid test because of their ductility (elongation of 0.25 to 0.6% - see table 3). However, because of the low ductility (0.25%) of the 80% density samples, we were interested in making this test. Two of the 80% density unirradiated samples failed in brittle fracture at a stress of 32 f 10 MPa while two more failed in the ductile mode as also did one of the 97% dense samples. The specimens that failed in the ductile mode did not exhibit the center diameter cracking characteristic of a valid splitting tension stress pattern. We give the results of the splitting tension tests by diametral line compression on the irradiated speci-
mens also in table 3. The irradiated specimens failed suddenly (nonductile) with the characteristic diametral line failure. One half of the cylinder exhibited secondary cracks, while one half appeared to have little secondary cracking. The applied strain rate was 0.004 m/m/min. 4.1. Hardness and surface roughness The hardness and surface roughness were measured on machined samples of each of the four densities. The samples were 7.6 mm in diameter by 6 mm long. The Rockwell B hardness, measured in four places, and the surface roughness data in rms micrometers are given in table 4. 4.2. Tritium release One of the concerns associated with structural failure of irradiated beryllium is the potential for tritium release. Analysis by Longhurst [4] of release measurement data by Baldwin [l] indicated a low solubility of tritium in the beryllium itself, but tritium produced in the beryllium, mainly from the reaction chain beginning with the ‘Be (n,a) 6He reaction, seems to reside mainly in microvoids and helium bubbles with some in oxide inclusions. The chemical state of the tritium released depends largely on which reservoir it comes from. Tritium coming from the voids and bubbles is mainly elemental gas while tritium coming from oxide inclusions tends to come out as tritiated water, HTO. When beryllium containing tritium experiences mechanical failure, we expect that some tritium will be released. To get some idea of the nature and magnitude of the release, the air in the vicinity of the beryllium specimen was sampled continuously with the sample stream routed through a dual-ion chamber tritium monitor and then through a water bubbler. The dualion chamber was for gamma compensation. The observed signal due to gamma effects in the ion chamber
J. M. Beeston et al. / Mechanical properties of irradiated Be
was of the same order as that from tritium. General observations were that for the high-density irradiated specimens where failure was generally by plastic deformation, although there was perceptible tritium release, it was very small, barely detectable above background. By comparison, when the low-density beryllium failed, there was massive brittle fracture, and a large number of fragments of various sizes were produced. Production of the free surface area was invariably associated with a substantial tritium release. It was not possible to get accurate quantitative information on the magnitude of the release because, as was apparent from smoke drift patterns, high air flows prevented thorough mixing of the tritium in the glovebox. Also, transient times for both the volume change in the ion chamber and the glovebox are nearly identical. That made it difficult to distinguish classical, fully-mixed dilution from a scenario where the tritium was released in a relative dense plume that was quickly transported to the vent. While the sampling point was located such as to be in the midst of the plume, only order of magnitude estimates of total release were possible. Our estimate was that a total of 217 t.t.Ci was released in our testing, of which 211 pCi came from two 80% dense sample failures. That corresponds to something on the order of 5 X 1O-4 of the tritium believed to be in the specimens. Tritium content in the water bubbler was only 68 k 14 nCi suggesting that less than 0.1% of the tritium coming from the samples was in the oxide form.
5. Discussion The ductility at the tensile yield strength of beryllium is reported by Martin and Ellis [5] to be low because of limited slip systems, so that plastic flow is mainly due to twinning. The ductility of beryllium in compression is notably higher than in tension tests. Upon irradiation at low exposures, the production of helium reduces the ductility due to the large amounts of helium produced and the pinning effects of the interstitial helium atoms. For this reason, and the added advantage of being able to use shorter specimen lengths in testing than would be required for tension tests (hence using less irradiation space per test specimen), we chose com.pression testing to investigate the mechanical properties of irradiated beryllium in this work. Although the ductility was higher in compression, the tensile yield strength at 0.2% offset was lower in compression than in tension for unirradiated material.
107
In comparing the tensile yield strength with the compression yield strength for the control specimens in table 3, one may see that the compression yield strength of unirradiated beryllium was only about 91% of its tensile yield strength while the compression ductility was four to six times greater (specimens did not buckle). The tensile strength of irradiated beryllium is difficult to measure at fluences greater than 1 X 10’” n/m* (E > 1 MeV) (helium contents of about 450 appm) [6,7]. However, the splitting tensile test, which produces diametral line compression allows a measurcmcnt to be made of the tensile strength in complex stress fields. Comparing data in table 3, the splitting tensile (yield) strength of the 80% dense irradiated specimens (helium content of about 730 appm) was found to be about 54% of the unirradiated specimen yield strength. For the 97% dense material the splitting tensile (yield) strength of the irradiated samples was 55% of the unirradiated specimen yield strength. The 80% initial density beryllium had lower strengths in both irradiated and unirradiated conditions than the 97% dense material. A reduction in the mechanical and physical property values with increased porosity and lower density was expected. In carlier work [8] it was shown that except for the fracture toughness and notched Charpy impact, the property value decrease with porosity was generally linear for unirradiated samples. With data for only two densities WC cannot determine such linearity for the irradiated material. Comparing the compression test data for the irradiated samples with data for the unirradiated specimens in table 3, one may see there was approximately a four-fold increase in the compression yield strength with a four-fold decrease in ductility. The bulk swelling, AV/V= 3A L/L of 0.54% was unexpectedly high. Comparison of the present measurements with other work [6,9] on beryllium at equivalent fluence and temperature revealed similar compression yield strength and ductility. The swelling, however, was a factor of four times higher than that given by the previously determined equation [9] AV/V(%)
= 0.549(@t)“““‘,
(li
where @t is the fluence in units of 10Zh n/m* (E > 1 MeV). We note that the swelling of the 80% dense beryllium was the same as that for the 97% dense beryllium. We rationalized that swelling must be due to operation of the gas laws in the closed pores or bubbles of both the 80%- and 97%-dense materials. Bubble behavior in irradiated beryllium has received considerable attention [lO,ll]. We expected some imprecision on the
10X
J.M. Beeston CI al. / Mrchund
swelling values since the accuracy of the measurement allowed some variation. However, the repeatability of the three tests for each density was very good. The evolution in the elemental form of most of the tritium that comes off when low-density specimens failed was consistent with the notion that it was resident in voids and bubbles, or closed porosities as opposed to being incorporated in hydroxides at oxide inclusions. When fracture occurred, it probably occurred along agglomeration surfaces for these voids and bubbles. That allowed tritium in the exposed dcfects to escape with effectively no activation energy. Tritium that may have been chemically bound to oxides would have required thermal energy to escape. The essential lack of tritium emission during plastic failures supports this hypothesis.
Acknowledgement This work was supported by the US Department of Energy, Office of Fusion Energy, Under DOE Idaho
Field Office Contract DE-A607-76ID01570. We also express appreciation for the excellent support services we received in the process of conducting the experiments.
properties
of irrudiatedBe
References [l] M.C. Billone, C.C. Lin and D.L. Baldwin, Fusion Technol. 19 (1991) 1707. [2] J.M. Beeston, G.R. Longhurst and T. Parsonage, Berylium Fabrication/Cost Assessment for ITER, Idaho National Engineering Laboratory Report, EGG-FSP-9120 (1990). [31 D.L. Baldwin, Second Workshop on Ceramic Breeder Interactions, Clearwater, Florida, November 22-23, 1991, PNL-SA-20198 S. [41 G.R. Longhurst, Behavior of Tritium in ITER Beryllium, Idaho National Engineering Laboratory Report, EGGFSP-9304 (1990). 151 A.J. Martin and G.C. Ellis, in The Metallurgy of Beryllium (The Institute of Metals, Chapman and Hall, Ltd., London, 1963) p. 3. [61 J.M. Beeston, L.G. Miller, E.L. Wood, Jr. and R.W. Moir, J. Nucl. Mater. 122 & 123 (1984) 862. 171 C.E. Ells and E.C.W. Perryman, J. Nucl. Mater. 1 (1959) 73. Nl D. Beasley and R.E. Cooper, Fourth International Conference on Beryllium, Metals Society, London, October 4-7, 1977, Paper 24. [91 J.M. Beeston, Properties of Irradiated Beryllium - Statistical Evaluation, Idaho National Engineering Laboratory Report, RE-E-76-031 (1976). 1101J.M. Beeston. G.R. Longhurst, L.G. Miller and R.A. Causey, Gas Retention in Irradiated Beryllium, EGGFSP-9125 (1990). [ill R.S. Barnes and R.S. Nelson, Theories of Swelling and Gas Retention in Reactor Material Proc., Metallurgical Society Conferences, Asheville, NC, September 1965, p. 22s.