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AECL-9999

ATOMIC ENERGY ( ^7 \'">•) ENERGIEATOMIQUE OF CANADA LIMITED VJ^A / DU CANADA LIMITEE

FUSION-BLANKET FABRICATION DEVELOPMENT AND IRRADIATION TESTING

COUVERTURE DE REACTEUR A FUSION - MISE AU POINT DE LA FABRICATION ET ESSAIS SOUS IRRADIATION

I.J. HASTINGS, A.J. ELLIOT, J.M. MILLER, B.J.F. PALMER and R.A. VERRALL

Presented ai the 10th Annual Conference of the Canadian Nuclear Society Ottawa, Ontario, 1989 June 4-7

Chalk River Nuclear Laboratories Laboratoires nucleates de Chalk River

Chalk River, Ontario KOJ 1J0

June 1989 juin ATOMIC ENERGY OF CANADA LIMITED

FUSION-BLANKET FABRICATION DEVELOPMENT AND IRRADIATION TESTING

I.J. Hastings**, A.J. Elliot+, J.M. Miller**'**, B.J.F. Palmer and R.A. Verrall

10th Annual Conference, Canadian Nuclear Society, Ottawa, Ontario

1989 June 4-7

Program Cofunded by Atomic Energy of Canada Limited Research Company and Canadian Fusion Fuels Technology Project

** Member, Canadian Nuclear Society

System Chemistry and Branch

Chemical Engineering Branch

Fuel Materials Branch Chalk River Nuclear Laboratories Chalk River, Ontario KOJ 1J0 1989 June

AECL-9999 CFFTP-G-8915 l-NKRCll' ATOMIQUE DU CANADA LIMITEE

COUVERTURE DE RÉACTEUR À FUSION - MISE AU POINT DE LA FABRICATION ET ESSAIS SOUS IRRADIATION*

I.J. Hastings, A.J. Elliot, J.M. Miller, B.J.F. Palmer et R.A. Verrall

10e Conférence annuelle de la Société nucléaire canadienne, Ottawa, Ontario du 4 au 7 juin 1989

RÉSUMÉ

Le programme de couverture de réacteur à fusion aux Laboratoires nucléaires de Chalk. River, financé conjointement par EACL et le PCTCT, vise la mise au point de la fabrication et les essais sous irradiation de la couverture. Les efforts canadiens de mise au point de couches fertiles solides sont concentrés sur le concept Spherepac. Les sphères, fabriquées par un procédé utilisant l'extrusion, sont solides, résistent à des cycles de chauffage rapides et présentent une taille de grain d'environ 1 jum. La résistance à la chaleur, la tenue mécanique et le rapport libération-production de tritium des sphères fabriquées par ce procédé font l'objet d'une caractérisation détaillée. Lors de l'irradiation en capsule à event CRITIC-I du Li20, on a obtenu une combustion massique totale du de 1 % et on a recueilli 2100 Ci (70 TBq) de tritium au cours d'une période d'irradiation de ?A mois. On participe également aux efforts sur le concept de solution aqueuse de sels de lithium (ALSB). On a réalisé la radiolyse en cellule gamma de solutions aqueuses de sels de lithium à l'appui du concept et retenu l'hydroxyde de lithium pour les prochains essais en réacteur.

Le Canada participe actuellement au projet BEATRIX-II, qui comprendra la mise à l'essai des sphères de zirconate de lithium de fabrication canadienne à l'installation FFTF, à Hanford, réalisé en collaboration avec le Japon et les É.-U.

Service des Matériaux pour combustibles

Laboratoires nucléaires de Chalk River Chalk River (Ontario) KOJ 1J0 1989 juin

AECL-9999 CFFTP-G-8915

*Programme financé conjointement par la Société de recherche d'Énergie atomique du Canada limitée et le Projet canadien sur la technologie des combustibles thermonucléaires. ATOMIC ENERGY OF CANADA LIMITED

FUSION-BLANKET FABRICATION DEVELOPMENT AND IRRADIATION TESTING*

I.J. Hastings, A.J. Elliot, J.M. Miller, B.J.F. Palmer and R.A. Verrall

10th Annual Conference, Canadian Nuclear Society, Ottawa, Ontario 1989 June 4-7

ABSTRACT

The fusion-blanket program at Chalk River Nuclear Laboratories concentrates on fabrication development and irradiation testing. Canadian efforts to develop solid-breeder blankets are focused on the sphere-pac concept. The spheres, fabricated by an extrusion-based process, are strong, survive rapid heating cycles, and have an average grain size of about one micrometre. In the recently completed CRIT1C-I vented-capsule irradiation of Li.,0, a total lithium hurnup of 1% was achieved, and 2100 curies+of tritium collected, over the 21-month irradiation. We are also participating in the development of an aqueous lithium salt blanket concept and has been chosen for further in-reactor studies. Canada is participating in the BEATRIX-II project, which will test Canadian-fabricated lithium zirconate spheres in FFTF, Hanford, with Japan and the US as the other partners. We are also members of the JAERI (Japan) Tritium Project, to improve blanket neutronics data.

*Program cofunded by Atomic Energy of Canada Limited Research Company and Canadian Fusion Fuels Technology Project + 2100 curies equal 70 TBq.

Fuel Materials Branch Chalk River Nuclear Laboratories Chalk River, Ontario KOJ 1J0 1989 June

AECL-9999 CFFTP-G-8915 1. INTRODUCTION

Canada is making significant contributions to the worldwide effort to develop fusion as an energy source. The lead agency in the Canadian program is Atomic Energy of Canada Limited (AECL) through the National Fusion Program, and major components are the Tokamak at Varennes, Quebec and the Canadian Fusion Fuels Technology Project (CFFTP) in Toronto, Ontario.

A program eofunded by AliCL and CFFTP, and focusing on fusion-blanket technol- ogy, began at Chalk River Nucleai Laboratories (CRNL) in 1983. The program was based on AECL's generic expertise in ceramics, irradiation testing and tritium technology, all key aspects in developing a breeder blanket. This paper updates the status of the program (1). The major components discussed are lithium-ceramic fabrication development, in-reactor testing, and work on an alternate concept, the aqueous lithium salt blanket (ALSB). Particular emphasis is given to the recently completed CRITIC-I experiment (CRITIC - Chalk River In-reactor Tritium Instrumented Capsule), now in the post- irradiation examination phase. Additionally, progress in international aspects of the program is outlined.

2. FABRICATION DEVELOPMENT

Canadian efforts to develop solid-breeder blankets focus on sphere-pac. AECL has substantial generic expertise through its experience with the concept in fission-fuel development (2). Sphere-pac blankets for fusion application require large numbers of tritium-breeding lithium-ceramic spheres. These are randomly packed into dense beds within the blanket modules incorporated into a fusion reactor. Because of their relatively small size, typically about 1 mm diameter, large numbers are required to fill a blanket bed. For example, a small-scale irradiation test on a 1-litre sphere bed requires in excess of 1 000 000, 1 mm diameter spheres. A full-size sphere-pac breeder blanket would be about 300 000 litres in volume, requiring more than 3xl0ll, 1 mm diameter ceramic spheres.

Work is under way at CRNL to develop technology for high-speed production of lithium aluminate and lithium zirconate spheres. Lithium aluminate is a well- established candidate for blanket application. Lithium zirconate has recently shown attractive low-temperature tritium-release properties during in-reactor rests (3). Emphasis in the fabrication program is on achieving the high production rates required to produce the large numbers of spheres necessary for a blanket module. Pilot-scale equipment capable of producing 1 mm spheres at the rate of 250 000 spheres per minute is currently being installed and commissioned. To date, more than 1 000 000 prototype lithium aluminate spheres have been produced at CRNL by the production process, based on extrusion. The spheres are mechanically strong, survive rapid heating cycles and have an average grain size of about 1 micrometre. In particular, the small surface-to-centre temperature differential (AT, less than 20°C), inherent in the sphere concept, provides one of its major advantages (4). Long-term irradiation testing (5) has identified cracking due to large ATs as a potential problem for blanket designs based on pellets or monolithic assemblies.

In addition to the development of a production process, good progress has been made in lithium-based powder synthesis. While lithium aluminate is available in industrial quantities via custom order, only small amounts of lithium zirconate have been produced worldwide. At CRNL, the precipitation technique has shown itself to be not feasible for quantity production. Current efforts focus on the solid-state reaction between zirconia and lithium .

The thermal, mechanical and tritium-release performance of spheres prepared by the extrusion-based process will be characterized in detail. An in-reactor CREATE (Chalk River Experiment to Assess Tritium Emission) test has just been completed on lithium aluminate spheres. In addition, high-flux irradiation tests are planned in the CRITIC facility of the NRU reactor at CRNL, and in the Fast Flux Test Facility (FFTF) at Han ford, Washington, as part of the BEATRIX-II program (International Breeder Exchange Matrix).

3. IRRADIATION TESTING

Two types of ceramic-blanket irradiation testing are performed at CRNL: CREATE tests, in which tritium-release information is obtained via post- irradiation annealing, and CRITIC tests, which permit on-line monitoring of tritium release as the irradiation progresses. Nine CREATE tests have been performed on lithium oxide and lithium aluminate, examining the effects of temperature, microstructure, sweep-gas composition and capsule material ( potential) on the amount and form of tritium release. Full details of these tests are given elsewhere (6-8).

The recently-completed CRITIC-I test in the NRU reactor at CRNL has produced a substantial amount of reactor-relevant data. Annular Li2<) pellets (30 mm I.D. x 40 mm 0-D.) were fabricated for CRITIC-I at Argonne National Laboratories under the BEATRIX-I program coordinated by the International Energy Agency (.TEA). The fabrication techniques and characterization results for CRITIC-I are described in detail elsewhere (9-12). The total weight of Li20 irradiated was 103 g, density 91.5% of theoretical, original isotopic content was 1.53 wt% ^Li and average grain size was 50 urn. The temperature of the ceramic was controlled by varying the composition of a He-Ar insulating gas layer (gap gas). Twelve thermocouples were located on the Li20 stack. The capsule provided approximately uniform ceramic temperatures (+50°C) in the range 370- 850°C. Full details of the capsule design are given elsewhere (9-12). A sweep gas flowed over the inner surface of the annular pellets to an analysis train which provided on-line monitoring of the tritium release. The first ionization chamber measured total tritium, while the second measured tritium in the reduced form only. Ethylene glycol bubblers were used to give an integrated measurement of the released tritium in both the oxidized and reduced forms (referred to as HTO and HT, respectively). A copper oxide bed was used to oxidize the HT to HTO. The bubbler solutions were changed periodically and tritium determined by liquid scintillation counting. A final ion chamber was included to measure the efficiency of the last bubbler and indicate releases, if any, to the NRU exhaust system. The sweep gas was monitored for moisture and oxygen content upstream and downstream from the capsule. The 30 m of sweep-gas line from the capsule to the analysis train was heated to 150°C to reduce tritium adsorption on the walls. The tritium that permeated into the gap-gas region of the capsule was oxidized and collected on a molecular sieve for periodic analysis. The irradiation test plan was divided into the following phases:

Phase 1 Conditioning the Li2O to 650°C - ceramic heated slowly to 650°C over 2 weeks.

Phase 2 Testing in the range 460-620°C - temperature step-change tests carried out, He

sweep gas for 6 weeks and He-0.01% H2 for 10 days.

Phase 3 Conditioning the LijO to 850°C - ceramic temperature increased to 850°C and held for 2 days.

Phase 4 Testing in the range 400-850°C - temperature step-change tests with He, He-

0.01? H2, He-0.1% H2, and He-1% H2 sweep gases.

Phase 5

Testing with H20 in the sweep gas, as described below.

Results from these phases, which include 95 temperature-transient tests, have been reported in detail (9-12). Figure 1 shows the results of a typical temperature transient, with tritium release shown as a function of tempera-

ture. This transient is a Phase 5 test, with moisture in the He-0.01% H2 sweep gas. Under these conditions most of the tritium was released as HTO. A

final burnup of 1% total lithium (58% ^Li) was achieved during the 21-month irradiation. The tritium generation rate was initially 5 Ci/day, dropping to .1.5 Ci/day after 461 full-power days of irradiation. Total tritium recovered was 2100 Ci.

Tritium release was found to be controlled by a surface-desorption release mechanism, and the influence of impurities (H20 or C02) in producing tritium as HTO during the initial irradiation phases was clear (10). The amount of tritium recovered in the reduced form (HT) increased from an initial value of approximately 50% with pure He sweep gas, to 99% with He-1% H2. The increas- ing H2 concentration in the sweep gas also reduced the time constants for tritium release (tritium residence time in the Li20). For the last 5 months of the irradiation, a small defect in the reactor insert resulted in -300 ppm moisture in the sweep gas. Post-irradiation examination showed the Li20 to be hard and durable, with only modest cracking. X-ray diffraction identified

Li20, with traces of LiOH. The post-irradiation tritium inventory was found to be less than 1 Ci, about 6 hours' tritium production. Figure 2 shows a scanning electron micrograph of a fracture-surface of the irradiated Li2O. Of note is the increased porosity (2)%) compared with 10% in the as-fabricated material (10), confirmed by mercury porosimetry. Figure 3 shows a higher- magnification micrograph; the bubble-like features are tentatively attributed to helium generated by the interaction of neutrons with lithium.

The CRITIC-I data show that tritium release is a complex process. Increasing

the H2 concentration in the sweep gas reduced the time constant for the tritium-release peaks, indicating a surface desorption-controlled release - 4 -

process. A desorption activation energy of 125 to 140 kJ/mol was obtained, in good agreement with previous results from post-irradiation studies. However, the initial part of the tritium-release peaks for temperature-increase tests were not consistent with simple first-order desorption kinetics. Kopasz et al. suggest (13,14) that the desorption activation energy changes with fractional coverage of the surface with tritium, and is in agreement with some of the CRITIC-1 data. This and othei. models are still being investigated.

Preparations are currently under way for CRTTIC-II. This will be an irradia- tion in the NRU reactor of CRNL-fabricated lithium zirconate spheres. The test will feature on-line tritium analysis, as for CRITIC-I. The scheduled insettion date is 1989 December.

4. AQUEOUS LITHIUM SALT BLANKET (ALSB)

An alternative to the solid ceramic breeder blanket is an aqueous solution containing approximately 2 mol drrr^ 6^ + . The ALSB is a simple, low- temperature, low-pressure concept (1), which is a candidate for the ITER/NET driver blanket. Solubility restrictions limit the choice to three lithium salts: the hydroxide, the nitrate and the sulphate. The ALSB has an advantage in that it draws on established technologies used in fission reactors and in tritium removal from aqueous solutions. However, it is recognized that radiolysis effects could form potentially-explosive mixtures of oxygen and , as well as degrading the counter ion. At CRNL, we are studying the radiolysis of lithium salt solutions, to select the most appropriate lithium salt for the ALSB concept and also to evaluate the use of an initial excess of hydrogen to suppress the radiolytic production of hydrogen and oxygen.

When degassed solutions containing either 3.6 mol dm~3 LiOH or 1 mol dm~^ M7SO4 were gamma irradiated (dose rate 1.4 Gy s~^) with >3.5xl0^ Gy, no oxygen was detected and only a small, steady-state concentration of hydrogen (1-2.5x10^0 molecules kg~l) was observed. In this case, trace impurities reacted with the hydroxyl radicals formed to leave an excess of hydrogen, which then suppressed further radiolysis of the water. On the other hand, when 5 mol dm~-^ LiN03 was gamma irradiated (< 5x10^ Gy) steady-state condi- tions were not observed: i.e., the formation of products was linearly dependent on dose over the range investigated. The product yields observed were: G(H2) of 0.026; G(02) of 0.43; and G(N02~) of 1.1. (G-values are defined as the number of species formed per 100 eV of energy absorbed.) The large nitrite ion yield indicates that the nitrate ion is being decomposed by direct action of the radiation in these solutions.

On the basis of the gamma radiolysis experiments (Linear Energy Transfer (LET) ~0.2 eV nm~l), 4.7 mol dm~3 LiOH solutions (natural abundance lithium) were chosen for in-reactor radiolysis experiments, where the radiation energy is essentially all derived from the recoil ions resulting from the reaction of thermal neutrons with 6Li+ (LET -24 eV nm"1 for the triton and -180 eV nra"1 for the alpha particle): n + 6Li > 3H + 4HG

Under these radiolysis conditions, the LET is comparable to, and the dose rate of 1(M Gy s~^ is about bZ of that expected in, an JTF.R/NF.T fusion reactor. Figure 4 summarizes the formation of liydt IH;CII ami oxygen in these solutions. The radiolysis does not approach steady state conditions, as both gases are formed linearly with dose, with (1-values of. 0.8 and 0.2, respectively. Furthermore, the gases are not foimed in stoichiome-tiic amounts and this is attributed to impurity effects in the small (2x10 -^ dm-*) quartz ampoules.

Although a dose of only 1.75x10^ (.'y is expected on a single pass through an TTER/NET fusion reactor, the production of gases at the rate shown in Figure 4 will require very large system operating pressures (-2-3 MPa) in order to keep these gases in solution, to prevent possible explosions (15). However, computer simulations at CRNL indicate that, for the radiolysis expected in a fusion reactor, the addition of modest amounts of hydrogen gas (~5xlO~^ mo! dm"-') to the solution should suppress the excess hydLogon and oxygen produc- tion (15). This effect is shown in Figure 5. Experiments are planned at CRNL to confirm these predictions.

5. INTERNATIONAL PROGRAMS

BEATRIX has been a successful example of international co-operation on solid- blanket technology. Canada, via the CRNL program, is a full partner in BEATRIX-! - the US, EEC and Japan are other participants. A recent matrix document showed 19 individual experiments within the matrix. Canada was the first to carry out testing under BEATRIX, on lithium aluminate from CEA (Sac lay, France); experiments have also been performed on material supplied by Japan and the US.

Canada is now participating in BEATRIX-II, advanced testing of candidate blanket materials in FFTF, Hanford, with Japan and the US as other partners. During 1988, the BEATRIX-II program, Phase I (16-18), progressed through design efforts into fabrication of its systems in order to meet a scheduled beginning of irradiation late in 1989. Funds were made available by Canada, Japan, and the US to pursue the design and fabrication program at Pacific Northwest Laboratory, Westinghouse Hanford Company and AECL. In addition, the Japan Atomic Energy Research Institute (JAERI) fabricated ceramic electrolysis cells for incorporation into the sweep-gas analysis lines. It appears that the final tasks before insertion into the reactor will he completed in time. Both the IEA and Hanford site-design-review meetings have approved the final design of the system.

The purpose of BEATRIX-II is to conduct an in-situ tritium recovery experiment in the high-energy-neutron environment of FFTF. During Phase I, two in-situ capsules containing LITO will be irradiated, with higher tritium production levels than in previous experiments. One capsule possesses the capability for temperature-change experiments, while the other exhibits a temperature gradient for evaluating the temperature stability in an engineering blanket configuration. Nonvented capsules are included for measurement of irradiation damage, thermal diffusivity, release kinetics, and beryllium compatibility. The tritium produced in the Phase 1 test and accumulated in the tritium removal system getter beds will he shipped to the Tritium System Test Assembly at. Los Alamos National Laboratory. The tritium will be processed into a form compatible with fusion fuelling, thus demonstrating the complete fusion-fuel cycle. The scope of Phase II (Cycle 12) irradiations, scheduled for late 1990, has just been established. CRNL- fabricated lithium zirconate spheres will be included in one of the vented capsules.

In addition to the BEATRIX program, we are also participating in the JAERI Tritium Project. Laboratories from Canada, US, Japan and Europe have provided lithium-based ceramics for iiradiation in the FNS and LOTUS high-energy (14 MeV) neutron facilities. On completion of irradiation, the ceramic will be measured for tritium content; all data will be pooled and reported, to improve blanket neutironies calculations.

6. CONCLUSIONS

(1) There has been substantial progress in the three main elements of the program,

- completion of the long-term, CRITIC-I in-reactor test that has provided significant reactor-relevant data on Li20, and preparation for CRITIC-II, with lithium zirconate spheres,

- good pjogress in lithium-based powder synthesis and the development of high-speed, sphere-fabrication technology, and

• selection of LiOH as the prime candidate for the aqueous lithium salt blanket concept following gamma-cell and in-reactor radiolysis testing.

(2) In international programs, lithium zirconate spheres fabricated at CRNL will be tested in Phase II of the BEATRIX-II irradiation in FFTF, Hanford. Additionally, the JAERI Tritium Project will result in improved blanket neutronics data.

7. ACKNOWLEDGEMENTS

The support of Drs D.P. Dautovich, K. Wong and P. Gierszewski (CFFTP) in the course of the program is acknowledged. D.S. MacDonald and D.H. Rose made significant contributions to operation and examination of CRITIC-I; M. Chenier performed the ALSB radiolysis measurements. The fabrication-development team includes J.D. Sullivan, J.D. Halliday, B. Clatworthy, L.E. Bahen and F. Gravelle. CRNL Site Support Services provided substantial effort in the CRTTIC-I project. 8. REFERENCES

(1) HASTINGS, I.J. and GIERSZEWSKI, P., "Fusion Fuel Blanket Technology", Proc. F.iigi nee ring Centennial Conference, Montreal, 1987 May 18-?.?., also available as Atomic Energy of Canada Limited, report AECL-9385 (1987).

(2) HASTINGS, I.J., CELLI, A., ONOFREI, M. and SWANSON, M.L., "Irradiation Performance of (Th,U)02 Fuel Designed for Advanced Applications", Proc. Third Annual CNS Conference, Toronto, Canada, 1982 June 8-9.

(3) JOHNSON, C.E., Argonne National Laboratory, private communication.

(4) SULLIVAN, J.D. and PALMER, B.J.F., "Fusion Breeder Sphere-Pac Blanket Design", Atomic Energy of Canada Limited, report AECL-9510 (1987).

(5) HOLLENBERG, G.W., "Swelling of Lithium Ceramics During Irradiation", Proc. First Int. Symposium on Fabrication and Properties of Lithium Ceramics, Pittsburgh, PA, 1987 April 26-30, A.Cer.S. Advances in Ceramics, Vol. 25, Eds. I.J. Hastings and G.W. Hollenberg, p. 183.

(6) HASTINGS, I.J., MILLER, J.M., VERRALL, R.A., BOKWA, S.R. and ROSE, D.H., "Irradiation of Lithium-Based Ceramics for Fusion Blanket Applications", Proc. Seventh Annual CNS Conference, Toronto, Canada, 1986 June 8-11, 7 122 U986).

(7) MILLER, J.M., BOKWA, S.R. and VERRALL, R.A., "Post-Irradiation Tritium Recovery from Ceramic Breeder Materials", Proc. Second Int. Conf. Fusion Reactor Materials (ICFRM-2), Chicago, 1986, J. Nucl. Mater. 141-143 294 (1986).

(8) MILLER, J.M., VERRALL, R.A., BOKWA, S.R. and ROSE, D.H., "The Effect of Material Characteristics on Tritium Release from Lithium Oxide", Proc. First Int. Symposium on Fabrication and Properties of Lithium Ceramics, Pittsburgh, PA, 1987 April 26-30, A.Cer.S. Advances in Ceramics, Vol. 25, Eds. I.J. Hastings and G.W. Hollenberg, p. 53.

(9) HASTINGS, I.J. et al., "Canadian Fusion Breeder Blanket Program", Proc. Second Int. Conference on Fusion Reactor Materials (ICFRM-2), Chicago, 1986, J. Nucl. Mater. 141-143 1044 (1986).

(10) VERRALL, R.A., MILLER, J.M., JOHNSON, C.E. and BOKWA, S.R., "CRITIC-I - Instrumented Lithium Oxide Irradiation: Part 2 - First Results", Proc. First Int. Symposium on Fabrication and Properties of Lithium Ceramics, Pittsburgh, PA, 1987 April 26-30, A.Cer.S. Advances in Ceramics, Vol. 25, Eds. I.J. Hastings and G.W. Hollenberg, p. 41.

(11) MILLER, J.M., VERRALL, R.A., MacDONALD, D.S. and BOKWA, S.R., "The CRITIC-I Irradiation of Li?0 - Tritium Release and Measurement", Proc. Third Topical Meeting on Tritium Technology in Fission, Fusion and Isotopic Applications, Toronto, Canada, 1988 May 1-6, Fusion Technology 14 649 (1988). - 8 -

(12) VERRALL, R.A., MILLER, J.M., HASTINGS, I.J. and MacDONALD, D.S., "CRITIC-I: Tritium Release from Large-Grained Li20", Proc. Second Int. Symposium on Fabrication and Properties of Lithium Ceramics, Indianapolis, IN, 1989 April 23-26, A.Cer.S. Advances in Ceramics, to be published.

(13) KOPASZ, J.P., TAM, S.W. and VERRALL, R.A., "Modelling Unusual Tritium Release Behaviour from Li20", Fusion Tech. ^5 1217, (1989).

(14) KOPASZ, J.P., FISCHER, A.K. and JOHNSON, C.E., "Desorption Activation Energies for Tritium Release from Ceramic Breeders", Proc. Second Int. Symposium on Fabrication and Properties of Lithium Ceramics, Indianapolis, IN, 1989 April 23-26, A.Cer.S. Advances in Ceramics, to be published.

(15) ELLIOT, A.J. and McCRACKEN, D.R., "Computer Modelling of the Radiolysis in an Aqueous Lithium Salt Blanket: Suppression of Radiolysis by Addition of Hydrogen", Fusion Eng. Design, to be published, also available as Atomic Energy of Canada Limited, report AECL-10013 and CFFTP-G-8914.

(16) HOLLENBERG, G.W., HASTINGS, I.J., MILLER, J.M., KURASAWA, T., UATANABE, H., BERK, S.E., BAKER, D.E., BAUER, R.E. and PUIGH, R.J., "A Fast- Neutron, In-Situ, Tritium-Recovery Experiment on Solid Breeder Materials", Proc. ANS Int. Conf. on Fusion Technology, Salt Lake City, UT, 1988 June 6-10.

(17) PUIGH, R.J., HOLLENBERG, G.U., KURASAUA, T., WATANABE, H., HASTINGS, I.J., MILLER, J.M., BERK, S.E., BAUER, R.E. and BAKER, D.E., "BEATRIX- II: In-Situ Recovery from a Fast-Neutron Irradiation of Solid Breeder Materials", Proc. 15th Symposium on Fusion Technology, Utrecht, The Netherlands, 1988 September 5-8.

(18) HOLLENBERG, G.W., "First. Annual Progress Report, BEATRIX-II Program, January 1988 to December 1988", Pacific Northwest Laboratory (US) report PNL-6818 (1989). - 9 -

20 ,- 800

585°C 600 I16 E \ 431°C LJ 400 JL & cc < LLJ 200 < _J /\ LLJ 01 8 o_ 2: 2: Total Tritium

l i 6 9 12 15 TIME (hours)

880CT17 TEST 91 He-0.01% H2 * MOISTURE IN SWEEP GAS

FIGURE 1. Tritium release as a function of time for a typical temperature transient in CRITIC-I. Most of the tritium in this case was rel- eased as HTO because of moisture in the He-0.01% H_ sweep gas. - Ill -

FIGURE 2. Sr.mninj; clcriron m i r ro;; raph ot i r r.-id i nL L-CI lithium oxide 1 rom CRITIC-1. XoU' ;iri,;|S ,,| pon-siLv.

FIGURE 3. Si-.-inninj; HiTiron mi rroj.i-.-iph of i n-;ul i .Hal lithium oxide troin CRITIC-I, sin.win;-. iV.iUiros t ^n t ;i t i vo 1 v i ik-nt i i" u-d as l'e 1 i inn buhb 1 cs . - 11 -

1 1 1 i ' ' ' ' /

- / •- YIELDS • HYDROGEN / ' 30 en o OXYGEN

to -

O 21 20 - /

O h- < 10 --

o-

i I i 12 3 4 DOSE/1024 eV kg'1

FIGURE h. Formation of Hydrogen and Oxygen as a function of dose. 50 m i E

O CALCULATED CONCENTRATION E 30 • HYDROGEN a "EFFECTIVE" OXYGEN

20 T:

10

4 -3 [H2] added/10" mol dm

I-7CURE 5. Suppression of excess Hydrogen and Oxygen production by addition of Hydrogen gas. ISSN 0067-0367 ISSN 0067-0367

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