CBBI-13 Workshop Program Wednesday, Nov. 30th, 2005 8:00 AM Registration (San Miguel Room) (closes at 12:00 PM) 8:30 AM Session A: Be Workshop Plenary Session I

(First Floor Ballroom) 12:00 PM Lunch 1:15 PM Welcoming Remarks (San Miguel Room) 1:30 PM Session B: Progress in the design of ceramic breeder blankets

for ITER-TBM and DEMO (Session Chair: Calderoni) B1: Current design status of the EU solid breeder Test Blanket

Module (Boccaccini) B2: Current status of design of solid breeder Test Blanket Module

of Japan (Enoeda) B3: Progress on US Helium-Cooled Ceramic Breeder (HCCB)

ITER TBM (Ying) B4: ITER TBM design progress and R&D plans in China* (Ying) 2:50 PM Break Session C: Out-of-pile testing facilities and mock-up tests for

ITER TBM (Session Chair: Enoeda) C1: Characterization of lithium orthosilicate pebbles for HELICA

and HEXCALIBER experiments (Alm) C2: Out-of-pile thermomechanical testing of breeder pebble beds

for HCPB TBM for ITER (Dell’Orco) C3: Investigation of fabrication technologies for Japanese ITER

Test Blanket Module (Hirose) C4: US thermomechanics R&D status and plan (Calderoni) 4:20 PM Discussion: Out-of-pile testing facilities and mock-up tests for

ITER TBM (Session Chair: Enoeda) (Adjourn 5:00 PM) Thursday, December 1st, 2005 8:30 AM Session D: Pebble bed thermomechanics (experiments and

modeling) (Session Chair: Boccaccini) D1: Thermal expansion behavior of a compressed Li2TiO3 pebble

bed (Suzuki) D2: Thermomechanical performance characterization of ceramic

pebble beds by Discrete Element Method modeling (An) D3: Thermomechanical behavior of LiSO4 and Li2TiO3 pebble

beds (Zaccari) D4: Measuring creep strain of individual Li2TiO3 spheroids

(Papp)

9:50 AM Break 10:00 AM Session E: Progress in ceramic breeder material development

and characterization (Session Chair: Dell’Orco) E1: Nanostructured ceramics blanket materials (Tiliks) E2: Thermal investigation of glassy lithium orthosilicate pebbles

(Knitter) E3: Experimental measurements of the effective thermal

conductivity and interface thermal conductance of a lithium titanate pebble bed (Abou-Sena)

11:00 AM Discussion: Pebble bed material development and thermomechanical properties (Session Chair: Dell’Orco)

12:00 PM Lunch 1:00 PM Session F: Irradiation testing, tritium release and tritium control

(Session Chair: Hirose) F1: Some considerations in the tritium control design of the solid

breeder blanket concepts (Boccaccini) F2: Results of the pebble-bed assembly irradiation* (Van der Laan

/ Hegeman) F3: Status of the high fluence irradiation of ceramic pebbles in the

HICU project* (Van der Laan / Hegeman) F4: In-pile behavior of lithium titanate in EXOTIC-9* (Van der

Laan / Hegeman) F5: Behavior of hydrogen isotopes irradiated in Li-containing

oxides (Luo) F6: Swamping effects on tritium permeation in solid breeder

blanket units (Guo) F6: Molecular simulation of radiation behavior of Li2O (Oda) 3:20 PM Break 3:30 PM Discussion: Irradiation testing, tritium release and tritium

control (Session Chair: Hirose) 4:00 PM Review of ICFRM Presentation 5:00 PM Reception / Dinner (Solstice Room and Terrace) Friday, December 2nd, 2005 8:30 AM Session G: Topical Discussions G1: Diagnostic needs and development and impact on TBM

design (EM and NT Modules)** (Ochiai) 10:00 AM Break G2: Materials database evaluation** (Sharafat) 11:30 AM Summary and Conclusions 12:00 PM Lunch 1:30 PM IEA Workshop (ends at 5:00 PM) *Abstract not available at print time **Topical discussions do not have abstracts

CBBI-13 List of Participants Ali ABOU-SENA UCLA ali@fusion.ucla.edu Zhiyong (John) AN UCLA an@fusion.ucla.edu Birgit ALM FZK birgit.alm@imf.fzk.de Lorenzo BOCCACCINI FZK boccaccini@irs.fzk.dePattrick CALDERONI UCLA patcal@fusion.ucla.edu Chang’an CHEN* CAEP chang_an.chen@163.netGiovanni DELL’ORCO University of Palermo giovanni.dellorco@brasimone.enea.itMikio ENOEDA JAERI enoedam@fusion.naka.jaeri.go.jp Kaiming FENG* SWIP fengkm@swip.ac.cn Wen GUO UCLA wen@fusion.ucla.edu Hans J.B. HEGEMAN JRC Petten hegeman@nrg-nl.comTakanori HIROSE JAERI hiroset@fusion.naka.jaeri.go.jpRegina KNITTER FZK regina.knitter@imf.fzk.de Deli LUO* CAEP luodeli2005@hotmail.comTianyong LUO Tokyo University luotianyong@flanker.q.t.u-tokyo.ac.jpKentaro OCHIAI JAERI ochiai@fnshp.tokai.jaeri.go.jp Takuji ODA Tokyo University oda@flanker.q.t.u-tokyo.ac.jpDaniel PAPP UCLA daniel@fusion.ucla.edu Shahram SHARAFAT UCLA shahrams@ucla.edu Tom SKETCHLEY UCLA tom@fusion.ucla.edu Satoshi SUZUKI JAERI suzukisa@naka.jaeri.go.jpJuris TILIKS University of Latvia jtiliks@lanet.lvJaap J.G. VAN DER LAAN JRC Petten vanderlaan@nrg-nl.comXisheng WANG* CAEP wxs755@sohu.com Alice YING UCLA ying@fusion.ucla.edu Nicola ZACCARI University of Pisa n.zaccari@ing.unipi.itPengcheng ZHANG* CAEP zpc113@sohu.com*Participants from China did not receive their visas in time to participate in the meeting

CURRENT DESIGN STATUS OF THE EU SOLID BREEDER TEST BLANKET MODULE

L.V. Boccaccini, R. Meyder, H. Neuberger

Association Euratom-Forschungszentrum Karlsruhe, Institut für Reaktorsicherheit, Postfach 3640, D-76021 Karlsruhe

Since about 10 years, EU is supporting a Solid Breeder Blanket Concept, the Helium Cooled Pebble Bed (HCPB), as candidate concept for the Demo and for the testing in ITER. During 2004-2005 the HCPB Test Blanket Module (TBM) has been deeply re-designed according with the revision of the corresponding Demo design. The result is a robust TBM box being able to withstand 8 MPa internal pressure in case of in-box LOCA, filled with 18 breeding units (BU); a breeding unit – that contain the breeding ceramics and Beryllium - has dimensions of about 200 mm in poloidal and toroidal direction and about 400 mm in radial direction. The paper presents the recent achievements in the design of this test component and in particular the thermo-hydraulics lay-out of the first wall and the new design of the breeder units. For the tests in ITER 4 different types of TBM are foreseen. The first, the so called Electro Magnetic (EM) module is used in the initial phase of ITER, without neutron flux. This allows checking the computer codes, used for predicting the electromagnetic forces. The second module, used from the D-D plasma phase of ITER is to test the codes for the neutron transport and interaction with the module and the Tritium generation rate. The third module is to check the knowledge of pebble bed behaviour and its interaction with the steel cooling plates. The fourth module finally is thought to be kind of a demonstrator. It should show the possibility of breeding sufficient Tritium for continuous power process, the high grade of heat being extracted and the stable operation under different loading conditions. The test configuration and strategy connected with these different modules will be discussed, as well.

B1

Current Status of Design of Solid Breeder Test Blanket Module of Japan

M. Enoeda, Y. Nomoto, S. Suzuki, D. Tsuru, K. Ezato, T. Hirose, H. Tanigawa, H. Nishi and M. Akiba

Blanket Development Group, Fusion Energy Development Division, Japan Atomic Energy Agency One of the main engineering performance goals of ITER is to test and validate design concepts of tritium breeding blankets relevant to a power-producing reactor. The tests foreseen on modules include the demonstration of a breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation. To accomplish these goals, the test blanket module (TBM) test program is being coordinated among parties and ITER international team under the frame of Test Blanket Working Group (TBWG). Among various blanket concepts, two DEMO-relevant TBM systems are proposed by Japan: a water cooled solid breeder (WCSB) blanket and a helium cooled solid breeder (HCSB), both with ferritic steel structure. Both concepts are based on the candidate DEMO designs in Japan. Both TBMs are planned to be delivered for testing in ITER from the first day of ITER operation. Recently, conceptual designs of both TBMs have been refurbished. The design of the TBM systems contains the design description of the test blanket module and its ancillary systems including interfaces with the ITER device together with essential analysis results on TBM performance and supporting R&D achievements and plan to deliver TBM on the first day of ITER operation. This paper presents the overviews of current status and important issues of the design of TBM’s. The structure of a “breeder out of tube (BOT)” concept is applied with Li2TiO3 breeder pebble beds, Be neutron multiplier pebble beds, F82H structure and water coolant of PWR conditions (280/325°C, 15.5 MPa) for WCSB and high pressure helium (300/500°C, 8 MPa) for HCSB. The configuration of the structure is designed to give the simulated performance for DEMO blanket (“act-alike”). Thermal and thermal-hydraulic design features show good heat removal performance of the TBM. Temperature limits of the materials are almost satisfied, and there is no excessive pressure drops. On the assumption that the test port is shared with the He cooled LiPb TBM, the installation concept of the TBM into the test port #18, is also developed. The ancillary systems, i.e., cooling system and tritium recovery system, have also been designed. By the performance analyses, nuclear performance was clarified to identify the basic design conditions on nuclear heating rate, tritium breeding ratio, induced activation and decay heat, incorporating the surrounding structure of the common frame. As the critical design issue, integrity of the module structure and capability of temperature control are evaluated to be sound enough. As the estimation of the functional performance, transient temperature evolution and tritium release behavior were analyzed and shown the estimation of the test performance of TBM. Also, safety related analyses were performed, in conservative manner, to have shown that the consequence had converged in the case of coolant ingress to the module, incorporating the effect of reaction heat generation by water-steam reaction. With respect to the development and delivery plan of TBM to ITER, supporting R&D showed sound progress on development of module fabrication technology, breeder and multiplier pebble fabrication technology and tritium recovery system. The R&D also included clarification of out-pile and in-pile blanket performance and neutronics performance and tritium production rate verification by 14 MeV neutron source. Further R&D plan on Engineering R&D using real scale blanket mockup was clarified toward the delivery of TBM to be operated from the first day of ITER operation.

B2

Progress on US Helium-Cooled Ceramic Breeder (HCCB) ITER TBM

A. Ying1, M. Abdou1, P. Calderoni1, Y. Kaoth2, R. Kurtz3, S. Reyes4, S. Sharafat1,

T. Tanaka5, S. Willms6, M. Youssef1, S. Zinkle2

1University of California, Los Angeles, Los Angeles, CA, USA

2Oak Ridge National Laboratory, Oak Ridge, TN, USA 3Pacific Northwest National Laboratory, Richland, WA, USA

4Lowance Livermore National Laboratory, Livermore, CA, USA 5Sandia National Laboratory, Albuquerque, NM, USA

6Los Alamos National Laboratory, Los Alamos, NM, USA

Abstract A planning and costing activity for the US TBM Program has been requested by the DOE for the reference TBM scenarios. For the HCCB, the reference scenario includes a submodule that has a size of 1/3 of one-half port and uses the EU (or Japan) first wall/box. The cost to be considered includes R&D, engineering design, mock-up/ prototype tests, fabrication, qualification, etc., as well as the cost of interface with ITER and other parties. To facilitate this costing activity, the TBM project is broken down into several key work subcategories: R&D; engineering; procurement and fabrication; project support; ancillary equipment; and assembly, testing, and installation. R&D is categorized according to safety and licensing requirements, first TBM design specifics, and enhanced confidence level in predictive capability. The main goal of R&D is to reduce risks, specifically those that impact device safety and availability and those that cause any unanticipated poor blanket performance, jeopardizing the objectives of the testing program. It appears that fabrication and qualification tests of the mock-up/prototype could drive the overall schedule, in which much attention should be given to the design-specific R&D of the first test article as compared to the R&D necessary to achieve experimental objectives for the later phases of the ITER TBMs. This paper reports the progress of this activity for the proposed HCCB ITER test blanket module with emphasis on schedule, deliverables, and critical path.

B3

Characterization of Lithium Orthosilicate Pebbles for HELICA and HEXCALIBER Experiments

B. Alm, R. Knitter

Forschungszentrum Karlsruhe, Institut für Materialforschung III

P.O. Box 3640, D-76021 Karlsruhe, Germany

ABSTRACT

Procurement, processing and characterization of lithium orthosilicate (Li4SiO4) (OSi)

pebbles were carried out for the out-of-pile experiments HELICA and HEXCALIBER at

ENEA in Brasimone. A quantity of 7 kg OSi pebbles with diameters of 0.2 - 0.4 mm was

produced, an amount of 6 kg was delivered to Brasimone in November 2004, 1 kg was

delivered to FZKa for quality control. Due to a rate of yield for this diameter range of only

about 40 wt%, 14 batches were produced, screened and mixed to gain 7 kg of the desired

material.

The chemical analysis revealed a slightly high excess of 3.3 wt% SiO2, the important

material properties like crush load, density, and the content of impurities, however, are well

within the range of the usually produced OSi reference material. During annealing at

970°C for 1 week, the decomposition of the high-temperature phase and the usual

coarsening of the microstructure took place without any significant changes of the

properties. From the results of quality control and characterization there is no indication

that the material will not behave well in the out-of-pile experiments HELICA and

HEXCALIBER.

C1

Out-of-Pile Thermo-Mechanical Testing of Breeder Pebble Beds for HCPB TBM for ITER

G. Dell’Orcoa, P. A. Di Maiob, R. Giammussob, A. Malavasia, L. Sansonea, A. Tincanic, G. Vellab

(a) ENEA Brasimone, 40032 Camugnano (Bo), Italy (b) DIN-Dipartimento di Ingegneria Nucleare, Università di Palermo, Viale delle Scienze, 90128,

Palermo, ITALY (c) DIENCA- Dipartimento di Energetica e Controllo Ambientale, Università di Bologna, Viale

Risorgimento, 2- 40136, Bologna, ITALY

The Helium Cooled Pebble Bed (HCPB) Blanket is one of the reference concept for the Test Blanket Module

(TBM) to be tested in ITER. In the HCPB TBM module, alternate staked beds of Lithiated ceramics and Beryllium

pebbles act respectively as Tritium breeder and neutron multiplier.

The thermo-mechanical behaviours of the pebble beds and their nuclear performances in terms of Tritium

production, are dependent from the reactor relevant conditions (heat flux and neutron wall load), the pebble sizes

and the breeder cell geometries (bed thickness, pebble packing factor, bed overall thermal conductivity).

ENEA, in the frame of the EU Fusion Technology Programme, has performed several experimental campaigns to

determine some of these behaviours by out-of-pile testing of small scale mock-ups.

The “Dipartimento di Ingegneria Nucleare” (DIN) of the Palermo University has adapted the thermo-mechanic

constitutive models (non-linear elasticity and plasticity models), available on commercial FEM code, for the

prediction of the thermal and mechanical performances of breeder pebble beds and for the comparison with the

experimental results of the ENEA tests.

More recently, among the EU Associations involved in the ceramic breeder qualification (ENEA-FZK-NRG), a

benchmark exercise has been launched aiming at selecting the pebble bed thermal mechanical constitutive models to

be implemented in a commercial FEM computer code.

The paper presents the main experimental results of two test campaigns on HELICA mock-up under testing at HE-

FUS 3 facility of ENEA Brasimone, the geometry of the mock-up, the adopted thermal and mechanical boundary

conditions and the test operating conditions.

HELICA mock-up represents a portion of a single breeder cell, made in martensitic steel, provided with a couple of

flat electrical heaters simulating the internal heat sources.

The experimental test campaigns on the breeder pebble bed behaviour are being carried out with the variation of the

Helium coolant and bed temperatures, at two different bed orientations (vertical and horizontal) and at fixed external

mechanical constraints both at steady state as well as at internal power step cyclic conditions.

The paper also presents the comparisons with the theoretical calculations, carried out by commercial FEM code, for

the benchmark exercise.

C2

Investigation of Fabrication Technologies for Japanese ITER Test Blanket Module

T. Hirose, Y. Nomoto, M. Enoeda and M. Akiba Japan Atomic Energy Research Institute, 801-1 Mukoyama, Naka, Ibaraki, 311-0193 Japan hiroset@fusion.naka.jaeri.go.jp

This paper presents recent results of R&D for fabrication process for Japanese ITER Test Blanket Module (TBM). Solid breeder blankets, which are cooled by pressurized light water, were selected as the primary candidate blanket of the fusion power demonstration plant in Japan. The Water Cooled Solid Breeder (WCSB) blanket is the primary candidate of Japanese TBM. The structural material for the TBM is reduced activation ferritic/martensitic steel, F82H. The gross weight of structural material is estimated to be 2.2t (1.9t for plates and 0.3t for tubes). The wall of the TBM has built-in cooling channels, which is fabricated by Hot Isostatic Pressing (HIP) method using square tubes and thin plates. These wall structures are connected by welding.

The square tubes for built-in cooling channels were successfully manufactured by cold rolling. The size is 11mm x 11mm, the wall thickness is 1.5mm and the total length is 3500mm. The tube is long enough to fabricate the first wall without any joint in the cooling channels. The surface roughness (Rz) is less than 1μm and it is smooth enough to HIP joint.

In the case of HIP bonding, nondestructive inspection is difficult because possible defects in the interface would include plane crack and/or film like obstacles with thickness of 1μm, which can hardly be detected by conventional method. For reliable blanket fabrication, it is necessary to establish the quantitative criteria on the pretreatment conditions of the HIP process. The degassing conditions and surface roughness were investigated as parameters of HIP conditions in this work. F82H joint was fabricated with HIP conditions of 1373K, 150MPa and 2h holding time. Normalizing at 1233K for 0.5h and tempering at 1023K were performed as post-HIP heat treatment. Although tensile property of the HIP joint was not sensitive to the surface roughness and degassing conditions, the impact property was significantly degraded by lack of degassing. The degradation in impact property was caused by the micro dispersed oxide on the bonding surface. The oxide was decomposed by thermal aging and the aging recovered the impact property. Therefore, to complete reliable joint, it is necessary to remove the surface oxide on the bonding surface.

C3

Status and plan of thermo-mechanics R&D for HCCB TBM in the US

P. Calderoni, A. Ying, M.A. Abdou

The thermo-mechanical performance of lithium-based ceramic materials is a critical issue for assessing the reliability of solid breeder blanket concepts over the lifetime of the component in a fusion power system. The investigation of the effect of the thermal cycling typical of a fusion energy power plant on the blanket components is complicated by the coupling of the thermal and structural behavior intrinsically related to the use of pebble beds as breeding material. For example, the formation of a void between the pebble bed and the structure due to creep relaxation would lead to a local deterioration of the interface heat transfer and the creation of a hot spot. The effect of thermal cycling is further aggravated by the degradation of materials mechanical properties due to neutron irradiation, so the final assessment of blanket component reliability in terms of its thermo-mechanical properties will be provided only by the results of the HCCB TBM during the DT phase of ITER operation.

However, thermo-mechanics out of pile tests are necessary to characterize the behavior of the ceramic breeder material under thermal and mechanical loads predicted to occur during the various phases of ITER operation and to establish a verified property database to be used for modeling purposes. This paper presents the current status of the experimental work on pebble beds thermo-mechanics in the US as well as the planned research within the Helium-Cooled Ceramic Breeder ITER Test Blanket Module program. The objective of the thermo-mechanics R&D for HCCB is the definition and experimental verification of temperature windows for the candidate ceramic materials within which the thermo-mechanical response of the pebble bed is acceptable, based on the predicted operational requirements in ITER. The analysis of the experimental results and the parallel development of predictive capabilities will be used as input for the design optimization of the HCCB TBM and for qualification and safety report of the test article.

C4

Thermal expansion behavior of a compressed Li2TiO3 pebble bed

H. Tanigawa, S. Suzuki, M. Enoeda and M. Akiba

Japan Atomic Energy Agency

In the blanket with ceramic breeders, it is planned that small pebbles of breeding materials are packed into containers and used as pebble beds. In the case of the pebble bed, thermal and mechanical conditions on the bed affect thermal and mechanical properties. Therefore, it is important to control and measure them at the same time. In the present paper, thermal expansion of a Li2TiO3 pebble bed was analyzed.

Our apparatus consists of a tensile test-apparatus, INSTRON, and a measurement chamber for thermal conductivity. Pebbles of Li2TiO3 with 2mm diameter were used. They were packed into a container made of alumina. The container was located in the measurement chamber. As reported in CBBI-12, it was difficult to check the height of the bed at high temperature because the measured deformation included thermal expansion of the load rods and the container. Instead of the bed, a column made of pure copper was installed and thermal expansion of the system was measured at different temperatures. Taking into account the estimated thermal expansion of the column, thermal expansion of the rods and the container could be analyzed. Based on the data, thermal expansion of the bed was measured under compression of 0.1MPa. The atmosphere was kept at 0.1MPa of He gas. Temperature of the pebble beds was regulated from room temperature to 973K.

Two types of beds were examined. One was the bed without pre-loading and its packing factor was 67.0%. The other was the bed compressed up to 10MPa before heating and packing factor was 67.9%. After the beds were heated up to 973K and cooled down to room temperature, their packing factor changed from 67.0 to 67.2% for the former bed and from 67.9 to 67.3% for the latter. Thermal expansion was the largest in the heating process for the pre-loaded bed and the smallest in the heating process for the non-loaded bed. In the cooling processes, thermal expansions of both beds were similar. When the pre-loaded bed was heated, stress and deformation caused by the pre-loading were relaxed and degree of compaction decreased. Hence, apparent thermal expansion increased. As regards the non-loaded bed, degree of compaction rose due to compression of 0.1MPa during the heating process. This led to a decrease in thermal expansion. These phenomena resulted in degree of compaction of the both beds in equilibrium after the heating. Therefore, thermal expansions were similar in the cooling processes for the both beds and packing factors were close after the experiments. These results suggest that residual stress in the bed cased by a compressive load can be annealed when the bed is heated without load.

D1

Thermo-mechanical Performance Characterization on Ceramic Pebble Beds by Discrete Element Method Modeling

Zhiyong An1, Alice Ying1, and Mohamed Abdou1

1Mechanical and Aerospace Engineering Dept., UCLA, Los Angeles, CA 90095-1597,

an@fusion.ucla.edu Many experimental and numerical approaches have been applied to quantify thermo-

mechanical characteristics of ceramic pebble bed assemblies. It is found that the macroscale

thermo-mechanical behaviors of a pebble bed material system are highly complex and can be

related to material properties, microstructure and size of particles, boundary conditions and

loading histories, etc. More works are needed to characterize the thermo-mechanical behaviors of

the pebble bed.

This paper presents the recent progresses on the development of a predictive capability for

ceramic breeder pebble bed primarily based on a discrete element method. Compared with

previous experimental methods, our numerical simulations show that discrete element method

has a more flexibility to assess the thermo-mechanics of ceramic breeder pebble bed.

In this paper, the numerical program using discrete element method (DEM) is mainly applied

to derive the effective constitutive relations of pebble bed. These constitutive relations are

dependent on pebble bed temperature and its boundary conditions and also are related to

microscale behaviors at the pebble contacts. Numerical results also include thermal creep

deformation, which becomes an important deformation mechanism at elevated temperatures.

Analysis will demonstrate the stress relaxation period of particulate material assemblies, which

has been proved that it can be much shorter than that of solid materials.

Corresponding Author: Zhiyong An Mechanical and Aerospace Engineering Department 44-139C ENGR IV, UCLA Los Angeles, CA 90095-1597, USA (310)794-4452 an@fusion.ucla.edu

D2

Thermo-Mechanical Behaviour of LiSO4 and Li2TiO3 Pebble Beds

Donato AQUARO1), Nicola ZACCARI1) Dipartimento di Ingegneria Meccanica Nucleare e della Produzione

University of Pisa Via Diotisalvi n.2 - 56100 Pisa (Italy)

(1)Dipartimento di Ingegneria Meccanica Nucleare e della Produzione University of Pisa Via Diotisalvi n.2 - 56100 Pisa (Italy)

This paper deals with a research activity on the breeding blankets of nuclear fusion reactors. In

particular, an experimental set up for the determination of the granular material conductivity in

presence of an interstitial pressurized gas is described. This research is performed in the frame of

the studies connected with the ITER plant. The ITER operation phase foresees to test Breeding

Blanket Modules (Test Blanket Module, TBM). One of the possible configurations of TBM, that

will be tested, has the neutron multiplier and the breeder made up of pebble bed. The knowledge of

the effective conductivity of the pebble bed versus the temperature and the bed deformation is of

fundamental importance for designing and optimizing these types of TBMs. The experimental set

up uses the so called ‘guarded hot plate’ method for the conductivity determination. The tests were

performed with a simultaneous compression of the bed (uniaxial oedometric deformation) which

permitted to obtain the effective bed conductivity versus the axial deformation at different values

of temperature. The effective conductivity of LiSiO4 and Li2TiO3 pebble bed were determined

considering several packing factors and several pressures of interstitial gas.

The distribution of the pebble bed packing factor was measured by means an ad hoc built

instrumentation based on the gamma ray backscattering. These measurements were performed off

line before and after the oedometric tests.

Moreover, a theoretical model and a discrete FEM model to simulate the thermal-mechanical

behaviour of the ceramic pebble bed are being developed by the authors. The results of the

theoretical models have been compared with the experimental ones, obtaining a good agreement in

terms of bed conductivity and stiffness.

D3

Measuring Creep Strain of Individual Li2TiO3 Spheroids

Daniel Papp, Pattrick Calderoni, Alice Ying1

1 University of California, Los Angeles, Los Angeles CA Predictive capabilities for the thermo-mechanical behavior of ceramic breeder pebble beds

are based on different approaches, one of which is the Discrete Element Method. This method requires characterization of the individual interparticle contact behavior, such as displacement as a function of contact force, temperature and time. As local contact stresses are higher than the ultimate compressive stress of the material, stress-dependent effects like creep and sintering between the particles cannot be predicted by data available on bulk properties. The later effects are important in proposed solid breeder pebble beds for fusion reactor blankets in the projected operational temperature range between 500°C and 900°C. Although basic mechanical properties of bulk Li2TiO3 material are known [1], experimental data on creep behavior have never been reported in the literature to the best of our knowledge. Extensive experimental work has been carried on to characterize the mechanical behavior of Li2TiO3 pebble bed assemblies as a whole [2], but the results can not be used to derive individual pebbles contact properties.

Research is ongoing to provide experimental data on creep strain of pairs of contacting Li2TiO3 pebbles of spheroidal shape, roughly 2mm in diameter. The goal of the experiment is to measure creep deformation under variable compressive forces between 3 and 15 N at temperatures up to 800°C. The equivalent compressive load on the pebbles cross sectional area is between 1 and 5 MPa. The deformation is measured with Linear Velocity-Displacement Transducers with 1 mm range and sub-micron precision. Preliminary measurements were made at room temperature before the installation of the heating equipment to determine the force-displacement characteristics of the pebbles. This behavior can be characterized as following a power-law dependence d = AFn. The preliminary results showed a linear behavior, with a scattering in the exponent between 0.7 and 1.2 for different pebbles attributed to their irregular shape. This effect is negligible at high temperatures where the creep deformations are expected to be one order of magnitude higher. Creep deformation data for a larger sample of the same material are being collected by measuring the displacement as a function of temperature, compressive force and time. Results obtained will be compared to creep behavior of bulk materials by means of Finite Element Analysis, with material properties based on data for other Li breeder ceramics [3].

1: P. Gierszewski, Fusion Engineering and Design 39-40 (1998) 739-743 2: L. Bühler, J, Riemann, Journal of Nuclear Materials 307-311 (2002) 807-810 3: ITER Solid Breeder Blanket Materials Database ANL/FPP/TM-263, Argonne National Laboratory, 1993

Corresponding Author: Daniel Papp UCLA – Department of Mechanical and Aerospace Engineering 44-114 Eng IV, 420 Westwood Plaza Los Angeles CA 90095-1597 Tel.: (310) 794 4452, Fax (310) 825 1715 Daniel@fusion.ucla.edu

D4

Nanostructured ceramic blanket materials

J. Tīliks, G. Ķizāne, A. Vītiņš, B. Leščinskis

Laboratory of Solid State Radiation Chemistry, Faculty of Chemistry, University of Latvia, Kronvalda bulvāris 4, LV-1010 Riga, Latvia

Lithium-containing ceramic materials is an important element of a blanket of a fusion

reactor. The main requirements for these materials is a high radiation and thermal stability, a low tritium retention under the operating conditions of the blanket – the temperature about 950 K, the fast neutron radiation about 1019 n⋅m-2⋅s-1 in an intense magnetic field 7-10 T. The efficiency of tritium release is inversely proportional to the grain size of the ceramics. The Li4SiO4 and Li2TiO3 ceramics synthesized by the present technology have the grain size 103-104 nm, which may increase under the conditions of high-temperature radiolysis.

The Li4SiO4 and Li2TiO3 ceramics obtained form nanosize powders (the particle size 10-50 nm, the surface 20-25 m2⋅g-1) were investigated for this study. The ultradisperse powders were synthesized from Li2CO3 and Si or TiO2 in argon atmosphere under the conditions of plasma chemical synthesis. The pressed pellets (∅10 mm) were sintered into ceramics at gradual heating to 950 K for 1 h. The ceramics obtained has the average grain size about 100 nm, the mechanical strength 1.5-2.5 kg⋅mm-2, a high radiation stability (the degree of decomposition was 0.03% at 100 MGy; for the standard ceramics that was 0.5%). The parameters of the tritium release from the nanostructured Li4SiO4 and Li2TiO3 ceramic pellets improved considerably in comparison with the standard ceramics: the temperature of tritium release decreased by 100 K, the activation energy of tritium release decreased from 11 kJ⋅mol-1 to 6 kJ⋅mol-1, the constant of the volume diffusion of tritium increased, the tritium retention decreased to 0.6% at 920 K, there was no magnetic field effect on the tritium release. Development of synthesis of the nanostructured ceramics in the form of pebbles is envisaged for further studies.

Corresponding Author: Juris Tiliks Laboratory of Solid State Radiation Chemistry, Faculty of Chemistry, University of Latvia, Kronvalda bulv. 4, LV-1010 Riga, Latvia Tel.+371-7033884 Fax +371-7033884 E-Mail Juris.Tiliks@lu.lv

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Thermal Investigations of Glassy Lithium Orthosilicate Pebbles

R. Knitter, B. Alm, C. Odemer

Forschungszentrum Karlsruhe, Institut für Materialforschung III

P.O. Box 3640, D-76021 Karlsruhe, Germany

ABSTRACT

Lithium orthosilicate pebbles have been developed at the Karlsruhe Research Center

in close cooperation with Schott AG, Mainz. In the fabrication process by melt-

spraying, a broad distribution of pebble sizes is obtained, but only pebbles with

diameters of 250 – 630 µm are selected for the use as breeder material in the HCPB.

Three different microstructures can be observed in the pebbles. While the

predominant microstructure is a dendritic solidification structure, some medium-sized

pebbles exhibit a microstructure that is supposed to be caused by homogeneous

nucleation. Most of the very small pebbles, however, are transparent and optically

amorphous and hardly have any cracks and pores.

With the long-term goal to optimize the mechanical properties of the pebbles, thermal

treatments of glassy pebbles has been carried out to induce a careful crystallization

with a minimum of cracks. The crystallization was studied by thermal analysis, and

the development of the microstructure and the phase composition was observed by

SEM and x-ray diffraction, respectively.

While samples treated at or quenched from temperatures ≥ 800°C consisted of the

two expected phases, ortho- and metasilicate, in samples quenched from or treated

at temperatures ≤ 600°C besides orthosilicate another phase was detected that has

not been observed before in the initial or conditioned material. The status-quo of the

experiments and the results gained so far will be presented.

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Experimental Measurements of the Effective Thermal Conductivity and Interface Thermal Conductance of a Lithium Titanate Pebble Bed

Ali Abou-Sena, Alice Ying and Mohamed Abdou Mechanical and Aerospace Engineering Department, University of California,

Los Angeles, CA 90095-1597

Corresponding author: Ali Abou-Sena Phone: 1-310-825-8824, Fax: 1-310-825-2599 Email: ali@fusion.ucla.edu

The use of lithium ceramics as tritium breeders in pebble bed form is a promising concept

for fusion blankets, and worldwide efforts have been dedicated to its R&D. The effective

thermal conductivity keff and interface thermal conductance h of the lithium ceramics

pebble beds have a significant impact on the thermal performance of the breeding fusion

blankets. There is a significant shortage of published data on h of the lithium ceramics

pebble beds. Also, the available data on keff is not sufficient due to the discrepancy among

the published values of keff and the small number of measured values. Therefore, more

data on both h and keff is still required for the Li2TiO3 pebble bed. The objective of this

study is to measure the interface thermal conductance and effective thermal conductivity

of a Li2TiO3 pebble bed as a function of the bed’s temperature. The pebble bed is single

size (1.7-2.0mm diameter pebbles) with a packing fraction of 61%. Helium at

atmospheric pressure was used as a cover gas. An experimental apparatus was designed

and built, based on the principles of the steady state and axial heat flow techniques, in

order to conduct the required measurements. The obtained results showed that the

effective thermal conductivity decreased from 1.40 to 0.94W/m.K with the increase of

the average bed temperature from 50 to 500°C. In addition the interface thermal

conductance increased from 2091 to 3600W/m2.K with the increase of the interface

temperature from 100 to 370°C. The measured values of h and keff will help to create a

complete database of these properties, which are needed for the design and analysis of

fusion blankets.

Keywords: effective conductivity, interface conductance, pebble bed, fusion.

E3

SOME CONSIDERATIONS IN THE TRITIUM CONTROL DESIGN OF

THE SOLID BREEDER BLANKET CONCEPTS

L.V. Boccaccini, N. Bekris, R. Meyder

Association Euratom-Forschungszentrum Karlsruhe, Postfach 3640, D-76021 Karlsruhe

The Tritium control is one of the most important issues in the design of a fusion blanket:

under this term it is intended not only all the processes and systems connected to the recovery of the Tritium produced in the breeding material, but also all the features required to minimise the permeation of the Tritium into the main coolant or to reduce the Tritium inventory in the materials for the implication in the safety due to the potential release in the environment.

In the Solid Breeder Blanket the tritium is produced by neutron reaction in a lithiated ceramics (Li4SiO4, Li2TiO3, Li2O, etc.). The breeding materials, generally in form of pebbles (diameter <1 mm), is purged by a low pressure flow of Helium that extracts the tritium from the pebble beds. Usually an addition of H2 to the purge flow contributes to facilitate the extraction of T from the pebbles. The outlet gas mixture (mainly He + HT +H2) is processed outside the reactor to recover the fuel for the thermonuclear reaction.

If the first objective of the design is the T production to assure the self-sufficiently of the reactor, the second one is to minimize the quantity of T that contaminates the main coolant flow. Sources of Tritium are the permeation inside the blanket module from the purge loop and the impingement of Tritium ions in the FW and hence into the cooling channels. A Coolant Purification System is used to process continuously a by-pass stream of the Coolant Loops keeping the level of T under a design value that is considered sufficient low to keep the potential losses of T in the steam generator below a safety acceptable level. Furthermore, the overall inventory of the T in the blanket materials should be minimised to avoid the risk of T release during accident.

The typical lay-out parameters of the T control design proposed for the HCPB blanket concept will be presented in the paper and discussed.

F1

Behavior of Hydrogen Isotopes Irradiated in Li-containing oxides

Tianyong Luo1，Takuji Oda1，Yasuhisa Oya2 and Satoru Tanaka1

1Department of Quantum Engineering and Systems Science,School of Engineering,

The University of Tokyo: Tokyo, Japan

2Radioisotope Center: The University of Tokyo，Tokyo, Japan

In order to make the tritium recovery process reliable, behavior of hydrogen isotopes in the blanket breeding materials should be clarified. For this purpose, many studies have been devoted, and the outline of tritium behavior, such as tritium release rates and diffusion coefficients, were obtained for each candidate breeders. However, to establish more secure and efficient fuel cycle, more detailed understandings in an atomic scale are needed on behavior of tritium and radiation defects and their interaction. From this viewpoint, we have conducted IR absorption analysis during deuterium ion (D2

+) irradiation, and thermal desorption spectroscopy (TDS) after the irradiation for Li2O, LiAlO2 and Li2TiO3.

In the IR spectra obtained during 3 keV D2+ irradiation for LiAlO2 and Li2TiO3, a

broad peak was observed corresponding to overlapped peaks from the stretching vibrations of O-Ds. Its shape was changed as the ion fluence increased. The LiOD phase, which is a dominant peak in Li2O, was not clearly detected in the ternary Li-containing oxides even over the fluence of 5×1021 D2

+•m-2. In the TDS experiment after the ion irradiation, both of D2O and D2 were desorbed from LiAlO2. The ratio of D2/D2O was dependent on the ion flux. It was also observed that the IR peak was decreased unevenly as the D2 and D2O being desorbed during the TDS experiment. These facts indicated that the deuterium near the surface holds multiple chemical forms and different manners of desorption due to the radiation effect and the structural features of the Li-containing oxides. In the presentation, these results will be discussed in detail, and a model to describe these behaviors of irradiated deuterium in Li-containing oxides will be proposed. Corresponding Author: Tianyong Luo Department of Quantum Engineering and Systems Science,

School of Engineering, The University of Tokyo 7-3-1 Hongo, Bunkyo-ku, 113-8656, Tokyo, Japan Tel: +81-3-5841-6970, Fax: +81-3-3818-3455 luotianyong@flanker.q.t.u-tokyo.ac.jp

F5

Swamping Effects on Tritium Permeation in Solid

Breeder Blanket Units

Wen Guo, Alice Ying, Ming-Jiu Ni, Mohamed A. Abdou Fusion Science and Technology Center, University of California, Los Angeles

420 Westwood Plaza, Los Angeles, CA 90095-1597, USA

E-mail wen@fusion.ucla.edu

Abstract In the helium cooled pebble-bed blanket, it is very important to limit tritium permeation from breeding zones to the coolant. In the case of no hydrogen addition to the purge stream, the probable tritium species are T2 and T2O, while T2O is assumed not to permeate through the coolant. The tritium permeation rate is (1) tDKPJ TT /2/1

2=

If some amount of hydrogen is added to the tritium, isotope effects should be considered. Tritium permeation rates from a given quantity of T2 molecules will be reduced by the addition of hydrogen to a point that the tritium permeation rates approach an inverse 0.5 power dependant on the hydrogen partial pressures [1]:

2/122/1/ HHTT PtkDKPJ = (2)

where, DK is the permeability and k is the equilibrium constant. A similar equation is given from [2], if 100 wppm hydrogen is added to helium purge gas, the tritium permeation flux is:

2/)8/3( HHTT PPtDKJ = (3)

where, PHT and PH2 are the partial pressures of HT and H2 on the breeder side of the coolant tubes. It can be seen from this equation that as the partial pressure of hydrogen is increased, the allowable partial pressure of HT is increased for a fixed tritium permeation flux. A FEMLAB-based model including convection-diffusion of the purge stream, diffusion and convection of tritium and the heat transfer is used to simulate tritium permeation from breeding zones to the coolant in the helium cooled pebble-bed blanket. Isotopic swamping effect due to additional hydrogen in the purge gas is evaluated and compared to no hydrogen case. Results show that the presence of hydrogen in a given amount of tritium will highly decrease the tritium permeation rate (6-12 times less) and be easier to satisfy the requirement that tritium leakage rate is less than 5 % of production rate Reference [1] Tritium Permeation through Steam Generator Materials [2] Tritium Percolation, Convection, and Permeation in Fusion Solid-Breeder Blankets

F6

Molecular simulation of radiation behavior of Li2O

Takuji Oda1, Yoshihisa Oya2, and Satoru Tanaka1

1Department of Quantum Engineering and Systems Science: The University of Tokyo,

Tokyo, Japan, 2Radioisotope Center: The University of Tokyo, Tokyo, Japan

Radiation behavior of a breeding blanket affects its performance in a fusion reactor, such

as mechanical strength, compatibility with structural materials, thermal conductivity, and tritium

release rate. The radiation effects will be amplified as a reactor is operated, thus sufficient

understandings are needed. In the case of Li-containing oxides under the fusion reactor condition,

key phenomena induced by the radiation are (i) easy displacement of Li compared with other

constituents, (ii) deficiency of Li by tritium-breeding reaction, (iii) formation of Li colloids, (iv)

formation of charged defects (F centers) which could interact with tritium meaningfully, and (v)

sintering via these fundamental processes. In the present study, we focused on (i) and (iv), and

conducted cascade simulation by molecular dynamics (MD) and analyzed interaction of hydrogen

isotopes with F centers using DFT calculation.

As a first step of the cascade simulation, a Buckingham-type potential model was created

empirically using GULP code, combined with the genetic algorithm. The created model had better

agreement with the experimental values than some reported models, on melting point, thermal

expansion behavior and mechanical property. Subsequently, a series of cascade simulation was

conducted under NEV ensemble at 0 K as a function of PKA energy and direction. Strong

dependences of threshold displacement energy on PKA directions, and easier formation of Li

vacancy than O vacancy were confirmed. In DFT calculation, stabilities of hydrogen interacting with

F centers and the UV/visible absorption spectra during the interaction were acquired. The results

fairly accorded with reported experimental results. In the presentation, we will discuss those

simulation results in detail, and review atomic scale information on the radiation behavior of Li2O.

Corresponding Author: Takuji Oda

Department of Quantum Engineering and Systems Science,

The University of Tokyo

7-3-1, Hongo, Bunkyo-ku, 113-8656 Tokyo, Japan

Tel.: +81-3-5841-6970, Fax: +81-3-3818-3455

E-mail Address: oda@flanker.q.t.u-tokyo.ac.jp

F7