JAERI-Conf 99-009
JP9950621
PROCEEDINGS OF THE THIRD JAERI-KAERI JOINT SEMINAR ON THE POST IRRADIATION EXAMINATION TECHNOLOGY
MARCH 25-26,1999, JAERI OARAl, JAPAN
September 1999
Japan Atomic Energy Research Institute WffiftiS (=f319-1195
(T319-H95
This report is issued irregularly. Inquiries about availability of the reports should be addressed to Research Information Division, Department of Intellectual Resources, Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan.
©Japan Atomic Energy Research Institute, 1999 JAERI-Conf 99-009
Proceedings of The Third JAERI-KAERI Joint Seminar on Post Irradiation Examination Technology March 25-26, 1999, JAERI Oarai, Japan
Department of JMTR
Oarai Research Establishment Japan Atomic Energy Research Institute Oarai-machi, Higashiibaraki-gun, Ibaraki-ken
(Received August 4, 1999)
Between the Department of JMTR of the Japan Atomic Energy Research Institute (JAERI) and the Nuclear Fuel Cycle Research Group of the Korea Atomic Energy Research Institute (KAERI), it has been periodically carried out the collaboration on technical information exchange by specialists and scientists, under the Arrangement of the Implementation of Cooperative Research Program in the Field of Peaceful Uses of Nuclear between JAERI and KAERI. And JAERI-KAERI joint seminar has been held every three years. The 1st and 2nd JAERI-KAERI Joint Seminars were held in November 1992 at JAERI and in September 1995 at KAERI, respectively. The 3rd JAERI-KAERI Joint Seminar was held on 25 and 26 March, 1999 at the Oarai Research Establishment of JAERI. In this seminar, total participants of 84 were joined from JAERI, KAERI, Hanyang University, Japan Nuclear Cycle Development Institute, Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd., Nuclear Development Corporation and others. Contributed presentations were 33 in three sessions; Current status and future perspectives on PIE (8 presentations), PIE techniques (11 presentations) and Evaluation of PIE data (14 presentations). Re-assembling technique for JOYO fuel, Nd-YAG laser welding technique, grain boundary analysis using FEG-TEM, lift time estimation of PWR Rod Cluster Control Assembly (RCCA) rodlet and failure analysis of Korea Nuclear Power Plant (KNP) fuel have been widely noticed as topic items on PIE. JAERI-Conf 99-009
And some comments from PIE user, were pointed out that the nano-PIE technique, the flexibility to ad-hoc demands on testing space or utilization, and the international collaboration were very important for the next generation's PIE
Keywords; PIE, Hot Laboratory, Re-assembling, JMTR, JOYO, HANARO, PWR, RCCA, Nd-YAG Laser, Failure Analysis, International Collaboration JAERI-Conf 99-009
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Foreword
It has been periodically carried out to bring together specialists and scientists in the field of the PIE activities and to strengthen the research cooperation between Japan Atomic Energy Research Institute (JAERI) and Korea Atomic Energy Research Institute (KAERI) through the mutual exchange of technical information on several kinds of items of Nuclear safety and Other related fields, under the Arrangement of the Implementation of Cooperative Research Program in the Field of Peaceful Uses of Nuclear between JAERI and KAERI since 1985. Between the Department of JMTR of the JAERI and the Nuclear Fuel Cycle Research Group of the KAERI, the 1st and 2nd JAERI-KAERI Joint Seminars on the PIE technology were organized by JAERI in November 1992 and KAERI in September 1995, respectively to summarize the results of mutual information exchange on the PIE activity. The 3rd JAERI-KAERI Joint Seminar was held in the HTTR main conference room at the Oarai Research Establishment of JAERI from March 25th to 26th, 1999 under the auspices of the Oarai Research Establishment of JAERI, the 2nd Seminar and three years later for summarizing the results of mutual information exchange on the PIE activity. The PIE-related persons including PIE users in and out the KAERI and JAERI took part in this seminar who belong to Hanyang University, Japan Nuclear Cycle Development Institute (JNC), Oarai Branch of Institute for Materials Research (IMR) of Tohoku University, Nippon Nuclear Fuel Development Co., Ltd. (NFD), Nuclear Development Corporation (NDC) and others. They discussed on recent topics on PIE and specially, Japanese PIE activities were briefly reviewed in this seminar by the result of participation of the JNC Oarai as a user on JMTR of JAERI. Contributed presentations of 33 were carried out in three sessions; Current status and future perspectives on PIE (8 presentations), PIE techniques (11 presentations) and Evaluation of PIE data (14 presentations) with total number of participants of 84 in two days. At the seminar, it was confirmed that key issues were to continue the mutual information change and the international collaboration and furthermore to grasp the perspectives of next generation's PIE. All the participants made an agreement to meet again in the next seminar three years later in Korea when the world-cup game on soccer will be held under the joint auspices of Japan and Korea.
JAERI-Conf 99-009
Contents
Opening Address Toshio Fujishiro 1 Key-Soon Lee 2
Session 1: Current Status and Future Perspectives on PIE 3
1.1 Over View of Nuclear Fuel Cycle Examination Facility at KAERI 5 Key-Soon Lee, Eun-Ga Kim, Kih-Soo Joe, Kil-Jeong Kim, Ki-Hong Kim and Duk-Ki Min (KAERI) 1.2 Activities on PIE of Nuclear Power Plant Fuels in KAERI 13 Eun-Ka Kim, Yong-Bum Chun, Gil-Sung You, Dae-Seo Koo, Duck-Kee Min, Key-Soon Lee and Seung-Gy Ro (KAERI) 1.3 Present Status of PIEs in the Department of Hot Laboratories 20 Tsuneo Kodaira, Tomohide Sukegawa, Hidetoshi Amano, Fumio Kanaizuka and Kiyomi Sonobe (JAERI) 1.4 Current Status and Future Prospects of JMTR Hot Laboratory 32 Osamu Baba, Norikazu Ooka and Taiji Hoshiya (JAERI) 1.5 Over View of Post-irradiation Examination Facilities for Fuels and Materials Development of Fast Reactor 44 Masahiko Itoh (JNC) 1.6(a) Activities of Oarai Branch IMR of Tohoku University as an Open Facility for Utilizing JMTR 53 Minoru Narui, Tsutomu Sagawa and Tatsuo Shikama (Tohoku Univ.) 1.6(b) Ventilation System of Actinides Handling Facility in Oarai-Branch of Tohoku University 66 Yoshimitsu Suzuki, Makoto Watanabe, Mituo Hara, Tatsuo Shikama, Hideo Kayano and Toshiaki Mitsugashira (Tohoku Univ.) 1.7 PIE Activities in NFD Hot Laboratory 75 Norikatsu Yokota, Keizo Ogata and Noriyuki Sakaguchi (NFD) 1.8 Current Status of NDC Fuel Hot Laboratory 89 Youichirou Yamaguchi, Takanori Matsuoka, Satoshi Shiraishi and Mitsuteru Sugano (NDC)
vii JAERI-Conf 99-009
Session 2: PIE Techniques 101
2.1 Development and Application of PIE Apparatuses for High-burnup LWR Fuels 103 Katsuya Harada, Naoaki Mita, Yasuharu Nishino and Hidetoshi Amano (JAERI) 2.2 A Technique of Melting Temperature Measurement and its Application for Irradiated High-burnup MOX Fuels 112 Takashi Namekawa and Takashi Hirosawa (JNC) 2.3 Development of PIE Techniques for Irradiated LWR Pressure Vessel Steels 119 Masahiro Nishi, Minoru Kizaki and Tomohide Sukegawa (JAERI) 2.4 High Resolution Grain Boundary Analysis of Neutron Irradiated Stainless Steel Using FEG-TEM 131 Mitsuhiro Kodama, Yoshihide Ishiyama and Norikatsu Yokota (NFD) 2.5 The Development of Crack Measurement System Using the Direct Current Potential Drop Method for Use in the Hot Cell —- - - 140 Do-Sik Kim, Sang-Bok Ahn, Key-Soon Lee, Yong-Suk Kim and Sang-Chul Kwon (KAERI) 2.6 Development of a Remote Controlled Small Punch Testing Machine for Nuclear Fusion Research 151 Masao Ohmi, Junichi Saito, Toshimitsu Ishii, Taiji Hoshiya and Shiro Jitsukawa (JAERI) 2.7 Newly Developed Non-destructive Testing Method for Evaluation of Irradiation Brittleness of Structural Materials Using Ultrasonic 163 Toshimitsu Ishii, Norikazu Ooka, Yoshiaki Kato, Junichi Saito, Taiji Hoshiya (JAERI), Saburo Shibata (IHI) and Hideo Kobayashi (Tokyo Institute of Technology) 2.8 Reassembling Technique for Irradiation Vehicle at Fuel Monitoring Facility (FMF) 173 Koji Maeda, Tsuyoshi Nagamine, Yasuo Nakamura, Takeshi Mitsugi and Shinichiro Matsumoto (JNC) 2.9 The Development of Electric Discharge Machine for Hot Cell Usages - 188 Wan-Ho Oh, Sang-Bok Ahn, Sang-Chul Kwon, Yong-Suk Kim and Key-Soon Lee (KAERI) 2.10 Development of Remote Laser Welding Technology 200 Soo-Sung Kim, Woong-Ki Kim, Jung-Won Lee, Myung-Seung Yang and Hyun-Soo Park (Yong-Sun Choo) (KAERI) 2.11 SEM Modification and Shielded Glove Box Design for the Radioactive Material 210 Ki-Seog Seo, Jeong-Hoe Ku, Kyoung-Sik Bang, Ju-Chan Lee, Gil-Sung You, Dae-Seo Ku and Duck-Kee Min (Dae-Seo Koo) (KAERI) JAERI-Conf 99-009
Session 3: Evaluation of PIE Data — - 217
3.1 Detection of Defects in Control Rods by Eddy Current Examination —- 219 Dae-Seo Koo, Jeong-Hoe Ku, Duck-Kee Min, Ro Seung-Gy, Young-Sang Joo and Yoon-Kyu Park (KAERI) 3.2 Life Time Estimation for Irradiation Assisted Mechanical Cracking of PWR RCCA Rodlets -- - -— 227 Takanori Matsuoka and Youichirou Yamaguchi (NDC) 3.3 Surveillance Tests for Light-water Cooled Nuclear Power Reactor Vessels in IMEF 245 Yong-Sun Choo, Sang-Bok Ahn, Dae-Gyu Park, Yang-Hong Jung, Byung-Ok Yoo, Wan-Ho Oh, Seung-Je Baik, Dae-Seo Koo and Key-Soon Lee (KAERI) 3.4 The Fracture Toughness Testing of Unirradiated and Irradiated Zr-2.5Nb CANDU Pressure Tube - - 255 Sang-Bok Ahn, Do-Sik Kim, Dae-Seo Koo, Sang-Chul Kwon and Yong-Suk Kim (KAERI) 3.5 Post-Irradiation Examination of PWR Fuels in KOREA - 267 Young-Bum Chun, Gil-Sung You, Dae-Seo Koo, Eun-Ka Kim,Duck-Kee Min and Seung-Gy Ro (KAERI) 3.6 Hydriding Failure Analysis Based on PIE Data - 278 Yong-Soo Kim (Hanyang Univ.), Hyun-Taek Park, Hwee-Soo Jun(Korea Electric Power Corp.), Yong-Bum Chun, Gil-Sung You, Duck-Kee Min,Eun-Ka Kim and Seung-Gy Ro (KAERI) 3.7 Re-Irradiation Tests of Spent Fuel at JMTR by means of Re-Instrumentation Technique - 286 Jinichi Nakamura, Michio Shimizu, Yasuichi Endo, Hideaki Nabeya, Kenichi Ichise, Junichi Saito, Kunio Oshima and Hiroshi Uetsuka (JAERI) 3.8 HANARO Fuel Gamma Scanning - - - 300 Kwon-Pyo Hong, Tae-Yon Kim, Dae-Gyu Park, Dae-Seo Koo and Bong-Goo Kim (Key-Soon Lee) (KAERI) 3.9 Metallurgical Properties of Power Reactor Fuels after Irradiation — 310 Gil-Sung You, Hang-Suk Seo, Sung-Ho Eom, Duck-Kee Min, Eun-Ka Kim, Dae-Seo Koo and Jun-Sik Ju (KAERI) 3.10 Post Irradiation Examinations for IASCC Study at JAERI - 325 Takashi Tsukada, Yukio Miwa, Hirokazu Tsuji and Hajime Nakajima (JAERI) 3.11 Determination of Irradiation Temperature Using SiC Temperature Monitors 335 Tadashi Maruyama and Shoji Onose (JNC)
IX JAERI-Conf 99-009
3.12 R&D Status and Requirements for PIE in the Fields of the HTGR Fuel and the Innovative Basic Research on High-temperature Engineering — 341 Kazuhiro Sawa, Masahiro Ishihara, Tsutomu Tobita, Junya Sumita, Kimio Hayashi, Taiji Hoshiya, Hajime Sekino and Etsurou Ooeda (JAERI) 3.13 Advanced Post Irradiation Examination for Fusion Reactor Development in JMTR 362 Kunihiko Tsuchiya, Etsuo Ishitsuka, Minoru Uda, Junichi Saito and Hiroshi Kawamura (JAERI) 3.14 Hot Cell Works and Related Irradiation Tests in Fission Reactor for Development of New Materials for Nuclear Application 373 Tatsuo Shikama (Tohoku Univ.)
Summarizing Talk Norikazu Ooka 380 Closing Address Key-Soon Lee 381 Osamu Baba 382
Appendix A Committee 383 Appendix B Schedule 385 Appendix C Program 386 JAERI-Conf 99-009
Key-Soon Lee 2
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(mm) 1.5 Mmfrmmmmrm it, m ^« E« m% ^m 5i* 1.7 NFVfryhyt^Z&tfZmr&mWL 75 $B3 *&, Wi SEii, *P fll^ (NFD) 1.8 NDC^f-4*>y h7#»^« - 89 %, \'m m es «, «» %m (NDO tc® XI JAERI-Conf 99-009 2.4 FEG-TEMiz^^^^Lj-mM^nrzx^yuxm^nwm.^^^ 131 mi ft&, eoj », mm ara (NFD) 2.5 jtSfE^&M&Srffl^fc*';/ ht;Wfl£g»»fe£SEcDlfflfS 140 Do-Sik Kim, Sang-Bok Ahn, Key-Soon Lee, Yong-Suk Kim and Sang-Chul Kwon (KAERI) 2.6 mMGmftniztbnmmmftmzt-jiK^i&ik&mttfflft 151 2.7 #! ©BBSS : 163 IE-, mm mm, ^m 11m, 2.8 FMF[I*^^MIt'J^©SffiiLS#i 173 Wffl Ste, *^ m, ^W ffiSI, H^: ^^, ^TC «-HS (JNC) 2.9 *7 h-fe;WBMflni3£g©lifg 188 Won-Ho Oh, Sang-Bok Ahn, Sang-Chul Kwon, Yong-Suk Kim and Key-Soon Lee (KAERI) 2.10 j&ffimftMu-*fm&&ffi 217 3.1 mwfcUM\z&%Mww 3.3 IMEF t*3ttS67K^$gg©fc*©-9—^7>X5^X h —- 245 Yong-Sun Choo, Sang-Bok Ahn, Dae-Gyu Park, Yang-Hong Jung Byung-Ok Yoo, Wan-Ho Oh, Seung-Je Baik, Dae-Seo Koo and Key-Soon Lee (KAERI) 3.4 CANDU EE^ffl Zr-2.5Nb ft(D&mW&ffllk 255 Sang-Bok Ahn, Do-Sik Kim, Dae-Seo Koo, Sang-Chul Kwon and Yong-Suk Kim (KAERI) xii JAERI-Conf 99-009 3.5 SISKfcttS PWRmM(DmmM%& - - 267 Young-Bum Chun, Gil-Sung You, Dae-Seo Koo, Eun-Ka Kim, Duck-Kee Min and Seung-Gy Ro (KAERI) 3.6 PIE x-*£«fc5#IlSffc8fcSl##f - 278 Yong-Soo Kim(Hanyang Univ.), Hyun-Taek Park, Hwee-Soo Jun (Korea Electric Power Corp.) Yong-Bum Chun, Gil-Sang You, Duck-Kee Min, Eun-Ka Kim and Seung-Gy Ro (KAERI) 3.7 teffl&$$fiffgftffi£fflufc JMTR iz&n&mmmtm -- 286 n%, ±m % mm 3.8 HANARO i^0^>YX + A'->y - - 300 Kwon-Pyo Hong, Tae-Yon Kim, Dae-Gyu Park, Dae-Seo Koo and Bong-Goo Kim (Key-Soon Lee) (KAERI) 3.9 mtspmn 3.12 tmmtm^v^ - 341 R mk en JEW, mm m, nm &w, # ©*, MM mr. mm is, ±& vim mm) 3.13 JMTR ££tt5l£m£*PMfg 3.14 Ii^^^»fC*3ttS»rtt^W^©fcJ&©^rtfigW&^^y h"t;H^lg 373 mm m% u- - - 380 Key-Soon Lee — - 381 Hi§ *& ------382 - - 383 tt&B X^rx^-Jl/ -- -- 385 ttfiC ^U^fyA - - 386 XIII JAERI-Conf 99-009 Opening Address by Toshio Fujishiro Chairman of Organizing Committee Director General, Oarai Research Establishment, JAERI Good morning ladies and gentlemen! It is a great pleasure for us to open "The Third JAERI-KAERI Joint Seminar on Post-Irradiation Examination Technology" here in Oarai Establishment of JAERI. As you may know, this Joint Seminar is held as one of the cooperative programs on PIE techniques based on the arrangement for implementation of "Cooperative Research Program between JAERI and KAERI" The 1st seminar was held in 1992 in JAERI and 2nd in 1995 in KAERI. Through these 2 seminars as well as other activities like the exchange of experts, we have successfully continued effective information exchange on operation and management of PIE facilities, and developments of PIE technique. We now largely depend on nuclear as one of the major energy resources for electricity both in Japan and Korea. Irradiation research of nuclear fuel and materials are vital for safe and economical of the current LWR plants and also for security of future energy source including the development of future fission and fusion reactors. PIE technology should be the most important technology to support this irradiation research. In this seminar, we have about 60 participants from KAERI, Hangyang University, Japan Nuclear Cycle Development Institute, Nuclear Develop Corporation, Nippon Nuclear Fuel Development Co., Ltd. and JAERI. 33 presentation are going to be presented during the full 2 days program. We expect that this 3rd Seminar will provide a good opportunity to exchange information and to establish good personal relationship for future cooperation. I hope active discussions will be made among the participants and this seminar will be successful. Thank you. - 1 - JAERI-Conf 99-009 Opening Address by Key-Soon Lee Director, Nuclear Fuel Cycle Examination Team Korea Atomic Energy Research Institute It is my great pleasure to have an opportunity to make an opening remarks for the 3rd KAERI-JAERI Joint Seminar on PIE technology being held this Oarai. First of all, I would like to express my sincere appreciation to Dr. Fujishiro, general director of Oarai Establishment JAERI, for his help on the joint seminar, and also I express my appreciation to Dr. Baba and Mr. Hoshiya and their staffs members for their efforts in organizing this joint seminar on the technology. We, both of Korean and Japanese scientists, are now preparing for the coming 21st Century. In this coming century, due to the lack of fossil resource in Korea and Japan, nuclear energy could be expected to be an unique energy source for the solution of the lack of energy resources in this two countries, as we understand the nuclear power is an economical and technology self-reliant energy source. Therefore PIE technology will be more important than ever which is essential to the development of nuclear technology to be needed for the safe utilization of the nuclear energy. At the time of the 1st seminar held at Oarai and 2nd seminar at the Daeduk Science Town, we had exchanged a lot of technical information and experiences on the PIE technologies. It is my sincere desire that this seminar also will contribute greatly to the utilization of PIE facility through exchanging of broad technical information, new ideas, and relevant experiences of PIE technology as we did at the previous seminars. As a representative of our participants from Korea, I would like to express that it is our great pleasure to be here for participation in this seminar and to share our friendship with all of you, and again I want to express my appreciation to staff members for preparation of this seminar. I hope this seminar could be remembered to be a useful seminar to all of us. Thank you very much. - 2 - JAERI-Conf 99-009 SESSION 1 CURRENT STATUS AND FUTURE PERSPECTIVES ON PIE PIE ACTIVITY-1 CHAIRS : N. Ooka (JAERI) and K.-S. Lee (KAERI) PIE ACTIVITY-2 CHAIRS : E.-K. Kim (KAERI) and T. Kodaira (JAERI) - 3 - This is a blank page. JP9950622 JAERI-Conf 99-009 1.1 Over View of Nuclear Fuel Cycle Examination Facility at KAERI Key-Soon Lee, Eun-Ga Kim, Kih-Soo Joe, Kil-Jeong Kim, Ki-Hong Kim and Duk-Ki Min Korea Atomic Energy Research Institute Yuseong-Ku, Taejon, Korea ABSTRACT Nuclear fuel cycle examination facilities at the Korea Atomic Energy Research Institute(KAERI) consist of two post-irradiation examination facilities(IMEF & PIEF), one chemistry research facility(CRF), one radiowaste treatment facilty(RWTF) and one radioactive waste form examination facility(RWEF). This paper presents the outline of the nuclear fuel cycle examination facilities in KAERI. INTRODUCTION The research and operation team for nuclear fuel cycle examination of Korea Atomic Energy Research Institute(KAERI) was organized in order to promote the operation efficiency of nuclear fuel cycle examination facilities. The team is responsible for operation of five examination facilities i.e., two facilities for post- irradiation examination of nuclear fuel and materials, one chemistry research facility, one facility for the treatment of low level radiowaste and one facility for safety tests of radioactive waste. The number of persons worked in five facilities is 59 persons and operation budget is about four billion Won(3.3 million US$). This paper introduces outline of the nuclear fuel cycle examination facilities operated by the team. GENERAL DESCRIPTION OF EACH FACILITY 1. Post-Irradiation Examination Facility (PIEF) - 5 - JAERI-Conf 99-009 Kori Unit 1, which was the first nuclear power reactor in Korea, was commercially operated in 1979. So the post-irradiation examination of spent fuel and structure discharged from this reactor was necessary in order to probe the reliability of nuclear fuel and structure material. In response to such approval, the KAERI decided to construct the PIEF for examination of spent fuel and structural material of power reactor. The PIEF started to design and construct in 1980 and completed at the end of 1985. The qualification of the equipments and test operation of the facility was conducted before 1987, and the PIEF has been put into service since 1987. A full size spent fuel discharged from commercial PWR could be subjected to post irradiation examination at this facility. The facility provides adequate services of examination for spent fuel as well as for examination associated with reactor safety, and fuel fabrication technology improvements. The PIEF is a two-story building with a basement. A total floor area of the building is about 5,947 m2 The PIEF has three large pools(9401 -9403), four concrete cells(9405 - 9407) and two lead cells(9408 and 9409). Specifications and detail examination items of the pool and hotcell in the PIEF are summarized in Table 1. 2. Irradiated Material Examination Facilities (IMEF) IMEF was constructed by domestic technology in order to mainly conduct post- irradiation examination of fuel and material irradiated in HANARO(High Flux Advanced Neutron Application Reactor), which provides high neutron flux levels among research reactor in the world. The facility commenced to design and construct in 1988 and completed at the end of 1993. The qualification of the equipments and test operation of the facility had been conducted by the end of 1994 and the IMEF has been put into service since 1995. The facility has three stories and a basement with a total floor area of about 4,000m2, and concrete cells of about 60m in total length. The facility consists of fuel examination cell line, material examination cell line and multiple examination cell line, and has 26 work units(one unit has one window and one pair manipulators) The activities conducted in the fuel examination cell line are a dismantling of capsules and fuel bundle, visual inspections, X-ray radiography, dimensional measurements, gamma scanning and eddy current tests for nondestructive test, and fission gas collections and sample preparation for metallography for destructive test, and the activities in the material examination cell line are the measurement of mechanical and physical properties of irradiated material, impact test, tensile test, heat treatment, and thermal conductivity measurement. A lead cell is also available for metallography, hardness test, and density measurement. - 6 - JAERI-Conf 99-009 The multiple examination cell line is used for the special research handling irradiated fuel and material. This cell line will be used for technology development of fuel fabrication using of PWR spent fuel a special project of Direct Use of PWR spent fuel in CANDU (DUPIC) research. In addition, a shielded electron probe microanalyzer and a transmission electron microscope installed in hot room are used in the examination of irradiated fuel and material. The specifications and detailed examination items carried out in the IMEF are summarized in Table 2. 3. Chemistry Research Facility (CRF) Chemistry Research Facility started to design in 1989 and constructed at the end of 1992. The facility has four stories with total floor area of about 3,300 m2. Major facilities of CRF consisted of 24 cold laboratory including clean laboratory, low level radioactivity measurement laboratory, and one radiochemical laboratory including one set of lead shielded line having three compartments with lead glass windows, two sets of steel boxes, four set of glove boxes and two liquid waste storage tanks capable of 10 cubic meters in capacity. In this facility 43 analytical instruments such as thermo-ionization massspectrometry(TI-MS), ICP-AES with shielded and non-shielded equipment, emission spectrograph, a,/3 ,y- measurement system, quadrupole mass spectrometer, EPMA, hydrogen determinator, XRD, XRF, XPS/Auger/SMS and ICP-MS etc. were installed and put into services of chemical analysis for the research and other works. A major activity of CRF is to do chemical analysis of irradiated samples for the post-irradiation examination such as burnup measurement and non-radioactive samples from the research projects such as fuel development and reactor material development. The lead shielded line is used for dissolution of irradiated fuel. A burnup measurement is performed by chemical analysis of irradiated fuel after dissolution, dilution, separation and determination of burnup monitors of uranium, plutonium and Neodymium using chemical hot cell, glove box and other analytical instruments. In addition, fission product determination, fission gas determination, hydrogen determination in zircaloy tube and some elemental analyses are also performed in this facility. Two liquid waste storage tanks are used for receiving non-radioactive waste solution and radioactive waste solution in each. The received wastes are transferred to Radioactive Waste Treatment Facility for the waste treatment. 4. Radioactive Waste Treatment Facility - 7 - JAERI-Conf 99-009 The radwaste treatment facility(RWTF) was designed and constructed by the domestic technology on the basis of the conceptual design from France. The RWTF has been normally operated since March, 1991 under the license by Korea Institute of Nuclear Safety(KINS). Major waste treatment process consists of liquid waste evaporation, bituminization, solar evaporation and solid waste compaction system. Figure 1 shows the radioactive waste treatment process flow diagram of the RWTF. The low-level waste(5X 10"6 fid/ml < LAW < 1 juCi/ro£) is usually treated by the evaporation process. The evaporator is the forced circulation, long-tube vertical type having high heat coefficients and its evaporation capacity is 1.2 m3/hr. For the bituminization process, the thin film evaporator is installed and its evaporation capacity is 40 I /hr at 240 °C. The straight asphalt 60/70 is used as a matrix agent for the process. The solar evaporation facility was designed and constructed by KAERI's own technology based on the Zero Release Concept for the very low-level liquid radioactive wastes(VLAW). The annual treatment capacity of the solar evaporation facility is 1,200 m3. The VLAW and the condensate from evaporation below the concentration of 5 X 10"6 y. Ci/m£ are treated in the solar evaporation facility. All of the radioactive wastes generated from the post irradiation examination facility(PIEF) and the laboratories in KAERI have been treated in the RWTF, whose major radionuclides are Co-60 and Cs-137. The generated wastes such as very low- active liquid waste(VLAW), low-active liquid waste (LAW), and medium-active liquid waste(MAW) from PIEF are transferred to the RWTF through the pipe line. Spent resin is also transferred by the pipe line. While corrosive active liquid waste(CAW) is transported by the specially designed tank-car. Recently, the radioactive wastes from the research reactor, HANARO, has begun to be generated. The HANARO has been operated since April, 1995. There are three major buildings; the reactor building, the radioactive isotopes production facility(RIPF), and the irradiation material examination facility(IMEF). The radioactive waste from those buildings are collected and transferred to the RWTF through the pipe line depending on the radioactivity level. The maximum volume of the waste treated is estimated to be 240m3/yr for the VLAW, 230m3/yr for the LAW, and 6m3/yr for the MAW. Approximately, 82% of the VLAW and 87% of the LAW at HANARO are generated from the RIPF building. In case of the RIPF building, 88% of the VLAW is the handwashed waste, and 97% of the LAW is from the hot sink, fume hood, and glove box. 5. Radwaste Form Examination Facility (RWEF) - 8 - JAERI-Conf 99-009 Radwaste Form Examination Facility was constructed to establish waste acceptance criteria for disposal site, and now carry out a characterization of waste form to provide technical data for feedback to solidification process(KAERI, KEPCO's NPPs) and for establishment of national regulation. This facility has two stories( 1,350 m2) and was completed to construct in the middle of 1992 through a national inspection of the equipments and installations. The characteristics(radioactivity, mechanical and physicochemical properties) of waste forms against radionuclides release are important. Principal examination items are radionuclides assay in 100~200drum by non-destructive method, compressive strength, long/short-term leach test(including 200drum), thermal cycling test, free liquid test, etc. Core drilling machine and core specimens sizing machine are used for lab. scale examination from real drum(rigid and flexible solidification matrix). These test are carried out remotely in concrete shielded cell. In addition, this facility being capable of handling high activity will be assist and support IMEF and other R&D studies, this year. OPERATION STAFF AND BUDGET Total number of operation staffs are 52 persons and the operational budget of the Nuclear Fuel Cycle Examination Facility is about four billion won (3.3 million US$) and more detailed information is presented in Table 3. - 9 - JAERI-Conf 99-009 Table 1 Specifications and fuctions of pool and hotcell in PIEF Inside Wall Number Pool Dimension Thickness of Functions Major Cell Equipment /Cell WxDxH(m) (cm) Windows Normal 9401 6.5 x 3.0 x Concrete Unloading /Pooll 15.5(D) 110 Normal 9402 6.5 x 3.0 x Concrete Storage /Pool 10.0(D) 110 Visual Inspection, Video Camera Normal 9403 7.5 x 3.0 x Eddy Current Test, Eddy Current Tester Concrete 1 /Pool 15.5(D) Dimensional Measurement, Gamma Scanning Equipment 110 Axial Gamma Scanning Saw Visual Inspection, Porfilometer, Heavy Eddy Current Test, 9404 6.5X1.5X3.5 Concrete 3 Dimensional Measurement, Eddy Current Tester /Cell 85 Axial Gamma Scanning Gamma Scaning Equipment. X-ray Radiography X-ray Equipment Heavy Rod Cutting Rod Cutter 9405 4.0X1.5X3.5 Concrete 2 FP Gas Collection Rod Puncturing Apparatus /Cell 85 Heavy Specimen Storage Storage Rack 9406 2.0X1.5X3.5 Concrete 1 Specimen Identification /Cell 85 Heavy Specimen Preparation Micro Cuter, Mounting Press 9407 3.0X1.5X3.5 Concrete 2 for Metallography Grinder/Polisher, Periscope /Cell 85 Specomen Preparation of SEM Metallography Microscope, 9408 Lead 1.2X1.8X2.5 2 Micro Hardness Tester /Cell 0.2 Hardness Measurement Macrosope Sectional Gamma Scanning Micro Gamma Scaning 9409 Lead 1.2X1.8X2.5 1 Equipment. /Cell 0.2 Density Measurements Balance -10- JAERI-Conf 99-009 Table 2 Specifications and functions of hotcell in IMEF Inside Wall Number Major Cell Cell Dimension Thickness of Functions Equipment W X D X H(m) (m) Windows Visual inspection, Porfilometer, Eddy current test, Dimensional Eddy Current Tester Heavy measurement, Ml 7.0X3.0X6.0 Concrete 3 Axial gamma scanning Gamma Scaning Cell 1.2 Equipment. X-ray radiography X-ray Equipment FP gas collection Rod Puncturing Apparatus Dismantling of capsule Milling Machine, Heavy and fuel bundle Rod Cutter M2 7.0X3.0X6.0 Concrete 3 Preparation of Electric Discharge Cell 1.2 Mechanical test Machine specimen Preparation of Micro Cuter, Heavy metallography sample Mounting Press M3 4.7X3.0X6.0 Concrete 2 Preparation of EPMA Grinder/Polisher, Cell 1.2 sample Periscope Specimen storage Storage Rack Heavy M4 Specimen 2.3X3.0X3.0 Concrete 1 Cell identification 1.2 Charpy impact test Impact Tester, Heavy M5a Heat treatment Heating Furnace 7.1X2.0X4.0 Concrete 3 Cell Physical properties Thermal Diffusivity 0.8 measurements Tester, Dilatometer Tensile/compression Dynamic Tensile Heavy test Tester M5b 4.8X2.0X4.0 Concrete 2 Fatigue test Static Tensile Cell 0.8 Tester, High Scope M6a Heavy Pellet manufacturing Pellet 11.7X2.0X4.0 5 Cell Concrete manufacturing Heavy Rod and bundle Rod and bundle M6b 11.7X2.0X4.0 Concrete 5 manufacturing manufacturing units Cell 1.1 Metallography Microscope, M7 Density measurements Micro Hardness 1.5X2.6X2.65 Lead 0.2 2 Cell Tester Balance - 11 - JAERI-Conf 99-009 Table 3 The Number of persons and operation budget of fuel cycle examination facilities of KAERI in 1998 KEARI Staff Post Doctor Others Budget (million US$) PEEF 12 1 0.94 IMEF 14 2 2 0.79 CRF 10 1 0.67 RWTF 12 1 0.65 RWEF 4 0.26 Fig. 1 Flow Diagram of Radioactive Waste Treatment Process -12- JAERI-Conf 99-009 JP9950623 1.2 ACTIVITIES ON PIE OF NUCLEAR POWER PLANT FUELS IN KAERI Eun-Ka Kim, Yong-Bum Chun, Gil-Sung You, Dae-Seo Koo, Duck-Kee Min, Key-Soon Lee, Seung-Gy Ro ABSTRACT The PWR irradiated fuels were transported from NPPs pool sites to KAERI PIE facility by using a shipping cask. Post-irradiation examination of reactor fuels has been performed to evaluate the performance and integrity in pools and hot cells since 1987. In this paper, 10 years of PIE activities for the power reactor fuels conducted at PIE facility in KAERI were summarized with a brief description of PIE capabilities. INTRODUCTION The nuclear power plant (NPP) in Korea went into commercial operation in 1978. Today total eleven(ll) PWRs and three CANDU NPPs are in operation. From the early of 1980's indigenous nuclear fuel development program has been started. Subsequently, post-irradiation examination (PIE) facility for the nuclear fuels was indispensable to make sure of the integrity and irradiation performance of these fuels. Construction of the PIE facility was completed at the end of 1985. After one year test operation of the facility, it has been put into service for PWR fuels from 1987. TRANSPORTATION OF IRRADIATED PWR FUELS The irradiated fuel assemblies have been transported from PWR nuclear power plant (NPP) pool sites to the KAERI PIE facility by using a shipping cask (KSC-1) which was designed for the transportation of one spent PWR fuel assembly. Seven spent PWR fuel assemblies and one basket containing the 46 defective fuel rods were transported for PIE by 1993. Two defective PWR fuel rods from Yongkwang Unit 4 and three defective PWR fuel rods from Ulchin Unit 2 were transported in 1996 and in 1998, respectively. Table 1 shows the status of PWR fuels transported to PIE facility in KAERI. -13- JAERI-Conf 99-009 POOL EXAMINATION PROCEDURE Fig. 1 shows the layout of ground floor plan of the PIE facility in KAERI. The shipping cask trailer comes in the receiving area through the shipping cask receiving entrance (6421) and the cask is transferred from trailer to decontamination pit by a 50 tone overhead crane. The cask is put upright at center of the room and the cask surface is washed with high-pressurized water spray. And then the cask is connected to the circulation loop for internal decontamination. With this close circuit processing device, cask is cooled down to mitigate the thermal shock when the fuel is loaded in the pool and the pressure and contamination level of radioactivity built up in cask are lowered. After decontaminating the cask, it is transferred to the pool 9401 for unloading of fuel assembly. Then, the fuel assembly is transferred to the storage pool 9402 for a temporary storage before inspection in the next pool. In the third pool, fuel inspection and dismantling are carried out. Before dismantling, the fuel assembly is inspected with the visual and dimensional inspection system (VDIS) by checking the geometrical changes, dimensions, and visual conditions. After finishing these examinations, the top nozzle of the fuel assembly is removed to extract the fuel rods to be examined in hot cell. The extracted fuel rods are transferred one by one to the hot cell 9404 using the lifting cart system. HOT CELL EXAMINATION The fuel rod transferred from the pool 9403 is placed on the rod examination bench for subsequent nondestructive test, which is vertically positioned in the NDT cell 9404. The nondestructive test of fuel rod is performed starting with visual inspection and photography, profilometry, eddy current test, X-ray radiography, and axial gamma scanning, etc. After nondestructive examination, the rod is tilted to the horizontal position and then transferred to the next cell, rod cutting cell 9405. Fission gas sampling is made by a puncturing device incorporated with a system outside the cell 9405 and the samples are sent to the analytical laboratory for chemical composition analysis. Cutting and drilling of the fuel rod are carried out in the cell 9405. Usually the PWR fuel rod is cut as long as 60cm, and six or seven samples of 2cm in length are taken from each rod. The 60cm long sections which are not examined are put in a container, and then transferred to the pool 9402 for a storage. For metallographic sample preparation, sectioning, resin impregnation, mounting, grinding, polishing, and chemical etching are performed in the sample preparation cell -14- JAERI-Conf 99-009 9407. Optical macro- and microscopic examination and photography are performed in the lead cell 9408. Sectional or radial gamma scanning and density measurement of fuel samples are carried out in the lead cell 9409. The burn-up is determined by Nd isotope ratio which is chemically separated from the irradiated fuel samples at cell 7409. Physical and mechanical tests for irradiated fuel materials are performed in the irradiated material examination facility (IMEF) located near PIE facility. PIE OF IRRADIATED PWR FUELS Post-irradiation examination for the discharged and failed PWR fuels has been performed at PIE facility in KAERI. As shown in table 2, irradiated PWR fuels took a series of pool examinations for fuel assembly, hot cell nondestructive examinations of fuel rods and destructive examinations of fuels in hot cells (1-4). PWR Fuel assemblies are examined on visual inspection, dimensional measurement, and burn-up distribution measurement by gamma scanning. Hot cell nondestructive examination includes visual examination and photography, profilometry, eddy current test, X-ray radiography, axial gamma scanning, and oxide layer thickness measurement. Destructive examination of fuel rods covers fission gas sampling and analysis, burn-up measurement, density measurement, metallography, and sectional gamma scanning. SUMMARY The performance and integrity test of the irradiated PWR fuels have been conducted at PIE facility in KAERI since 1997. The spent PWR fuels are continuously examined in KAERI PIE facility as one assembly per annum. Recent concentrations of PIE items are put on the enhancement of the irradiation characteristics of fuels to endure under more severe conditions and environment in consideration of high burn-up. PIE database for supporting safe operation of NPPs and improvements of fuel performances are under construction by using the data obtained through PIE so far. REFERENCES 1. S.G. Ro et al., "Post-Irradiation Examination of Kori-Fuel" KAERI/GP-77/88 (1988). -15- JAERI-Conf 99-009 2. E.K. Kim et al., "Hot Cell Examination of Kori-3 Defective Control Rod" KAERI/TR-141/89(1989). 3. S.K. Lee et al., "Confirmation of Failure Causes of the Kori-2 Cycles 7&8 KOFA Fuel and Remedies against the Fuel Failure" KAERI/TR-387/93 (1993). 4. S.G. Ro et al., "Post-Irradiation Examination of Yongkwang-4 Defective Fuel rods" KAERI/TR-739/96 (1996). -16- j 74W LEPMALab 7430 Low Active 7411 ialfon Radwaste tamrina L* 3. Aniilysis Chemical Hol-Cdl SMctromttr] Lab. I ftutr. Decon. $421 Cask Receivtng Area htcasuriae 8401 Lab. Intervention Act* 7410 Atomic Spectroscope Lab. 6405 O Safety O Control General Chwnbtey Lab. Changing Room Fig. 1. Layout of Ground Floor Plan of the PIEF JAERI-Conf 99-009 Table 1. Transportation of PWR Fuels Irradiated in NPPs Fuel Assy. Date of Date of NPP FA No. Status Type Discharged Transported Kori Unit 1 C15 14x14 17 Apr.'82 April, 1987 Intact 5) A39 55 30 Jan.'81 May, 1987 Intact 55 A17 55 27 Oct. '79 June, 1987 Intact 55 Basket (46 Rods) 55 May, 1988 Defect 55 G23 5? 24 Oct. '86 May, 1990 Intact 55 J14 55 20 Jan.'89 July, 1991 Intact 5) F02 55 17 Sep.'85 May, 1992 Intact Kori Unit 2 J44 16x16 29 May '92 April, 1993 Defect Yongkwang B209-R8 (Rod) 16x16 24 Sep.'95 April, 1996 Defect Unit 4 55 D108-K2 (Rod) 55 55 April, 1996 Defect Ulchin Unit 2 J09-L1 (Rod) 17x17 12 May '97 July, 1998 Defect 55 J09-K1 (Rod) 5) July, 1998 Intact 55 J12-A13 (Rod) July, 1998 Defect - 18 - JAERI-Conf 99-009 Table 2. PIE of Irradiated PWR Fuels Burnup Burnup FA No. Reactor Core Location PEE Status Cycle (MWD/MTU) C15 Kori-1 1/2/3 K3/F8/H3 32,300 Dismantled A39 53 1/2 K7/G7 25,300 5) A17 33 1 J6 17,071 5) Basket (46 » PIE of 4 defect rods) rods G23 4/5/6/7 A8/AS/B7/D7 35,500 Dismantled J14 5) 7/8/9 E9/J5/H11 37,840 n NDEin F02 33 4/5/6 B6/K9/L10 28,300 Pool J44 Kori-2 7/8 C8/C7 35,018 Dismantled B208-R8 (Rod) Yongkwang-4 1 C7 - NDT,DT D108-K2 (Rod) 55 1 D13 - »3 J09-L1 (Rod) Ulchin-2 7 F2 11,806 33 J09-K1 (Rod) 5) 7 F2 35 NDT J12-A13 (Rod) 5S 7 A8 7,210 NDT,DT -19- JP9950624 JAERI-Conf 99-009 1.3 Present Status of PIEs in the Department of Hot Laboratories Tsuneo KODAIRA, Tomohide SUKEGAWA, Hidetoshi AMANO, Fumio KANAIZUKA and Kiyomi SONOBE Department of Hot Laboratories, Tokai Research Establishment Japan Atomic Energy Research Institute Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195, Japan ABSTRACT The Department of Hot Laboratories (DHL) operates three hot cell facilities including the Research Hot Laboratory (RHL), the Reactor Fuel Examination Facility (RFEF) and the Waste Safety Testing Facility (WASTEF). The RHL is performing post irradiation examinations (PIEs) for fuels and materials irradiated in research and test reactors. The RFEF is principally examining the reliability of operating power reactor fuel assemblies for PWR, BWR and ATR. In the WASTEF, development and characterization tests of advanced waste forms have been carried out for a safety examination on disposal of high level waste. The present paper mainly describes current status of PIEs in these facilities and several technical topics concerning measurements of physical and mechanical properties for light water reactor (LWR) fuels and materials. 1. INTRODUCTION DHL is operating three hot cell facilities, i.e. the RHL , the RFEF and the WASTEF. The RHL was established in 1961 and expanded in 1965 in which there are 10 /? y concrete and 38 /3 y lead cells. The RFEF was established in 1979 and it is equipped with 6 j3 y concrete cells and lay ones with 2 lead cells. The WASTEF was established in 1981 and it is equipped with 3 j3 y concrete cells, lay ones with 1 lead cell and 6 glove boxes. In this report, current activities of these facilities are described and several R & D works are also presented as technical topics. -20- JAERI-Conf 99-009 2. CURRENT STATUS 2.1 RHL Figure 1 shows typical PIE flow diagram in the facility, where material examination occupies relatively important position. The activities in the facility are divided into the following four categories: Visual Inspection Dimensicn Ivfeas Irradiated Specimen X Ray Radiography at Eddy Qnot Test y-Ray Scanning Density Kfeasunexnent N4cro-Analysis: Fuel Specimen XRay rjiffiacticn FP C3as Analysis Visual Inspection dimension tvfeas. Pvfeffierial Specimen Hardness Test Tensile Test Bending Test pg "Tests oifFuel ancl Impact Test fcr Txicai ftwer St. Fatigue Test Fracture: Toughness Crack Propagation Punch Fig.l Flow Diagram of PIEs in the RHL (a) PIEs of HTTR Fuel Specimen The High Temperature Engineering Test Reactor (HTTR) established the first criticality on November, 1998 at the Oarai site in JAERI, where prismatic fuel compacts including coated particle are used for the driver fuel. JAERI has been studying the behavior of this type of fuel for long period and still now carry out the irradiation test using the Japan Material testing Reactor (JMTR). For PIEs of the HTTR fuel specimen, the facility has developed lots kinds of special devices such as high temperature annealing and defective fuel detecting devices, for dealing with the coated fuel particles with the size less than 1 mm in diameter. (b) PIEs of NSRR Experimented Fuel Specimen One of the important subjects for reactor fuels is safety investigation under transient power increase. The fuel durability under the condition is examined using the Nuclear - 21 - JAERI-Conf 99-009 Safety Research Reactor (NSRR). Specimens, prepared from the reactor fuel rods with high burnup after finished PIEs, have been supplied for reactivity initiated accident (RIA) experiment. For the specimen after the NSRR power transient test, main targets in PIE are ceramography and EPMA in the defected region. (c) PIEs of Several Materials The facility equips several kinds of apparatus for examining mechanical properties of the materials on reactor pressure vessel, fuel cladding, first wall of fusion reactor and so on. These apparatus are mainly installed in the lead cell line in the facility, which are on tensile, impact, fatigue, fracture toughness and small punch tests. (d) Monitoring Tests for Reactor Fuel and Pressure Vessel from JAPCO For 33 years, the facility has cooperated with the fuel monitoring works on the uranium/magnox driver fuels and the surveillance test on the reactor pressure vessel in the Tokai Atomic Power Station of the Japan Atomic Power Company (JAPCO). PIE items are visual inspection, dimensional measurement, X-ray radiography, metallography and density measurement for the fuel and Charpy impact and tensile tests for the pressure vessel. These monitoring tests have successfully finished this year due to the termination of the reactor operation. 2.2RFEFr> The RFEF is principally used for examining reactor fuels. The flow diagram on PIEs is presented in Figure 2, and the working activities in the facility can be divided into the following six categories: Fuel Asserrtiy Fuel Rod Fuel Specimen Visual Inspection Visual Inspection Cfexwrography Dimension Mam. r-Ray Scanning Buxnup Ivfcas. Rod GapNtas Ciiraisicn IVfcas Density IVfeas, Eddy Curoent Test Qtt-Gas Analysis Oddc Thickness IVfcas XRay Efcffincticn XRay Radiography rvfcJting Pfcint Nfeas. ¥YC Cep rvfaw Therm. ESflu. Ivfcas. FP Gas Analysis IM^, ERVIA. Punc ture Test 7 Scanning Struztual Tvlalenul CWVting Specimen Rod VUthdravteft Ftorce IVfcas. Tensile Test Burst Test SCC Test t^drogn Analysis Fig.2 Flow Diagram of PIEs in the RFEF - 22 - JAERI-Conf 99-009 (a) PIEs of Reactor Fuels PIEs for PWR, BWR and ATR fuel assemblies are being conducted after the establishment of the facility in 1979. The developing targets of reactor fuels in Japan are aiming at higher performance and higher burnup from the standing point of economical view, so that the contents of PIEs are extended and focussed on detailed and microscopic investigations. As a series of high burnup program on PWR fuel, the fuel assembly with the burnup of about 48 GWd/tU was transferred into the facility in March 1996, and PIEs are under way. (b) PIEs of Plutonium Fuels JAERI is investigating the irradiation behavior of uranium/plutonium mixed fuels such as (U,Pu)O2, (U,Pu)C and (U,Pu)N. Up to the present, total seven fuel specimens irradiated in the JMTR have been supplied for PIEs using the cell in the facility. For these examinations, the cells are kept in argon gas atmosphere for preventing the oxidation of specimen. (c) Other PIEs PIEs were conducted for the two kinds of specimens. One is PIEs for the higher burnup UO2 specimens irradiated in the Halden reactor in Norway up to the burnup of about 60GWd/tU. Principal target in PIEs is in ceramography on rim structure. Another is the examination of Rock-Like fuels which are possible to dispose directly. Small fuel samples were irradiated for four cycles at the JRR-3M to clarify irradiation behavior under LWR condition. NDEs (visual inspection, X-ray photography, etc.) and DEs (puncture test, metallography, EPMA, etc.) were performed. (d) Re-assembling of Fuel Rods Finished PIEs The fuel rods finished PIEs in the facility should be re-assembled together with the residual fuel rods and transferred to the reprocessing plant in accordance with the government policy. Re-assembling, nowadays, becomes popular work in the facility. (e) Re-fabrication sand Re-irradiation Test JAERI is performing the reactivity experiments using the NSRR and the power ramping experiments using Boiling Water Capsule (BOCA) instrument in the JMTR, respectively. In the present, reactor fuel rods with high burnup are supplied for these experiments. Re-fabrication means the preparation of capsule for re-irradiation, and total 59 capsules, including PWR, BWR and ATR spent fuel rods with high burnup, have been re-fabricated in the facility. (f)VEGA An experimental program, VEGA (Verification Experiments of FP Gas/Aerosol release) has been performed at JAERI to investigate the release of FP (Fission Products) -23- JAERI-Conf 99-009 including non-volatile and short life radionuclides from irradiated fuel at -3,000 °C under high pressure condition up to 1.0 MPa. One of special features of this program is to investigate the effect of ambient pressure on the FP release from fuel that has never been examined in previous studies. In the experiment, the Japanese PWR/BWR irradiated fuels and TMI-2 debris sample will be used as the test sample. The test facility has installed into the beta/gamma concrete No.5 cell at the RFEF. Four experiments in a year will be schedule after FY 1999. 2.3 WASTEF In the WASTEF, safety examination on disposal of high level waste has been performed, as seen in Figure 3. The working activities are divided into the following five categories: Fabrication ancl Characteristi 'WSiste Fcms Fabrication, and CHiaracteristics of HxaninatLcns en Glass "Wfeiste Forms in Storage and disposal Fig.3 Flow Diagram of Safety Examination in the WASTEF (a) Fabrication of Glass Waste Forms The facility installs a glass vitrification apparatus in the concrete cell No.2, by which a glass waste forms is fabricated for examining its characteristics. High level liquid waste (HLLW) is transferred from the reprocessing plant in the former Power Reactor and Nuclear Fuel Development Corporation, PNC, (The Japan Nuclear Cycle -24- JAERI-Conf 99-009 Development Institute, JNC, at present) to the facility. Basic characteristics, homogeneity and chemical composition are mainly evaluated by using the fabricated waste forms. This study will be terminated within this fiscal year (FY 1998). (b) Fabrication of Synroc Waste Forms Synroc waste forms, one of the advanced materials for confining especially transuranium (TRU) elements, is under investigation. The present waste forms have relatively excellent characteristics comparing with that for glass waste forms; high density, high thermal durability and high confining performance for TRU nuclides. The cooperative study for examining the characteristics in Synroc waste forms has been undertaken between JAERI/Japan and ANSTO/Australia. This research will be finished within FY 1998. (c) Safety Examination on Disposal The volatility of FP and TRU nuclides is one of the subjects to be studied under storage condition and the facility equips a volatility measuring apparatus. On the other hand, de-vitrification behavior in amorphous glass and alpha acceleration test on radiation damage are also carried out for the subjects under disposal condition. The most important examination under disposal condition is leachability of FP and TRU nuclides from the waste forms. The reaching test is now carried out in the reducing condition and the mobility of these nuclides in a specific rock is also examined. (d) Technical Development on Chemical Analysis Chemical analyses occupy an important position in the facility, since the whole specimens after examinations should be subjected to the chemical analyses. The facility has several apparatuses for chemical analyses such as Induction Coupled Plasma (ICP), atomic absorption analysis and radiation spectrometers. Furthermore, technical developments on chemical analyses are continuously carried out according to further new demands. (e) Other Tests New studies are under way concerning the corrosion test of reprocessing plant materials such as stainless steels and zirconium alloys, and the evaluation of basic characteristics on TRU nitrides such as AmN as a R&D for the advanced nuclear fuel cycle. PIEs about the irradiation assisted stress corrosion cracking (IASCC) of stainless steels for a LWR are also planning to be initiated next year. - 25 - JAERI-Conf 99-009 3. TECHNICAL TOPICS For the evaluation of safety, reliability and integrity of LWRs, various PIEs have been carried out in the Development of Hot Laboratories concerning reactor pressure vessel steels, fuel pellets, claddings and so on. The developments of innovative and advanced apparatuses and technologies are essential in order to produce useful and high quality data in accordance with current and future PIE needs. Several technical topics are summarized as follows. 3.1 New PIE apparatuses for thermal properties measurements and microscopic analysis on LWR fuels Measurement of thermal properties and microscopic observation and analysis of irradiated fuels are very important to evaluate fuel behavior in the high burnup state. The RFEF is under development of the Pellet Thermal Capacity measurement (PTC) and Micro Density Measurement (MDM) apparatuses for the former and the Ion Microprobe mass Analyzer (IMA) and shield-type Field Emission Scanning Electron Microscope (FE-SEM) for the latter. These are outlined as follows and more details are described in the separated report2). (a) Pellet Thermal Capacity measurement apparatus (PTC) and Micro Density Measurement apparatus (MDM) Two new apparatuses are developed to improve the accuracy of thermal conductivity of irradiated UO2. The thermal conductivity ( K ) is calculated from the following equation: K = a • Cp * p where a is the thermal diffusivity from the Pellet Thermal Diffusivity Measurement apparatus (PTDM), Cp is the thermal capacity and p is the density. The thermal capacity and the density are generally referred from previous works. Experimental data are strongly needed for the more detailed evaluation of high burnup fuel behavior. PTC based on differential scanning calorimetry and MDM based on immersed method (meta-xylene) are developing to measure the thermal capacity and density using small sample as the same as that for PTDM. Relation among PTDM, PTC and MDM is shown in Figure 4. (b) Ion Microprobe mass Analyzer (IMA) Quantitative and qualitative analyses of micro-region in pellet and cladding are strongly needed to clarify the irradiation behavior of fuels with higher burnup in LWR. The RFEF is now under development of IMA to analyze isotope distribution on the surface as well as the depth direction. IMA consists of beam, sample, vacuum, detection control and data processing systems in the wall (200mm thickness) for radiation shielding. 26 - JAERI-Conf 99-009 The beam system has gas and Cs ion guns as a primary ion source. The gas ion gun is used for releasing positive secondary ion efficiently in order to analyze FP, TRU and so on. Minimum beam radius is less than 0.6 n m. The Cs ion gun is applied to release negative secondary ion efficiently for the analysis of oxygen within oxidation films on the cladding surface. Minimum beam radius is less than 2 n m. The sample handling system mainly consists of analysis chamber, sample entry chamber and sample stage. It is possible to evacuate in the analysis chamber beyond 4X 10~8Pa . The sample stage is possible to move in straight, rotation and tilt motions. The detection system is mounted for mass-analysis (1 to 450amu) of released secondary ions. IMA will be used for irradiated specimens in 1999. Main body of DS C Micro-weighine machine _ Spec imen DSC Signal Heat power suppIy Conttroller and data process Ing unit I 1 - ! 1 P D M 1 r i J=fellet <&• i Ther mal Property ure i in Irradated fuel i pellet measu j Figure 4. Relation among PTDM, PTC and MDM (c) Shield-type Field Emission Scanning Electron Microscope (FE-SEM) The remarkable phenomena occurred in LWR fuel pellets are rim-region on UO2 pellet, hydride and zirconium oxide layer in zircaloy cladding with increasing burnup. In response to these, FE-SEM is under development with remote handling operation. FE-SEM consists of field emission gun, vacuum control and Energy Dispersive x-ray Spectroscope (EDS). The field emission gun has higher brightness than tungsten-hairpin filament type. Resolution of FE-SEM is 2nm and it will be available in 1999 for irradiated sample. -27- JAERI-Conf 99-009 3.2 Development of mechanical testing technologies for LWR pressure vessel steels Mechanical properties of reactor pressure vessel (RPV) steels and structural materials of LWRs at the post irradiation state are the key parameter for the evaluation of safety, structural integrity and lifetime as well as the material development. The mechanical tests at the RHL have been performed for 37 years to support R&D works at JAERI. Recently, the existing Charpy impact testing machine was remodeled in order to improve its accuracy and reliability. By this remodeling, absorbed energy and other useful information can be delivered from one-time blowing. In addition, the remote machining technology from actually irradiated RPV steels has been developed in order to clarify the aging behavior of LWRs at the RHL. Another new technique is developed to determine the post-irradiation fatigue characteristics of structural and fuel cladding materials as low and high-cycle fatigue tests technology with the function as tensile test equipment. This paper outlines two mechanical testing apparatuses and techniques and remote machining of mechanical test pieces for irradiated LWR-RPV steels as follows. More details are described in the separated report3). (a) Remodeled Charpy impact testing machine The Charpy impact testing machine was redesigned and modified in order to clarify the neutron irradiation embrittlement behavior of LWR-RPV. This machine instrumented with electronic measuring devices to detect an impact force and a displacement of specimen has an automatic specimen setting system. The block diagram of instrumented Charpy impact testing machine is shown in Figure 5. displacement aeasuring sytten force measuring system Figure 5. Block diagram of instrumented Charpy impact testing machine - 28 - JAERI-Conf 99-009 The load capacity is 300J and it is possible to test in the temperature range from -140°C to 240°C by using two types of agitated liquid baths. The test specimen is transferred from the cooling (or heating) bath to an anvil of the machine using industrial robot, and struck by a hammer within 4 sec after removal it from the medium. The test items are V-notch Charpy impact test and K, d dynamic fracture toughness test. The sensor for the load detection was composed of two semiconductor active strain gages on the tup and two dummy gages put on near the hammer. Moreover, a potentiometer for the displacement detection was inserted and fixed to the hammer shaft. These signals from sensors are recorded in the wave-memory with the capacity of 32Kwords x 2 channels. Collected data are utilized for data processing and analysis. (b) Remote machining from irradiated RPV steels In case of irradiated material, since all of manipulation must be handled remotely, machining of the mechanical test specimen should be performed accurately in accordance with the material testing standards such as ISO, ASTM and JIS. However, the remote machining with high accuracy has never been done up to date because the requirement is quite difficult. Therefore, the original machine for general use is modified according to some requirements from radiation environment, free maintenance and higher performance. Moreover, the innovational techniques are applied to achieve the allowable machining by means of remote handling. A numerically controlled machine tool is selected and developed as the most useful apparatus for hot cell work without a human error. As shown in Figure 6, a computerized numerical control (CNC) milling machine developed is composed with the main body for machining and a control system included a personal computer. The machining programs developed in the RHL are a Charpy impact test specimen type A of 10 x 10 x 55(mm) so-called V-Charpy, a plate type tensile test specimen with parallel part of 22.95LX 3Wx 3t(mm), and a three point bending type fracture toughness test specimen with knife-edges. (c) Remote system technology for fatigue testing One of the important research subjects on the LWR fuel cladding performance at extending burnup is to understand the mechanical properties. The RHL developed an electro-hydraulic fatigue testing machine with two kinds of load cells and servo valves in tandem and in parallel respectively. By exchanging the test fixtures with remote handling, the machine is utilized for a high-cycle fatigue test with arc-shaped specimen machined from LWR fuel cladding, a low-cycle fatigue test with round specimen from structural materials, a crack propagation test and a high-frequency test. Moreover, tensile test, plain strain fracture toughness test and the fatigue pre-cracking for fracture toughness specimen are also possible. -29- JAERI-Conf 99-009 Figure 6. Computerized numerical control (CNC) milling machine 4. CONCLUDING REMAKS As described above, the DHL is actively carried out in a wide range of PIEs and are being contributed for the advance of R&D works in JAERI. The development of new and advanced PIE techniques is now very important from the viewpoint of progress of innovative and basic researches as well as R&D in nuclear energy. 5. ACKNOWLEDGEMENT The authors wish to express their sincere thanks to the staffs of Department of Hot Laboratories for performing PIEs and useful discussions. - 30 JAERI-Conf 99-009 6. REFERENCES 1) T. Kodaira, T. Yamahara, T. Sukegawa, Y. Nishino, H. Kanazawa, H. Amano and M. Nakata ; HPR-349 "Current Status of PIE Techniques in RFEF", Enlarged HPG meeting (Lillehammer, Norway, 1998) 2) K. Harada, N. Mita, Y. Nishino and H. Amano ; "Development and Application of PIE Apparatuses for High Burnup LWR Fuels", Third JAERI-KAERI Joint Seminar on PIE Technology (Oarai, Japan, March 25-26, 1999) 3) M. Nishi, M. Kizaki and T.Sukegawa; "Development of PIE Techniques for Irradiated LWR Pressure Vessel Steels", Third JAERI-KAERI Joint Seminar on PIE Technology (Oarai, Japan, March 25-26, 1999) - 31 - JAERI-Conf 99-009 1.4 CURRENT STATUS AND FUTURE PROSPECTS OF JMTR HOT LABORATORY Osamu BABA, Norikazu OOKA and Taiji HOSHIYA Department of JMTR, Oarai Research Establishment, JAERI Oarai-machi, Higashi-ibaraki-gun, Ibaraki-ken, 311-1394 Japan ABSTRACT A wide variety of post-irradiation examinations (PIEs) for research and development of nuclear fuels and materials to be utilized in nuclear field is available in three kinds of (3 — y hot cells; concrete, lead and steel cells in the JMTR Hot Laboratory (JMTR HL) associated with the Japan Materials Testing Reactor (JMTR). In addition to PIEs, re-capsuling including re- instrumentation on the irradiated specimen is currently conducted for the power ramping tests of the LWR fuels using the Boiling Water Capsule (BOCA) or for the re-irradiation tests in the different neutron fields (coupling irradiation test). The newly developed techniques by the JMTR HL have provided us with the key information about the irradiation effects on mechanical and physical properties of the specimen in various environments as fission and fusion reactors. These techniques are focused on several topics as follows; (1) miniaturized specimen test as an advanced mechanical test, (2) slow strain rate tensile test (SSRT) and crack propagation measurement in high temperature and pressure water for the study of Irradiation Assisted Stress Corrosion Cracking (IASCC) of LWR core internals, (3) handling technique on materials containing tritium for the research and development of tritium breeders and neutron multiplier for fusion reactors, (4) jointing method using the conventional Tungsten Inert Gas (TIG) welding for re-assembling of irradiation capsules and/or re-fabrication of specimen, and (5) Nondestructive examination using ultrasonic wave and infrared thermography for the quantitative evaluation of irradiation embrittlement of structural materials in fission and fusion reactors. As there are various PIE facilities around Oarai site, mutual exchange of PIE information, interchange of researchers and mutual utilization on PIE facilities are desired to raise the scientific and technical potential on PIE and to get the break-through of the study in the field of nuclear applications. INTRODUCTION The JMTR HL associated with the Japan Materials Testing Reactor (JMTR) was put into service in 1971 to examine specimens irradiated mainly in the JMTR. A wide variety of PIEs - 32 JAERI-Conf 99-009 for research and development of nuclear fuels and materials is available in three kinds of P - 7 hot cells in the JMTR HL. These examinations are on LWR high burn up fuels subjected to power ramping tests, NSRR test fuel, structural materials for LWRs, HTGRs and fusion reactors, shape memory alloys and others. In addition to PIEs, re-capsuling including re-instrumentation [1] is currently conducted for the power ramping tests using the Boiling Water Capsule (BOCA) [2] or for coupling irradiation tests [3]. The newly developed techniques by the JMTR HL have provided us with the key information about the irradiation effects on mechanical and physical properties of the specimen in various environments as fission and fusion reactors. In this paper, the facilities of the JMTR HL are described. Current PIE activities and advanced PIE techniques in the JMTR HL are also presented. FACILITIES AND FUNCTIONS Figure 1 shows the process of PIE for irradiation sample in the JMTR HL. The JMTR HL is located adjacent to the JMTR, and the irradiated capsules can be transferred through a canal. Some of the irradiated specimens and radioisotope (RI) drawn out from the capsules in the JMTR HL are supplied to the hot laboratory of Tohoku University and RI production facility of JAERI, respectively. In addition, a lot of fuel rod segments, which have been irradiated in power reactors and transported to the JMTR HL, are transferred through a canal to the JMTR after re- instrumented and re-assembled into new capsules in the JMTR HL. Figure 2 shows the arrangement in the ground floor of the JMTR HL. Three trains of 3 - y cells, i.e. 8 concrete cells attached with 4 microscope lead cells, 7 lead cells and 5 steel cells are available for PIEs on irradiated fuels and materials. No a type of cell for MOX fuels is provided. Six globe-boxes for handling tritium-containing materials can be used for the study of fusion blanket. Dismantling irradiated capsules, re-capsuling, re-instrumentation, destructive and nondestructive examinations, microstructure observations of fuel and material specimens are performed in the concrete cells. The lead and steel cells are used for many kinds of material tests such as tensile test, Charpy impact test, miniaturized specimen test, SSRT, stress corrosion cracking test, fatigue test, fracture toughness test, creep test and so on. Furthermore, a scanning electron microscope (SEM) is installed in the lead cell to observe the fracture surface of the tested specimens. CURRENT PIE ACTIVITIES IN THE JMTR HOT LABORATORY 3.1 Re-instrumentation for fuel of the light water reactor The information on FP gas pressure and centerline temperature of fuel rods during power transient is very important to realize load-following operation and achieve high burn up of LWR fuels. Special techniques on re-instrumentation of FP gas pressure gauge and/or centerline thermocouple were developed in 1990 and 1994, respectively and have already been put into service for the BOCA. Figure 3 shows the re-instrumentation procedures of a center-line thermocouple to an irradiated fuel pellet. This technique consists of several steps; fixation and freezing the irradiated fuel pellets, drilling to make a center-hole, removing of small tips of pellets -33- JAERI-Conf 99-009 during drilling and welding the top cover attached an FP pressure gauge and a centerline thermocouple. 3.2 Slow strain rate tensile testing (SSRT)forthe study of Irradiation Assisted Stress Corrosion Cracking (IASCC) By the SSRT method, the susceptibility of the irradiated specimen to the stress corrosion cracking (SCC) can be evaluated under the environment of high temperature and high-pressure water. The experiment device for SSRT has a tensile-test machine, an autoclave and a water circulating system with a water purification system as shown in Fig. 4. The IASCC susceptibility and fracture morphologies are obtained from the fractions of SCC area, which are measured by a remote-controlled scanning electron microscope (SEM). In the recent study of IASCC susceptibility, the dependence of the type 304 and 316 stainless steels to IASCC susceptibility on alloy composition, neutron fluence and dissolved oxygen is reported using SSRT experiments [4]. 3.3 Miniaturizing testing It is very important to develop material testing technology with miniaturized specimens (0.1 to lmm in minimum dimension), especially for the development of fusion reactor materials. An ion-accelerator base intense high-energy neutron source has been planned to build for the irradiation tests for the development of fusion reactor materials. Because of the limited volume available for the irradiation by an ion-accelerator, the miniaturization of the specimens is inevitable. In addition, this technology is beneficial for reducing radioactive wastes and efficient use of surveillance test specimens of LWRs. Developments of remote operation techniques and equipment are necessary due to the limited manipulation in the hot cells. Developments of the small punch test machine [5], electrical discharge machining device and a computer-aided micromanipulator for handling the miniaturized specimens have been successfully carried out at the JMTR HL. By using the small punch testing apparatus shown in Fig. 5, the load-displacement curves are obtained from the punching force and the displacement of the puncher. The deformation of the specimen is similar to that of the bulge test. From the analysis of the curves, ductile to brittle transition temperature, fracture toughness and so forth are suggested to be obtained. Furthermore, the research and development on hardness test, impact test and so forth with miniaturized specimens have been carried out as the future techniques. 3.4 Tritium handling Beryllium is a candidate for neutron multiplier and plasma-facing material in fusion reactors, and beryllium irradiation studies have been performed to obtain engineering data for blanket design. The most important point of PIE for beryllium is to manage tritium released from irradiated samples. Figure 6 shows a new facility for PIE of irradiated beryllium in the JMTR HL [6]. This facility consists of four glove boxes, a dry air supplier, a tritium monitor and -34- JAERI-Conf 99-009 removal system, and a storage box for irradiated samples. Maximum amount of tritium to be handled in the facility is 7.4 GBq/day"1. 3.5 Welding technique Welding technique in hot cells is one of the key issues to support the PIE. In the JMTR HL, four kinds of welding techniques have already been developed to fabricate (1) new capsules with irradiated specimens for re-irradiation tests(re-capsuling technique), (2) new specimens from irradiated and tested materials for re-irradiation, (3) instrumented fuel rods from irradiated ones as the FP gas pressure gauge and thermocouple for measurement of centerline temperature (re-instrumentation technique), and (4) Co-60 source from irradiated reactivity adjusting elements in the JMTR. FUTURE PIE TECHNIQUE 4.1 Achievement ofPIEs in short turn-around time Shortening the turn-around time for PIE is one of the key issues to improve the utility of the JMTR HL. For this issue, the following items are thought necessary. (1) Establishing automatic machines for time consuming tests and pre-test procedures such as high and/or low temperature Charpy impact test sample polishing for metallography etc. (2) Modularize PIE apparatus for quick replacement and installation. (3) Modification of PIE apparatus for easy and quick decontamination. 4.2 Application of advanced NDT technique Continuous monitoring on mechanical and/or physical property changes of the same samples through neutron fluence is not possible with destructive testiness. Many samples from the same material are usually prepared and used for this purpose accepting some error included using different samples. From this point of view, nondestructive testings (NDTs) can be a powerful tool to trace these changes induced by neutron irradiation on the same sample, if the correlation between the parameters affecting NDT characteristics and these properties becomes clear. According to a preliminary research using ultrasonic wave tests (UTs) [9], changes in ultrasonic velocity and attenuation coefficient have found some correlation with embrittlement of materials by fast neutron irradiation. Figure 7 shows the schematic diagram of the ultrasonic wave measurement system installed in the lead cell of the JMTR HL to examine the characteristics of the ultrasonic wave for the irradiated small specimens, used Charpy test ones. It may be possible that NDT can be applied to evaluate some mechanical properties of irradiated specimens, although NDT is normally used as a detecting tool for the defects. In the JMTR HL, the application of NDT technology to PIEs is thought very effective and desirable from the view points of monitoring the same samples, saving testing time, rad-waste management and so on. - 35 - JAERI-Conf 99-009 FUTURE PERSPECTIVES OF JMTR HOT LABORATORY There are many PIE facilities around Oarai site as shown Fig. 8. Mutual exchange of PIE information among these facilities, interchange of researchers and mutual utilization of PIE facilities are very desirable to raise the scientific and technical potential in the irradiation research and to get break-through of the study in the field of nuclear application. ACKNOWLEDGEMENT Authors would like to acknowledge to colleagues in the JMTR HL for the support of this presentation. REFERENCES [1] M. Shimizsu, J. Saito, K. Oshima, Y. Endo, T. Ishii, T. Nakagawa, S. Souzawa, K. Kawamata, Y. Tayama, H. Kawamura, H. Sakai and R. Oyamada, JAERI-Tech 95- 037(1993) 185 (in Japanese). [2] Department of JMTR, Annual Report of JMTR FY1996 (Apr. 1, 1996 -Mar. 31, 1997) , JAERI-Review 98-004, p20. [3] Department of JMTR, ibid., p50. [4] T. Tsukada et al., Proceedings 7th Int. Symp. on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol. B (1997) p795. [5] M. Ohmi et al., J. of the Atomic Energy Soc. of Japan, Vol. 39, No. 11 (1997) p966-974 [6] E. Ishitsuka et al., Fusion Engineering and Design 41 (1998) pi95-200 [7] T. Shikamaetal., Nucl. Instr. and Meth. B 91 (1994) 342-345 [8] N. Sekimura, JAERI-Conf 97-006 (1997) 137-143 [9] T. Ishii and N. Ooka, Proceedings of EC-IAEA Specialist Meeting (1999-3) -36- I Flow of Irradiation Sample HL of Tohoku Fuels Materials R Univ. Hot Laboratory- (HL) > m 2 n ri Hot LaboraioP. o O o Re-irradiation of Fuels and Materials * Re-instrumentation Machining TC&FP pressure gage Welding J{ ' Capsuling Post Irradiation Examination rradiafion Fig. 1 Flow diagram of PIE for irradiation sample _rv J—qpj f Service Area Hoi Mock-Up Radiation Dark II Preparation I gaKS Rmim Loading Dock Management 111—I S-2 S-l JRoom J..Berylliu. m r\^l Characteristics Operation stee! Cells Area . Lead Cells Office Durk IJControl Service Area Room n n rv^ . n 9 JMTR- Tinnrrf tf Tr TTTT T 6 Microscope Concrete Cells O Lead Ce!ls Operation Area JMTR Hot Laboratory: (I 1 2 3 4 5 6 7 8 » Kim I , I . i , I , i , [ , I i I i I , I , I Fig.2 Functional Area of JMTRHL Hz Removing of Freezing Drilling UO2 chips > en to dual 1 instrumentation- O device O 0 o II U n n o a N2 "IT" o GOO Xe Kr| He! Inserting of Sleeve thermeeouple ing out inserting Fig. 3 Re-instrumentation procedures of themocouple to irradiated fuel rod JAERI-Conf 99-009 Water makeup system Monitoring/purification system Ion exchanger Drain out .—C i DO pH Cond, High pressure Pressure SSRT autoclave pump regulator Cooler •JCUUH'I! /Accumulator MJS316 (ID Safety 1.71 valve High pressure filter r c~ Heat exchanger Preheater External reference electrode On top of hot cell L- High temp./press.circulation system Hot cell Fig.4 Schematic flow diagram of IASCC test facility. 40 Load cell Extensometer Punching rod Load celf Vacuum .or > en n o © o Push rod for Af specimen clamp en !UiUCi Size of body height: 1.73m width : 0.73m length : 0.66m Fig.5 Remote controlled SP testing machine JAERI-Conf 99-009 Storage box Exhaust Tritium process monitor blower GB-5 GB-1 GBJ-2/3 GB-4 Dry air supplier )V3 GB-6 Tritium removal system GB : Glove box V : Changing valve of normal and tritium removal mode — Normal mode ~ Abnormal mode ( after detecting tritium leakage ) Fig.6 Ventilation system at Be PIE Facility Manipulator J Data processing unit Ultrasonic test instrument Clamping rig of probe v Probe - Irradiated charpy impact specimen Fig.7 Schematic view of experimental apparatus for nondestructive evaluation of irradiation embrittlement in hot cell.2 -42- JAERI-Conf 99-009 Accelarator HTTR JMTR Collaboration of PIE facilities JNC ' J . i FMF AGF MMF LWR LWR Fusion High burnup fuel RPV Blanket Material R MOX fuel Core internal Advanced material Oarai Res. Est. Tohoku Univ. IJOYO(JNC) NFD MMF(JNC) University—National laboratory—Private laboratory Cooperative research, Research exchange, Mutual facilities utilization, Information exchange etc. Fig.8 Cooperation of post irradiation examination facilities 43 - JP9950626 JAERI-Conf 99-009 1.5 OVER VIEW OF POST-IRRADIATION EXAMINATION FACILITIES FOR FUELS AND MATERIALS DEVELOPMENT OF FAST REACTOR Masahiko Itoh O-arai Engineering Center, Japan Nuclear Cycle Development Institute 4002 Narita-cho, Oarai-machi, Higasiibaraki-gun, Ibaraki, 311-1393 Abstract The hot cell complex for post-irradiation examination of the fast reactor fuels and materials was constructed and has been operated at the o-arai engineering center of Japan Nuclear Cycle Development Institute. The complex Consists of three hot cell facilities. They are the Fuel Monitoring Facility (FMF), the Alpha-Gamma Facility (AGF) and the Materials Monitoring Facility (MMF). The FMF is located adjacent to the experimental fast reactor "JOYO" and started operation in November 1978. In this facility, nondestructive examination of fuel subassemblies and other core components, in addition to some destructive examination of fuel and absorber pins, are carried out. The selected pins and materials, sectioned to the appropriate size at the FMF, are sent to the AGF and the MMF for further detailed examinations. The AGF has been operated successfully since October 1971. The functions of this facility are the physical, metallurgical and chemical examinations of irradiated plutonium-bearing fuels. The MMF was constructed at 1972 and has been operated since 1973 for the reactor materials. In this facility, various tests are conducted on core materials, structural material and control rod materials irradiated in fast reactor. Structural materials irradiated in JMTR and pressure tubes irradiated in prototype advanced thermal reactor "Fugen" are also examined. INTRODUCTION In the fast reactor, core fuels and materials are exposed in severe environment comparing to thermal reactor. For the commercial application , it is vitally necessary to improve the performance of fuels and materials. The Power Reactor and Nuclear Fuel Development Corporation was established in 1968 .(restructuring to Japan Nuclear Cycle Development Institute (JNC) at October, 1998) Since 1968, irradiation experiments on fuels and materials have been performed in overseas reactors and Japan materials testing reactor (JMTR), while the construction - 44 - JAERI-Conf 99-009 of post-irradiation examination facilities in O-arai engineering center had also been promoted. The alpha-Gamma Facility (AGF)° for the fuel examination, the Materials Monitoring Facility (MMF)2) for reactor materials examination and the Fuel Monitoring Facility (FMF)36) for the examination of core components of experimental fast reactor "JOYO" were constructed and started operation in 1971,1973 and 1978 respectively. The core modification from the breeding core (MK-1 core) to the irradiation core (MK-n core) was performed in 1982 . Irradiation of fuels and materials in MK-n core was started in 1983. Special devise for irradiation was developed, that is reloaded type irradiation vehicle which is describe elsewhere in detail71. To examine a lot of material specimens in the irradiation vehicle, Material Monitoring Facility was extended and the operation of extended facility was started in 1983. The prototype fast breeder reactor "Monju" reached its fast criticality in 1994. To evaluation of irradiation performance for core component of "Monju", Fuel Monitoring Facility was extended . The construction of extended Facility was started in 1991, and completed in 1995. Pre-operational test for the facility is underway . In JNC , it was established the plan of the Advanced Nuclear Fuel Recycle which aims to reduce the burden of geological disposal by separating long-life nuclide such as Neptunium , Americium and Curium (Minor Actinides) in spent fuels and burning as the Minor Actinides fuels . MA containing fuels have high radio-activity , so it is necessary to fabricate remotely. Therefore, it was planed to install the fuel fabrication apparatus in the hot cell in AGF. Installation of MA containing fuel fabrication apparatus was completed in1998. The functionary tests for the apparatus is underway. FUNCTION OF POST-IRRADIATION EXAMINATION FACILITIES The role of the post-irradiation examination facilities in O-arai Engineering Center is to examine the fuels and materials irradiated in "JOYO" and other fast reactors for the development of high performance fuels . The fuel performance test and surveillance test of structural materials for "JOYO" are also conducted . Furthermore , surveillance test for the pressure tubes of advanced thermal reactor "Fugen" is conducted in MMF . The outline of post-irradiation examination facilities is as follows. Fuel Monitoring Facility The facility has four stories above ground and two stories underground.The layout of the first floor is shown in Fig.1. FMF is divided into two area , FMF-1 and FMF-2. The three main concrete cells .Examination cell, Decontamination cell and Clean cell, in FMF-1 are located in first floor. These cell's wall is made of heavy concrete. The shielding capability of the wall facilitates the examination of fuels and materials of 6X 1016Bq for gamma ray in the examination cell. The metallography cell with steel shielding - 45 - JAERI-Conf 99-009 is also located in the first floor. Specimen transfer is carried out with a pneumatic transfer tube interconnected between the decontamination cell and metallography cell. The handling activity of this cell is 1.1X1013Bq for gamma ray. Another concrete cell for the X-ray radiography of assemblies is provided in the second basement under the examination cell. The Examination cell, Decontamination cell and Metallography cell are large a - y type cells . The Clean cell and radiography cell are 0 - y type . The atmosphere of Examination cell is maintained as high purity nitrogen with H2O and O2 being less than 100ppm .Metallography cell is also nitrogen atmosphere with H2O and O2 being less than 500 ppm . In FMF-2 , The two main a-y type concrete cells , Examination cell and Decontamination cell, are located in first floor and the /? - y type X-ray CT cell is located in the second basement The a-y type concrete cell's wall is made of heavy concrete . The shielding capability is 6.7X1016Bq for gamma ray. The examination cell is maintained of high purity nitrogen atmosphere. The role of this facility is as follows . • non-destructive examination and disassembling of "JOYO" core component • Metallography of fuel pin • Preperation of the sample for AGF and MMF The post-irradiation examination items for hot cells are shown in table 1. Alpha-gamma facility The main functions of this facility are fuel burn-up analysis .physical property measurement, chemical analyses for MA elements in MOX fuel and MA bearing MOX fuel fabrication . The layout of the first floor is shown in Fig. 2 . The facility is three stories high with basement. The hot cells are located on first floor. The inner box-type cell is a characteristic, in which air tight stainless steel movable boxes are installed in radiation shielding such as concrete , lead , etc. The specimens are accepted using a transfer container at the No.1 cell. Inter-cell transfer is provided by a conveyer installed under the inner boxes. The specimens transfer from concrete cell to lead cell are carried out with apneumatic transfer tube which is installed between No.7 cell and No. 13 cell. The shielding capability for gamma ray is 1.6X1014Bq for No.1 cell to No.3 cell, 2.6X 1013 Bq for No.4 cell to No.7 cell and 2.6X 1012 Bq for No.11 cell to No.18 cell. The main post-irradiation examination items are shown in table 2. Materials Monitoring Facility The main functions of this facility are the evaluation of irradiation effect for mechanical properties, physical properties , microstructure for reactor materials. The layout of the first floor is shown in Fig. 3. The facility is two stories high with basement. - 46- JAERI-Conf 99-009 This facility is divided into two area , MMF-1 and MMF-2 . MMF-1 has six concrete cells (one is located in basement) and two lead cells . The No.1 concrete cell is a - y type, the other are /3 - y type . The front wall for the cells from No.1 cell to No. 3 cell is made of heavy concrete and others are ordinary concrete .The shielding capability of each cells for gamma ray of "'Co is shown in table 3. In the laboratories , the physical properties measurement apparatus are provided . MMF-2 has four concrete cells and one iron cell. one concrete is divided into two parts by stainless steel wall shown in Fig.3. No.1 cell and No.2-1 cell are a-y type and another cells are j3 - y type. It is possible for two a-y type cells to change the atmosphere of air to nitrogen . The shielding capability for gamma ray of ^Co is 1.8X 1013 Bq for No1 to No.4 cell and 3.7 X1013 Bq for No5 cell. To evaluate the microstructure of core materials, transmission electron microscope with accelerate voltage of 400keV is provided in the EM laboratory . The main post-irradiation examination items are shown in table 4. POST-IRRADIATION EXAMINATION After the criticality of "JOYO", the examination of fuels and materials irradiated in "JOYO" have been performed mainly . The "JOYO" core component such as core fuel assembly, control rod , reflector etc. and irradiation rig are transferred to the FMF. After non-destructive examination , the assemblies are dismantled and sectioned to the small segments to transfer to other facilities . The post-irradiation examination flow is shown in Fig.4 . Up to now, 63 fuel assemblies, 21 irradiation vehicles with fuel pin , 18 control rods , 39 irradiation vehicles for material etc. are provided to post-irradiation examination . The results of post-irradiation examination contribute to fuel design for "JOYO" and "Monju". Furthermore ,15 surveillance test rigs for "JOYO" and 14 surveillance capsules for "Fugen" were also provided to post-irradiation examination , the results was used to confirm the integrity of reactor pressure vessel and pressure tube , respectively. CONCLUSION Post-irradiation examination facilities in O-arai engineering center, JNC, are described. These facilities have been operating successfully since operations were started. Irradiation performance of the "JOYO" fuel design was confirmed through its PIE results. The fuels and materials development of the fast reactor is being steadily progress in JNC . Further research and development for fuels and materials is being pursued to utilize the PIE facilities. - 47 - JAERI-Conf 99-009 REFERENCE 1) K.Uematsu , Y.lshida , S.Kobayashi and J.Komatsu ; Proc. 22nd Conf. "Remote Systems Technology", 1974, pp.3-10 2) K.Uematsu , Y.lshida , K.Suzuki; pp11-19 in reference 1 3) K.Uematsu , Y.lshida , S.Seki and T.Hayashi; pp.20-29 in reference 1 4) T.ltaki, T.Shimada and H.Tachi ;Proc. 30th Conf. "Remote Systems Technology vol.2", 1982,pp.16-23 5) S.lwanaga , Y.Nakamura , T.Nagamine and A.Koizumi; Proc. International Conf. "Fuel Management and Handling ", Edinburgh, BNES,20-22,March,1995,PaperNo.56 6) T.ltaki, J.Komatsu S.Yamanouchi and Y.Enokido ; pp24-31 in reference 4 7) K.Maeda , T.Nagamine , Y.Nakamura , T.Mitsugi and S.Matsumoto ; This Seminar, 25-26, March, 1999 Table 1 Function of Cells in FMF Area Function Visual Examination Profilometry Gamma Scanning Exam. Cell Pin Puncturing Dismantling and Sectioning Strage Can Welding FMF-1 Decon. Cell Decontamination for In-cell Equipment Clean Cell • Reassembling for Irradiation Vehicle Metallography • Microstructural Analysis Cell Radiography Cell • X-ray Radiography • Visual Examination • Profilometry Exam. Cell • Gamma Scanning • Eddy Current • Dismantling FMF-2 • Strage • Reassembling for Irradiation Vehicle Decon. Cell • Decontamination for In-cell Equipment X-ray CT Cell • X-ray Radiography and X-ray Computer Tomography -48- JAERI-Conf 99-009 Table 2 Function of Cells and Laboratries in AGF Area Function No. 1-1 • Inspection of MA Containing MOX fuel No. 1-2 No 2 • Storage No 3-1 • Fuel Fabrication Concrete No 3-2 Cell No 4 • Sample Preparation for Physical Property Measurement No 5 • Sample Preparation for Metallography No 6 • Sample Preparation for Chemical Analysis No 7 • Transfer by Pneumatic Tube No 8 • Decontamination of In-cell Device No 9 No 11 • Microchemical Analysis No 12 • Metallography No 13 • Transfer by Pneumatic Tube No 14 • Analysis for Fission Product Released from Lead Fuel Cell No 15 • X-ray Analysis No 16 • Density Measurement No 17 • Melting Temperature Measurement No 18 • Thermal Diffusivity Measurement Chemical Laboratry • Chemical Analysis for MA Physical Laboratry • Analysis of MA (ICP-AES) Laboratry • Burn-up Measurement by Mass Spectroscopy Table 3 Handling Activity of cells in MMF-1 Shielding Materials Radioactivity Area of front wall (Bq) No.1 Heavy Concrete 9.7X1012 No.2 Heavy Concrete 3.7X1013 No.3 Heavy Concrete 4.2X1012 No.4 Ordinary Concrete 1.8X1012 No.5 Ordinary Concrete 3.1X1012 No.6 Lead 1.5X1011 No.7 Lead 1.7X1012 No.8 Ordinary Concrete 3.7X1013 -49- JAERI-Conf 99-009 Table 4 Function of Cells and Laboratories in MMF Area Function © Cladding Test Cell • Mechanical test for fuel cladding • loading , unloading Loading Cell © Machining Cell • Specimen preperation ® Metallurgy Cell • Specimenn preperation for metallurgy © Mechanical Test Cell • Mechanical test for structural materials MMF-1 © Microscope Cell • Optical microscope observation © Uniaxial Creep Cell • Uniaxial creep test for structural materials Storage Cell • Specimenn storage Gas Analysis • Retaind gas analysis Room Physical • Thermal conductivity mesurement Laboratory • X-ray diffraction analysis No. 1 Cell • Fuel removal from cladding No. 2-1 Cell • Decontamination No. 2-2 Cell • loading , unloading No. 3 Cell • Creep-fatigue test for structural materials MMF-2 No. 4 Cell • Dimensional measurement • Density measurement • Visual examination No. 5 Cell • Uniaxial creep test for structural materials EM Room • Microstructural analysis by transmission electron microscope -50- JAERI-Conf 99-009 LAJt-J LJ LAJ Operation Corridor 1 \ / \ / \ / \ / \ 1 \ / \ Examination Cell Jecontamt- Clean - lation Cell Celt Fxamination Cell Decontami- \ / \ / \ / \ / \ / \ / \ / \ 1 \ 1 \ \ natior Cell Operation Corridor Service Area \i \ i \ i \i \ i \ i \ in Operation Corridor lAHAf Maintenance Room T_OT_ FMF-1 FMF-2 Fig . 1 Layout of the Fuel Monitoring Facility Ventilation and Power Supply I lOm Fig . 2 Layout of the Alplr-Gamma Facility - 5i - JAERI-Conf 99-009 I LAJ> MMF-1 MMF-2 Fig . 3 Layout of the Materials Monitoring Facility AdvancedThermal Reactor "Fugen" I I I Control Rod | R Pressure Tube Fuels Monitoring Facility FMF Assembly Parts, Controll Rod I and Structural Material i J Materii Fuels Materials Fig . 4 Flow Sheet of Irradiated Fuels and Materials -52- JP9950627 JAERI-Conf 99-009 1.6 (a) ACTIVITIES OF OARAI BRANCH IMR OF TOHOKU UNIVERSITY AS AN OPEN FACILITY FOR UTILIZING JMTR Minoru NARUI1, Tsutomu SAGAWA2, and Tatsuo SHIKAMA1 1 Oarai Branch, Institute for Materials Research, Tohoku University, Oarai, Ibaraki, 311-1313 Japan 2 Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki, 311-1394 Japan ABSTRACT Hot cell facilities in a university open facility of Oarai Branch of IMR, Tohoku University (Oarai Branch) will be described briefly, for general aspects and a beta-gamma handling facility. A related topics concerning alpha-gamma facility will be described in a separate paper. Hot cell works in Oarai Branch is related strongly with utilization of Japan materials Testing Reactor (JMTR) in Oarai Research Establishment of Japan Atomic Energy Research Institute. Steady efforts are made to develop advanced techniques for studies of nuclear materials utilizing a JMTR and its related hot cells in Oarai Branch and JAERI. One of examples will be mentioned in the paper. Acquisition of systematic irradiation data is essential for understanding fundamental processes of irradiation effects and for establishment of reliable database of irradiation effects in fusion reactor materials. It will take several years with expensive several different irradiation rigs in a fission reactor irradiation. There, it will take seriously long time to carry out needed iterations between irradiation tests&evaluation and materials developments. An irradiation rig was developed to carry out irradiation in multiple conditions of temperatures and irradiation fluences. Irradiation tests of fusion reactor materials were successfully carried out in JMTR. INTRODUCTION A university open facility of Oarai Branch of Institute for Materials Research in Tohoku University is working for materials study utilizing Japan Materials Testing Reactor in Oarai Research Establishment of Japan Atomic Energy Research Institute. Major playing theaters for promoting researches are -53- JAERI-Conf 99-009 development of radiation techniques and development of post irradiation examination techniques with miniature specimens. Oarai Branch is furnished with two hot laboratories, one for beta-gamma emitting materials and the other for alpha-gamma emitting materials. They are operated under intimate collaboration with the JMTR and its hot cells. The first part of the present paper will describe general architecture of Oarai Branch for study of nuclear materials and show general schema of beta-gamma facilities. The alpha-gamma facility will be described in a separate paper presented in this seminar[l]. The rest of the paper will mainly focus on the newly developed irradiation technique. An appropriate control of irradiation conditions in a fission reactor[2] is needed to obtain reliable database which can be analyzed in comparison with other irradiation data obtained in different irradiation sources such as charged particles irradiation [3]. However, controlled irradiation is expensive and time-consuming in a fission reactor and it will take more than several years to obtain data in irradiation conditions covering needed range of parameters. Also, realization of the same irradiation in different irradiation rigs in different irradiation cycles is not a easy work in a fission reactor, where existence of other irradiation rigs will change a local neutron flux and a gamma heating rate. A irradiation rig which can realize multi irradiation parameters was developed and controlled irradiation of fusion reactor materials was successfully carried out in Japan Materials Testing Reactor (JMTR) in Oarai Research Establishment of Japan Atomic Energy Research Institute. Ten irradiation conditions were realized in one pseudo-shroud-type irradiation rig in JMTR. The present paper describes a detailed structure of developed rig and a history of successful irradiation. Examples of results of analyses of irradiated fusion reactor materials will be found elsewhere [4]. HOT CELL FACILITY IN OARAI BRANCH Interests of researchers in universities in Japan have a wide-range spectrum from development of fission nuclear fuel recycling system and development of new materials for advance nuclear systems such as a nuclear fusion reactors to fumdamental studies such as activation analyses. Various kinds of materials are irradiated in JMTR and they must be handled in hot cells in Oarai Branch. Irradiated capsules are disassembled in a hot cell in the JMTR and subcapsules were transported to Oarai Branch. Hot cell facility in Oarai Branch is separated into two parts, one for handing beta-gamma emitting materials and the other for handling alpha-gamma emitting materials. Fig. 1 shows a layout of Oarai Branch. Irradiated beta-gamma emitting materials are handled in the beta- gamma laboratory in Fig. 1. There, subcapsules are opened and specimens are identified and - 54 - JAERI-Conf 99-009 sorted and they are delivered to each researcher. Fig. 2 shows a layout of the beta-gamma laboratory, which has 6 small lead-shielded hot cells. Researchers prefer to use miniature specimens for restriction of irradiation space and reducing radioactivity. Thus, hot cells are designed to handle these small specimens. Some of material-testings are carried out in the hot cells. Examples are sharpy-tests and tensile tests. Major parts of PIEs are carried out, out of hot cells, as specimens do not have strong radioactivity. Details of hot cells and related works have been already described in the first and second Korean/Japan seminar [1], Figure 1 Layout of Oarai Branch of IMR of Tohoku University - 55 - JAERI-Conf 99-009 Waste Measuring Room Room Service Area Storage Room Isolation Room 1 0 ® 0 Control Loading Dock H Room Steel Cells nn _r Operation Area Lead Cells Changing Room Radiation Management Roon Entrance Office Office Figure 2 Layout of beta-gamma laboratory with 6 hot cells -56- JAERI-Conf 99-009 DEVELOPMENT OF IRRADIATION RIG Figure 3 shows general structures of present irradiation system. Ten small subcapsules were accommodated in a temperature controlled irradiation rig [3], which has independently regulated two temperature zones. The irradiation rig, a protecting tube, a junction box, and a lifting device are forming a psudo-shroud system and the small subcapsules could be lifted up from and inserted down into the temperature controlled rig in the JMTR core, during a reactor operation. Irradiation temperatures were controlled by electric heaters encased in the rig as shown in Fig. 3. Two different temperature zones were set up above and below of the midplane of reactor core, which could be independently controlled irrespective of the reactor power. Figure 4 shows cross sectional view of the rig accommodating the subcapsules. Five transfer tubes were installed in the rig, being thermally bonded by aluminum block. The electric heaters were coiled on the outer surface of the aluminum block, which ensures temperature homogeneity . A reflecting tube and a gap between the reflecting tube and a wall of rig (described as outer tube in Fig. 4) will setup appropriate heat-removal rate for temperature controls. Schematic view of a train of two subcapsules is shown in Fig. 5. Two subcapsules were connected through alumina made thermal insulator and they could be inserted to or removed from the irradiation rig during a reactor operation. Temperature of each subcapsule was monitored by a thermocouple inserted into an aluminum made top cap of the subcapsule shown in Fig. 6. Each of two subcapsules could be irradiated at different temperatures otherwise in nearly the same irradiation conditions. -57- JAERI-Conf 99-009 Capsule control panel Junction box ^ Lifting device \ Lead out tube Vacuum tube F/M Specimen Transfor tube I Insulator TIG / Irradiation rig T/C Electric eater Thermal bond Sub capsule Reflecting tube (T)~@ : Control : Full cycle Figure 3 Schematic view of pseudo-shroud irradiation rig developed forcontrolled fission reactor irradiation. - 58 - JAERI-Conf 99-009 Reflecting tube Electric heater (SUS304,t=0.7) Transfer tube Specimen 72 Thermal bond (A1050) Outer tube Sub capsule (<*> 5.7X04) (<*>40x Figure 4 Cross section of irradiation rig accommodating 5 trains of subcapsules -59- JAERI-Conf 99-009 Wire (SUS304) Sub capsule (A1050) : Control(removal during cycle) : Full cycle Arrangement of Transfer tube Insulator (AQ2O3) Figure 6 Cross sectional view of a subcapsule in a transfer tube. Irradiation temperature is monitored by thermocouple at top cap. )~@ : Control(Removal during cycle) D : Full cycle Arrangement of Transfer tube Figure 5 Outlook of a train of two subcapsuels for tow different temperatures -60- JAERI-Conf 99-009 HISTORY OF IRRADIATION TESTS The present irradiation rig has a potential of temperature control by changing helium gas pressure in a gap between a reflecting tube and outer tube shown in Fig. 4, independently of the reactor power (gas-pressure-temperature control; GPTC). Figure 7 shows one example of temperature control by this GPTC method, without using electric heaters at the reactor shutdown. The gamma heating rate was changing from about 7W/g to less than lW/g, but the temperature of a subcapsule could be kept constant at about 680K. However, the reactor power changed fast by abrupt insertion of stopping control rod at the end of reactor shutdown and the temperature could not be controlled well by the GPTC. In general, the GPTC has advantages over the electric-heater-temperature-control (EHTC). A structure of rig could be simple and more specimens can be irradiated. The most important to point out is that an electric heater has always a possibility of failure. Although electrical heaters could occasionally survive more than one year irradiation in JMTR ( 5 cycles in a year in average), up to more than 1021n/m2 fast (E>lMeV) neutron fluence, their reliable life-time will be a-few-cycles JMTR irradiation. Development of fusion reactor materials are demanding irradiation exceeding 1022 n/m2 of fast neutron fluence. The GPTC will not have a life limitation as far as a geometry of the gas gap is not changed drastically by the swelling. 500 r 400 33 to 300 O I re 100 - 0 50 100 150 Time from start of shutdown (min) Figure 7 Example of temperature control by GPTC at reactor shutdown -61- JAERI-Conf 99-009 38h(9.Q8X1018n/cm)t 92h(2.20X10 n/cm) t 19 186h(4.45X10 n/cm') 354K8.46X10 n/cni) 20 590h(1.41X10 n/cm) 673K 573K (Improved Temperature Control) Control lMay,8(19:50) 50MW ! Control Start (Reactor Power) JUN,2(10:30) Stop QMW OMW T May,8,96(9:57) JUN,2(12:30) Figure 8 Irradiation history using the present rig. Subcapsules were inserted into and removed from the rig during reactor operation. -62- JAERI-Conf 99-009 (X/zCi) Fluence Monitor u 250 500 750 1000 (mm) T3 C [From Core Center] Figure 9 Results of dosimetry of fast neutron fluence as a function of distance from midplane of reactor core. - 63- JAERI-Conf 99-009 However, the GPTC can not control abrupt change of a reactor power in the present system because it takes several minutes to attain equilibrium gas pressure in the gas gap. The basal temperature control was carried out by the GPTC and the EHTC was used for compensating abrupt but small change of temperature especially at the reactor start and shutdown. Figure 8 shows temperature history of irradiation using the present irradiation rig in 1996. The subcapsules were inserted into the irradiation rig after the reactor power and the temperatures of the rig were stabilized as shown in Fig. 8. Then, subcapsules were sequentially removed from the rig during the reactor operation. Irradiation with five different levels of neutron fluence from 9xlO22n/m2 to 1.4xlO24n/m2 at two different temperatures of 573 and 673 K were realized in one exertion of irradiation. A previous preliminary results, however, suggested that specimens were exposed to low flux fast neutron irradiation even when they were not inserted in the reactor core. Figure 9 shows induced radioactivity of iron dosimetry foil as a function of a distance along the height of reactor core. The induced radioactivity is roughly proportional to the fluence of fast neutron. It can be seen that substantial neutron flux exists even in an out-of- core-region. A length of the rig was enlarged as long as possible and the subcapsules were placed at about 300mm above the edge of the reactor core to avoid exposure to the low flux neutrons, where the fast neutron flux is about 10-4 of that at the core center. More than a few thousands TEM (Transmission Electron Microscope) specimens of different fusion candidate alloys were irradiated. Microstructural modifications due to a fission reactor irradiation were examined in comparison with those under other irradiation sources such as high energy electrons in HVEMs (High Voltage Electron Microscope) [4], where extensive data were accumulated and fundamental processes of irradiation induced microstructural evolution was analyzed as a function of a variety of irradiation conditions. CONCLUSION A rig was developed for controlled irradiation in JMTR fission reactor. Irradiation of five different fast neutron fluence at two different temperatures could be carried out successfully using the developed rig. History of irradiation confirmed the well-control of temperature and neutron fluence. References l.M.Narui, et al., The 2nd KAERI-JAERI Joint seminar on PIE technology, Sept. 20-22, 1995, Taejon, Korea. P.242 2. Y.Suzuki et al., presented in this seminar. -64- JAERI-Conf 99-009 3. M.Kiritani et al., J. Nucl. Mater, 191-194 (1992) 100 4. M.Narui et al., J. Nucl. Mater., 212-215(1994)1645 5. N.Yoshida, presented in the ICFRM-8 (Int. Conf. on Fusion Reactor Materials) Sendai, 1997. -65- JP9950628 1.6 (b) VENTILATION SYSTEM OF ACTINIDES HANDLING FACILITY IN OARAI-BRANCH OF TOHOKU UNIVERSITY Yoshimitsu SUZUKI, Makoto fATANABE, Mituo HARA.Tatsuo SHIKAMA, (Hideo KAYANO) .Toshiaki MITSUGASHIRA Oarai-branch, Institute for Materials Research, Tohoku University Narita-machi,Oarai-machi, Higashi ibaraki-gun, Ibaraki, Japan ABSTRACT We have reported the development of the facility for handling actinides in Tohoku University at the second KAERI-JAERI joint seminar on PIE technology. Actinide isotopes have most hazurdous a-radioactivity. Therefore,a specially designed facility is necessary to carry out experimental study for actinide physics and chemistry. In this paper, we will describe the ventilation system and monitoring system for actinide handling facility. INTRODUCTION The Oarai-branch of Institute for Materials Research in Tohoku University has been an open facility to researchers in Japanese universities who want to utilize irradiation service in the Japan Materials Testing Reactor (JMTR) of the Oarai Laboratory of the Japan Atomic Energy Research Institute since its foundation in 1969. To meet increasing interest in studies on actinide physics and chemistry,a new laboratory for actinides experiment(LAE) was constructed during 1987-1991. In order to promote experimental works on actinides research, it is neccessary to insure the health physical protection system for the daily use of alpha-emitters. LABORATORY FOR ACTINIDES EXPERIMENT The LAE is three stories high with one basement and has earthquake-proof structure made of reinforced concrete. Most of the experimental equipment is settled in the first floor. Figure 1 shows a layout of the first floor, i. e., main hole of LAE. Radiation protection area(RPA), shown in bold line, is divided into two zones,a low level zone and a high level zone. The rooms for contamination test, radiometry, physical characterization, and cell operation are in the low level zone and samples enclosed in proper vessels or capsules are treated in these rooms. -66- JAERI-Conf 99-009 Thus, these rooms are regarded as contamination free area. In the rooms of pyrochemistry and solution chemistry, unsealed actinide samples are treated and these rooms are called as umber rooms. Handling of highly radioactive materials is performed in the hot cell and precise experiments are carried out in the rooms of solution chemistry and pyrochemistry by using glove boxes or drafthoods. Mass Flow of Actinide Research A small quantity of TRU has been produced from commercially available actinides such as U,Np,and Am by neutron irradiation in JMTR. Materiales for the irradiation are also prepared in LAE. The materials are encased in a sealed subcapsule usually made of fused silica which is then encapsulated into a stainless steel clad for a short term irradiation in hydraoulic rabbit or a stainless steel capsule-rig for a long term irrsdiation. Every enclosure is examined by X-ray autoradiography before irradiation and transported to JMTR. After the irradiation, the rabbit or the rig is cooled in a water pool in JMTR to wait for the decay of short-lived induced radioactivities. Then, the subcapsule is retrieved in a hot cell in JMTR and transported to LAE by using a container applied with thick lead shield. The container is transported into LAE through the loading dock and settled onto the transfer port, named gamma-gate, of storage cell by using a 10 tons crane. The gamma-gate has an air tight structure applied with 250 mm thick lead shield and the irradiated subcapsule can be transferred into the storage cell. The irradiated actinide containning specimen is retrieved by using a cutter in the storage cell and then transferred to the working cell through a transfer port which consists of a pair of double-sealed and air-leak-tight doors on each side of the two cells. The chemical separation of TRU is carried out in the working cell and the isolated TRU is transferred to the cell glove box attached on the side wall of the working cell. Then, TRU thus isolated may contain impurities of other actinides and lanthanides. The final purification to remove these impuritiesis carried out in the room for solution chemistry by using a glove box or a draft-hood. Sometimes, commercially obtained actinides contain impurities. The removal of these impurities is also performed by using these equipments. Specification of Hot Cell Figure 2 illustrates the structure of the hot cell. The cell shield is made of 350 mm thick SS-41 steel and the inside of the shield is covered with polished lining made of 4mm thick SUS 316L plates. Each SUS 316L plate was welded together to make air-leak-tight structure of the two cells. Seven air-tight manipulators (Sargent Model-L), five lead glass windows,250 mm thick lead equivalent, and a control desk are attached in the front wall of the two cells. Major experiment in the working cell can be carried out by using four manipulators inspecting the - 67 - JAERI-Conf 99-009 inside condition through one of the three windows. The setting of the experimental apparatus inside the cell can be made by using a glove port attached at the rear side of SUS lining wall. Many feed-throughs are available to get easy access for electricities, gases and chemical reagents. An isolation room is provided for the maintenance and decontamination of apparatus settled in the cells. The room also has air-tight structure and the maintenance works will be done appling a tunnel suit attached to the tunnel port. These maintenance works will be supported by supporters in service room. Air in the isolation room is taken from hot area through a gallery attached at the wall of service room and transported through one of the two overhanged HEPA filter cases. Another HEPA filter case is equipped at the rear wall of the cell shield to transport the air into the cells. Air ducts are attached to each filter case to insure the down blow of the air. Radiometry [J Physical L Characterization Figure 1. Layout of the LAE -68- JAERI-Conf 99-009 MAINTENANCE GLOVEBOX FOR MANIPUtATORS [SERVICE ROOMl URTtGHTDOOR TUNNEL PORT WASTE PORT I IISOLAT1ON ROOMl WALL UNING (SS41 350mm) DOOR (SUS-316L 4mm) TUNNEL PORT \GLOVE POR Figure 2. The structure of the hot cell For the cleaning of the exhaust from the cell, cell filters are also equipped and the exhaust is treated finally by using a filter assembly of hot cell. The filter assembly is consisted from two ULPA filters and one chacorl filter. The ULPA filter can remove 99. 999% of airborne radioactive particles of 0.0001 mm in diameter. Leak-tight automatic-valves are attached to the cell ventilation system. These assemblies can sustain a pressure difference upto 500 mmAq. The measured leak rates are less than 0. 01 volX/h and less than 0. 007 volVh for the cells and cell glove box, respectively. -69- JAERI-Conf 99-009 Maximum Limits of Main Actinide Nuclides for Daily Use Table 1 shows maximum permissible amount of main actinide nuclides which can be used in LAE. To keep maximum freedom and to minimize the official inspection due to NPT, the amount of Pu is restricted to be less than lg/y. The actinide elements which can be used in substantial study of solid state physics and chemistry are Th, U, Np, and Am. The conceptual design of health physical protection system and monitoring system is made to sustain the experimental use of these nuclides. Table 1. Maximum limits of main actinide for lailc y use Nuclide In hot cell In glove box 226Ra 400 GBq 10 MBq 2 3 2^ (500 g) (1,000 g) irrad. Th 750 GBq (100 g) 10 MBq ( 50 g) 231Pa 20 GBq ( U.4 g) 50 MBq 233U (100 mg) ( 40 mg) 2S5JJ 5 GBq 500 MBq 238U (500 g) (1 000 g) irrad. U 750 GBq (100 g) 10 MBq ( 50 g) EU«20X) ( 15 g) ( 5 g) HUO90X) ( 50 mg) ( 20 mg) 237Np 5 GBq (191.52 g) 200 MBq 2 3 8pu 20 GBq 20 GBq 2 3 9pu (200 mg) ( 100 rag) 24 'Am 200 GBq ( 1. 57 g) 5 GBq 243Am 40 GBq ( 5. 41 g) 5 GBq 24ZCm 500 GBq ( 4. 07mg) 100 MBq 243Cm 200 GBq (104.61 rag) 500 KBq 244Cm 100 GBq ( 33. 35mg) 100 KBq 248Cm 50 MBq (327. 78mg) 100 KBq 251Cf 2 GBq ( 34.12mg) 40 MBq 252Cf 50 MBq ( 2. 5 Hg) 100 KBq 253Es 20 GBq ( 0. 8 fg) 200 MBq Ventilation System of LAE LAE has maximum 100,000 m3/h once through ventilation system, and usually operate at about 50,000 m3/h. Inlet air is taken through role type dust filters and salt removing filters by using three ventilators settled in the cold-ventilator room at the third floor of LAE. The inlet air is introduced into RPA through three air lines,low level line,hot area line,and umber line. The air in the low level area - 70 - JAERI-Conf 99-009 is passed into the hot area and a small portion of the air in the hot area is introduced into cell lines as explained in former section. The exhaust from the hot area and umber rooms is transported to one of the 9 filter cases, each of which is consisted with a prefilter and two stage of ULPA filter that has the ability removing 99.999% of 0.1// particle. The air in umber room is partially introduced into draft hoods which are attached with two stage alkaline scrubers. The exhaust from these scrubers is also transferred into the same filter cases. All the filter cases and fans for exhaust are settled in the basement floor. The outline of these ventilation system is shown in Figure 3. Pneumatically controlled valves and dumpers are attached on each ventilation line and the pressure in the RPE is always controlled below atmospheric level. The conditions of air flow as well as the status of valve operation can be seen in a display board settled in the control and monitoring room. All exhaust is released from 40 m tall stack after the monitoring of alpha and beta-gamma radioactivites contained. o;u GLOVE BOXES I :::-•»••::. I f K u :y -e- LAB -e- AIR EA INLET R:ROLL FILTER WASTE S:SALT CLEAR FILTER P:ME FILTER C:CHARCOAL FILTER STORAGE -J H:HEPA FILTER U:ULPA FILTER HOT -e- AREA P!H I .•:•••••*»-••••. l P :.' CELL 1 t v •• CELL 2 P:'c -©- WASTE -e- AIR WATER INLET EXHAUST P-H Figure 3. The outline of ventilation system - 71 - JAERI-Conf 99-009 MONITORING SYSTEM OF LAE The monitoring system of LAE is composed of in-cell monitors,exposure rate monitors for RPA, air line monitors,and exhaust air monitors. The specifications of these monitors are shown in Figure 4. All monitors are connected to a large display board settled in the control and monitoring room. Variable alarm level setting is possible for each monitor from the display board. Once the alarm is raised, inter-lock protection, such as the shut-down of the fans and alarm rings for workers, are automatically triggered. In-Cell Monitor and Exposure Rate Monitor Each cell is equipped with an ionization chamber to inspect the exposure rate. If it exceeds the alarm level, inter-lock system is triggered to prohibit opening shielding doors and the maintenance works in the isolation room. It is neccessary to cancel the alarm condition by using manipulators. Usually, the maximum of the exposure rate in the RPA is controlled to be lower than 0. 001 mSv/h. The exposure rate monitors are settled in high level zone and continuous monitoring is made regardless of experimental works. Air Line Monitor A SSB detector, a Nal scintillation detector and a GM counter are applied to the air line sampling system in LAE to monitor continuously the dust trapped on the surface of a filter. The signals from the SSB are recorded in an analyser which calculates the real alpha contents subtracting radon daughter contributions. In spite of this substraction ability, back-ground level of the alpha counts from SSB changes according to the condition of ventilation system. Thus,the filter is exchanged daily and the alpha-counts are measured separately by using a- spectrometer in order to confirm the contamination free atmosphere in LAE. Monitors for Exhaust The exhaust which passes through ULPA filter is bypassed to the monitoring system. The monitoring system for exhaust is equipped with five detectors and a dust sampler as shown in Table 2. This monitoring system is very important to ensure the non-release of radioactive substances into environment. Two ZnS detectors are applied to measure the contents of alpha radioactivity in the dust continuously collected on the role-type filter sampler. One is used to measure the dust just after the sampling and the other is used to measure about 24 hours later to wait for the decay of the signals due to 222Rn daughters. Actualy, the sensitivity about -72- JAERI-Conf 99-009 MONITORING SYSTEM OF LAE a, 0, 0 • 7 ST A C K D U ST •G A S MO N IT OR CONTROL a, 0 ' 7 PANEL RO OM DU ST •G A S MON IT 0 R 7 j DATA ; AR E A MO N IT 0 R i TREATMENT: 7 1 DISPLAY 1 I N C E LL M 0 N IT 0 R : & • •' PRINT-0UT1 a, 0, 7 WA ST E WA T ER MO N I T 0 R Figure 4 Monitoring system of LAE STACK DUST-GAS SAMPLER SAMPLER DETECTOR RADIOACTIV RAY MEASURING PROCEDURE Zn S a 1 JUST AFTER SAMPLING DUST Zn S a 2 1 DAY AFTER SAMPLING PLAST I C 0'7 JUST AFTER SAMPLING I ON I ZAT I ON 0 REAL TIME, CONTINUOUSLY GAS C HAMBER Na I 0'7 IODINE REAL TIME, CONTINUOUSLY Table 2 Stack dust-gas sampler - 73 - JAERI-Conf 99-009 10 Bq/cm3 is attained by using this system. Main component of these alpha- counts is due to the daughters of220 Rn which may be generated in the concrete of LAE and fluctuates according to the humidity. Monitoring of Waste Water In LAE, the washing for cleaning experimental tools is prohibited. Thus, the water mainly comes from scrubers. All waste water is stored in one of the four tanks, each has a capacity of 4 m3. The monitors of waste water are equipped but have never been operated, because the total volume of the waste water generated during these 7 years is less than 3 m3. Instead, sample is taken periodically into a 1, 000 cm3 marineri-beaker and gamma-ray spectrum is inspected. After the inspection of the gamma-ray spectrum,alpha-contents are measured by using SmF3 coprecipitation technique or liquid scintillation spectrometry combined with pulse shape discriminator. We have developed a new type extracting scintillator for the selective extraction of TRU. The main component of the extracting scintillator is 20 vol% TBP-toluene, which contains 0. 02g/cm3 of naphthalene and 4 mg/cm3 of PBBO. The water sample is mixed with calcium nitrate tetrahydrate to be the IhO/Ca ratio less than 8 and the salt solution is shaken with an equal volume of the extracting scintillator. All TRUs can be extracted quantitatively and the organic phase can be applied to the accumulation of alpha-spectrum. Reference Suzuki, Y. ;Hara, M. ;Shikama,T. ;Mitsugashira, T. ;Kayano, H. 1990:Application for Permission of Uses of Radioactive Materials and Nuclear Fuel Materials to the Science and Technology Agency of Japanese gaverment. Suzuki, Y. ;0chiai,A. ;Shikama, T. ;Mitsugashira, T. ;Kayano, H. 1995: Development of Facility for Handling Actinides in Tohoku University. KAERI-NEMAC/TR-32, 90-100 - 74 - JP9950629 JAERI-Conf 99-009 1.7 PIE ACTIVITIES IN NFD HOT LABORATORY Norikatsu Yokota, Keizo Ogata and Noriyuki Sakaguchi Nippon Nuclear Fuel Development Co., Ltd. 2163 Oarai-machi Narita-cho, Higashi-Ibaraki-ken Ibaraki-ken, 311-1313, Japan ABSTRACT Nippon Nuclear Fuel Development Co., Ltd. (NFD) has been operating hot laboratory facility since 1977 for post-irradiation examinations (PIE) of boiling water reactor (BWR) fuels and structural materials. Various examination techniques have been developed to meet the research requirements. The BWR fuel design, which has been revised for a step-by-step burnup extension, has been verified at each step through comprehensive PIEs. A large number of fuels and materials have been examined in various research and development programs. High burnup fuel pellets were extensively examined in terms of fission gas behavior and microstructural evolution. Cladding waterside corrosion performances were studied from a viewpoint of the base metal metallurgical conditions. An electro-chemical technique was applied for determining oxide film characteristics. Reactor core structural materials have also been studied for plant life extension and development of remedies. INTRODUCTION Nippon Nuclear Fuel Development Co., Ltd. (NFD), a subsidiary of Hitachi and Toshiba, has been operating hot laboratory facility since 1977 for extensive post-irradiation examinations (PIE) of boiling water reactor (BWR) fuels and structural materials. NFD hot laboratory has a capability to accomodate full size commercial BWR fuel bundles. Comprehensive PIE programs have been carried out on many BWR fuel bundles [1-4] including failed fuel bundles [5] and MOX fuel bundles[6], as well as test fuel and material specimens irradiated in test reactors. To meet the demands for detailed mechanistic understanding of fuel -75- JAERI-Conf 99-009 performance, more precise microscopic and specific PIE techniques are required. Various examination techniques have been developed during the course of the PIE programs to meet the research requirements. This paper presents the overview of recent PIE activities in NFD hot laboratory. OUTLINE OF THE NFD HOT LABORATORY Layout of the NFD hot laboratory is shown in Fig. 1. It consists of a storage pool, an inspection pool, six concrete cells, six steel shielded cells, two waste storage cells, and an isolation area. It has a capability to accomodate full size commercial BWR fuel bundles. Standard PIE procedures include non-destructive examinations followed by destructive ones. A window is installed in the side wall of the inspection pool so that the fuel bundle can be observed directly. Non-destructive examinations on fuel bundle and fuel rods can be performed in the inspection pool and the monitoring cell. Concrete cells, shielded by concrete walls lined with stainless steel sheets for easy decontamination, are used for destructive examinations such as mtallography and ceramography. Steel shilede cells annexed to the concrete cells are used for the tests on relatively small specimens, such as mechanical testings. Various kinds of microscopic equipment, including transmission electron microscope (TEM), scanning electron microscope (SEM), electron probe micro- analyzer (EPMA), mass spectrometer, and X-ray diffractometer (XRD), are installed in precise measurement labs. Special apparatus for specific examinations on irradiated materials, such as thermal diffusivity measurement by laser flash method, post-irradiation annealing, etc., are also installed there. OVERVIEW OF PIE ACTIVITIES 1. Research and development trends for BWR fuel Burnup extension with the design changes followed in Japanese BWRs is shown in Fig. 2[7]. Following the research activities in the early use of fuel with the focus mainly placed on enhancing the reliability and capacity factor, one of the largest technical efforts during last decade has been burnup extension aiming at reduction of spent fuel amount and improvement of the fuel cycle economy. The BWR fuel design, which has been revised for a step-by-step burnup -76- JAERI-Conf 99-009 extension, has been verified at each step through comprehensive PIEs[l-4] in terms of the required performance and design margins. A large number of fuels and materials irradiated in commercial reactors as well as those in test reactors have been examined in various research and development programs. Significant amounts of data have been accumulated especially on high burnup phenomena. Utilization of plutonium uranium mixed oxide (MOX) fuel in light water reactors is another important issue in Japan. Programs for the MOX fuel have also been carried out including PIEs on MOX fuel bundles irradiated in a commercial BWR and test fuel irradiations in test re actors [6,8,9]. 2. PIEs on fuel pellet andFP'behavior Fission product (FP) gas release and FP gas bubble swelling are of practical importance in determining high burnup fuel performance. Since such phenomena are closely related to the fuel microstructure, irradiation-induced microstructural evolution is also of major interest at high burnup. FP gas release rate of irradiated fuel rods can be measured by both destructive (puncture test) and non-destructive methods. The latter utilizes the measurement of 85Kr gamma activity at the rod plenum. Some FP gas release data[4] are shown in Fig. 3. Fuel rod axial distribution of FPs is measured by gamma scanning in a hot cell. Radial distributions in a fuel pellet cross section are examined with micro-gamma scanning device, EPMA, and ion micro-analyzer (IMA). A large amount of data has been accumulated in the fuel performance database to support development of a fuel performance analysis code which employs mechanistic models of FP gas behavior. FP gas behavior in a fuel pellet was studied using post-irradiation annealing technique [10]. A small specimen of irradiated fuel pellet was heated in the high temperature furnace and the released 85Kr was measured by gamma activity. Typical results are shown in Fig. 4 demonstrating burst FP gas release at some temperature. Further, high temperature and high pressure furnace has been developed[ll] to study FP gas behavior under various restraint pressures simulating restraint force by pellet-cladding interaction (PCI) which could suppress FP gas bubble growth and its release. Fig. 5 shows some results of de- pressurization experiments under isothermal annealing at 1500°C to simulate a rapid removal of PCI restraint. A large burst release was observed immediately after the de-pressurization. Such results are consistent with the irradiation data obtained in power bump tests. Microstructural evolution in high burnup UO2 and U02-Gd20.3 fuel pellets, such as formation of rim structure and pellet-cladding bonding, were extensively examined by detailed PIE techniques utilizing EPMA, XRD, SEM, and TEM[12]. - 77 - JAERI-Conf 99-009 Typical ceramograph, and SEM and TEM images of fuel pellet irradiated to 60GWd/t are shown in Fig. 6[13]. Sub-divided grain structure (rim structure) was seen at the pellet periphery. Its formation mechanism was studied from the detailed examinations as shown in Fig. 7 [14]. Tangled dislocations were organized into sub-grains of 20-30nm in size with high angle boundaries which were regarded as the nuclei for recrystallization resulting in sweeping out of FP gas atoms into small intragranular bubbles. Regions of the pellet-cladding bonding layer of the high burnup fuel above 40GWd/t were examined in detail[15]. AZrO2 layer consisting of cubic polycrystals was found near the cladding inner surface. In a region near the UO2 pellet surface, both a cubic solid solution of (U,Zr)O2 and an amorphous phase existed. The derived formation mechanism of bonding layer is summarized in Fig. 8. A key process is the phase transformation of ZrO2 film from monoclinic phase which is stable at temperatures below 1170°C to cubic phase. This observed phase change is attributed to fission damage. 3. PIEs on fuel cladding materials Cladding waterside corrosion and associated hydrogen pickup are also important issues at high burnup. Oxide thicknesses of the cladding outer surface are measured non-destructively by eddy current method and destructively by metallography. Typical data of oxide thickness are shown in Fig. 9[16]. Fuel rods of recent step II design showed a remarkable improvement in corrosion resistance. Hydrogen contents are measured by hot vacuum extraction method. Corrosion performances of irradiated Zircaloy-2 and other zirconium alloys were studied from a viewpoint of the base metal metallurgical conditions utilizing SEM and TEM[16-18]. Fig. 10 shows schematic illustrations indicating the irradiation-induced change of precipitates in Zircaloy-2 cladding tubes. Zr(Fe,Cr)2 type precipitate in irradiated Zircaloy-2 undergoes crystalline-to-amorphous transition and has Fe-depletion in the amorphous region, i.e. dissolution of Fe atoms from the precipitate to the matrix. Zr2(Fe,Ni) type precipitate also changes its shape and both Fe and Ni diffuse away into matrix. Such radiation-induced dissolution was modeled taking the precipitate size into consideration and calculated as shown in Fig. 11. Dissolution rates of various Zircaloy-2 cladding tubes were estimated and relationship with their corrosion behaviors is also shown in Fig. 11. The dissolution rate has a clear effect on the corrosion performance. An electro-chemical technique was applied for determining oxide film characteristics. The electric resistances of oxide films formed on various Zr alloys in out-of-pile steam and in BWRs were measured utilizing an AC impedance device shown in Fig. 12[19]. Some results are shown in Fig. 13. Impedance - 78 - JAERI-Conf 99-009 responses differed remarkably between pre- and post-transition oxide films, although TEM observations gave no clear difference in crystal shapes and sizes. The impedance measurement results on various Zr alloys irradiated in a BWR demonstrated that the alloys whose oxide had a lower electronic conductivity showed a better corrosion performance. Mechanical properties of cladding tube are not supposed to significantly change at high burnup, but they should be watched in relation to oxide buildup and hydrogen pickup. Tensile, internal pressurization burst, fatigue and charpy tests are performed in hot cells. 4, PIEs on BWR core structural ma terials Reactor core structural materials have also been studied for plant life extention and development of remedies. Slow strain rate tensile and uniaxial constant load tests have been performed on irradiated specimens under simulated BWR conditions. Microstructures of irradiated stainless steels were examined utilizing TEM equipped with field emission electron gun (FE-TEM) which enabled to analyze microscopic area with the resolution of 1 nm to investigate irradiation- induced defects and segregation affecting the mechanical and chemical behaviors such as irradiation assisted stress corrosion cracking (IASCC). SUMMARY The major interest of BWR fuel R&D has shifted to burnup extension. To meet the demands for detailed mechanistic understanding of fuel performance, especially at high burnup, more precise microscopic and specific PIE techniques are required. Various examination techniques have been developed in NFD during the course of the PIE programs to meet the research requirements. A large number of fuels and materials irradiated in commercial reactors as well as those in test reactors have been examined in various research and development programs. Detailed PIEs on fuel pellets, cladding materials, and core structural materials have been successfully carried out and significant amounts of data have been accumulated. -79- JAERI-Conf 99-009 REFERENCES 1. Y. Mishima and T. Aoki, Proving Test on the Reliability of BWR 8X8 Fuel Assemblies, Paper Presented at IAEA International Symposium on Improvements in Water Reactor Fuel Technology and Utilization, IAEA-SM- 288/58, Stockholm, Sept. 15-19 (1986). 2. M. Oishi, High Burnup Fuel Behavior Studies in NUPEC, Paper Presented at IAEA Technical Committee Meeting on Fuel Performance at High Burnup for Water Reactors, Nykoping (1990). 3. H. Ohara, et al., Fuel Behavior During Power Ramp Tests, Proc. International Topical Meeting on Light Water Reactor Fuel Performance, West Palm Beach, April 17-21 (1994). 4. H. Hayashi, et al., Irradiation Characteristics of BWR Step n Lead Use Assemblies, Proc. 1997 International Topical Meeting on Light Water Reactor Fuel Performance, Portland, May 2-6 (1997). 5. K. Ogata, et al., Post Irradiation Examination on the Failed Fuel Rod in the Hamaoka Atomic Power Station Unit 1, Proc. International Topical Meeting on Light Water Reactor Fuel Performance, West Palm Beach, April 17-21 (1994). 6. M. Ichikawa, et al., J. At. Energy Soc. Japan, 39, 2 (1997) pp.93-111 (in Japanese). 7. K. Ogata, et al., BWR Fuel Performance and Recent R&D Activities in Japan, Proc. 1997 International Topical Meeting on Light Water Reactor Fuel Performance, Portland, May 2-6 (1997). 8. M. Oguma, et al., Technology Development for Japanese BWR-MOX Fuel Utilization, Paper Presented at IAEA Technical Committee Meeting on Recycling of Plutonium and Uranium in Water Reactor Fuel, Newby Bridge, July 3-7 (1995). 9. K. Asahi, et.al., Irradiation and Post Irradiation Testing Program of BWR MOX Fuel Rods, Proc. International Topical Meeting on Light Water Reactor Fuel Performance, West Palm Beach, April 17-21 (1994). 10. K. Une, et al., J. Nucl. Sci. Technol., 27, 11 (1990) p. 1002. 11. S. Kashibe and K. Une, Effect of External Restraint on Density Change and Fission Gas Release in UO2 Fuels, Paper Presented at Enlarged Halden Programme Group Meeting, Lillehammer, March (1998). 12. K. Une, et al., Effect of Irradiation-Induced Microstructural Evolution on High Burnup Fuel Behavior, Proc. 1997 International Topical Meeting on Light Water Reactor Fuel Performance, Portland, May 2-6 (1997). 13. M. Hirai, et al., Performance of Improved UO2 Pellets at High Burnup, ibid. 14. K. Nogita and K. Une, J. Nucl. Mater., 226 (1995) p.302. -80- JAERI-Conf 99-009 15. K. Nogita and K. Une, Formation Process of Pellet-Clad Bonding Layer in High Burnup BWR Fuel, Paper Presented at IAEA Technical Committee Meeting on Advances in Pellet Technology for Improved Performance at High Burnup, Tokyo, Oct. 28-Nov. 1 (1996). 16. Y. Etoh and S. Shimada, J. Nucl. Sci. Technol., 29, 4 (1992) p.358. 17. Y. Etoh, et al., J. Nucl. Mater., 200 (1993) p.59. 18. Y. Etoh, et al., The Effect of Microstructure on Corrosion Behaviors of Zry-2 in BWRs, Twelfth International Symposium on Zirconium in the Nuclear Industry, Toronto, June 15-18 (1998). 19. S. Nanikawa, et al., Correlation Between Characteristics of Oxide Films Formed on Zr Alloys in BWRs and Corrosion Performance, Twelfth International Symposium on Zirconium in the Nuclear Industry, Toronto, June 15-18 (1998). 81- JAERI-Conf 99-009 Cross sectional view Inspection window Concrete cells Pool Precise measurement labs. Waste storaae area • I I chemicat labJS Steel shielded cells Grand floor olan Fig. 1 Layout of the NFD hot laboratory -82- JAERI-Conf 99-009 60 ; 3OOO0O0OO aoooooooo O Water rod ooooooooo ooooooooo * : Small number of step III LUAs ooooooooo ooooooooo T3 OOOO^NOOO OOOf \OOO were loaded in 1996 Qocy-wooo oool looo ©ookjooc - ioL_Jooo CD oooooooo ooooooooo ooooooooo 50 ooooooooo ooooooooo D. oooooooo oooooooo POOOOOOOQi POOOOOOOQ 9X9 (Type A) 9x9 (Type B) C i_ DOOOOOOO ooeUooo 1 D oooooooo STEP-IE (45)* " oooooooo oooooooo OOOOQOOO 40 OOOOOOOO OOOOOOOQ STEP- n (39.5) oooOoooo cd oooooooo o oooooooo STEP-1 (8 x 8BJX33) OOOOOOOQ 8x8RJ(29.5) I © 30 7x7R(27.5) cd 8x8(27.5) 7x7(21.5) < 1970 1980 1990 2000 Year of loading Fig. 2 Change of average bundle burnups at discharge 50 • 1 1 • A,V: conventional 8X8 D :HPF • 40 • O :Step! LUA a) Black symbol;(U,Gd)O2 w -S> 30 a: to (0 A A| A O 20 • 1I A A ANT g w A OT A A ^ A A 10 M A • D 10 20 30 40 50 60 Rod Burnup (GWd/t) Fig. 3 Burnup dependence of fission gas release -83- qq h Restraint pressure (Mpa) era at 00 00 NJ 6o o o O O en 00 w CD P3 CD GO P Temperature (°C) CD Kr-85 activity (Arbitary unit) GO O CD 3 GO Hi 5" era h GO 0 k CD O O > CD en t3 2 B CD n to Hi o H 0 P 3 3_ 3 0 3 CD CD al i CD O O f 00 3 0 era "—'0 "73 0 t3 -p»- O p 0 CO T3 0 & HJ CD p en 13 0 P CD 0 ->• -»• ro GO ST. ^^ o> o GO o o o o C o o o HJ Fractional release (%) 3 P N 3 Temperature (°C) CD O P 3 tn 3 oq JAERI-Conf 99-009 , ( 'ii d subble , ;tal!izing rngion 'T?V^s-f Radiation damage 'of- 3a accumulating region | 200nm Ceramograph SEM images TEM image of the rim region Fig. 6 SEM and TEM images of fuel pellet irradiated to 60GWd/t Point defects Cladding inner surface oxidation Interstitials Vacancies (Monoclinic ZrOa) Fission damage —•( RecombinationJ in ZrO2 Phase transformation of Dislocations Bubbles ZrO2 Monoclinic-*Cubic Sub-divided grains ^~*\ closure J L Strong contact of cubic ZrO2 and UO2. Recrystallization Formation of (U,Zr)O2 solid soliution Recrystallized grains Coarsened bubbles Rim structure ing layer Fig. 7 Formation mechanism of the Fig. 8 Formation mechanism of the rim structure bonding layer - 85 - JAERI-Conf 99-009 20 30 40 50 60 Local burn up (GWd/t) Fig. 9 Maximum oxide thicknesses on three types of Zircaloy-2 cladding tubes irradiated in BWEs TEM images TEM images Fe I Low o 15 i fluence 'a. 'o ' "o I Q. ^ Crystal 03 | Medium Fe Ni Fe I fluence T3 C 03 /Amorphous c o High fluence b Zr-Fe-Cr type precipitates Zr-Fe-Ni type precipitates Fig. 10 Schematic illustrations indicating the irradiation-induced change of precipitates in Zircaloy-2 cladding tubes 86- JAERI-Conf 99-009 1°.4 Particle size : 0.25 urn Measured (a) Dissolution rate ' Calculated as a function of U 0.2 neutron fluence •£ 4 6 8 10 12 14 Fast neutron fluence (E>1MeV, x 1025 n/m2) (b) Relationship between oxide thickness and dissolution rate 10-9 10-8 Initial dissolution rate of Fe+Ni (at%/s) Fig. 11 Dissolution rate of Fe from precipitates in Zircaloy-2 and relationship between oxide thickness and dissolution rate of Fe+Ni Working electrode (Specimen) Lock-in amplifier Reference T electrode Potentiostat (Ag/AgCI) \ T Personal computer Counter electrode (Pt) Control unit Fig. 12 Block diagram of the AC impedance measurement system -87- • Zircaloy-2 • Pre-transition A High FeNi/Zry-2 • Post-transition O 0.5Nb/Zry-2 10 10 10 10 • i a a B CD 8 CD 10 OL 10 — a. D) CD 13 o O 10° CD 106 CD to Li k x— 10* 104 2 >rn 10' 10 N 2 0 n 0) 0) o 3 CD CD • -30 Z3 -30 6 > o cb~ % CD" 1 ^0O O 2 • ^^ 3 % CD -60 CQ -60 1 CD CD CD Jll I, 1 -90 -90 10" 10u 10' 10" 10"4 10* 10u 10' 10" Frequency (Hz) Frequency (Hz) (a) Unirradiated Zry-2 (b) Irradiated for 4 cycles (Oxide film formed out-of-teactor) (Oxide film formed in-reactor) Fig. 13 Typical results of impedance measurements JAERI-Conf 99-009 JP9950630 1.8 Current Status of NDC Fuel Hot Laboratory Youichirou YAMAGUCHI1, Takanori MATSUOKA1, Satoshi SHIRAISHI1 Mitsuteru SUGANO2 1 .Hot Laboratory Experiment Department, Nuclear Development Corporation Tokai-mura, Naka-gun, Ibaraki-ken, Japan 2:Nuclear Fuel Engineering Department, Mitsubishi Heavy Industries, Ltd. Wadasaki-cho, Hyogo-ku, Kobe, Japan ABSTRACT Nuclear Development Corporation(NDC) fuel hot laboratory was established to investigate the causes for leaked rods and to confirm the integrity of the precedence irradiation fuels, in 1986. After that, it obtained a license to conduct PIE of the structural materials, such as stainless steels, inside the reactor in addition to fuels. So far we have conducted PEE of fuels and metallic materials including fuel assembly components and reactor internal components irradiated in Japanese PWR plants or some test reactors. To meet these PIE needs, we are making efforts to improve facilities and to install high advanced equipments. This paper describes current status of the facilities and PIE techniques in NDC fuel hot laboratory. INTRODUCTION As hot laboratories to conduct PIE, NDC is equipped with the fuel hot laboratory and the material hot laboratory. Each has /? - y cells, and we can handle nuclear fuel substances in only fuel hot laboratory. The fuel hot laboratory has been established to investigate the causes for leaked rods, to confirm the integrity of the precedence irradiation fuel, and so on. In recent years, fuel leakage does not occur in the domestic PWR plants. The current major works are the PIE related to research and development of high burnup fuels and reactor internal component materials against the background of PWR plant life extension. Implementation of PIE for about lOyears caused increase in the volume of waste generated in the fuel hot laboratory, and this resulted in -89- JAERI-Conf 99-009 additional installation of a waste storage warehouse in 1996. In the material hot laboratory, PIE for the surveillance test in the PWR nuclear reactor vessel steel has been conducted since 1972. This laboratory has mainly been used in mechanical tests such as Charpy impact test and tensile test. The material hot laboratory is also used to conduct a performance test of charcoal filter used in the nuclear power plant. In addition to these hot laboratories, NDC has the fuel test facility to manufacture of unirradiated pellets on a trial basis and to perform initial characteristics tests, and the facility to test the unirradiated fuel assembly structure. This paper describes the equipments and PIE techniques paying attention to the fuel hot laboratory of these facilities. 1. Current Status of the fuel hot laboratory As shown in the layout of Fig 1, the fuel hot laboratory has a pool, six cells and two instrumental analysis rooms. Table 1 shows the cell specification, and Fig. 2 shows the PIE flow classified according to test sites. (l)Pool Almost all the PIE samples are supplied from the PWR plants or test reactors. A fuel assembly or fuel rods are transported in the cask called MSF-1 which contains one fuel assembly, as illustrated in Fig 3. The fuel assembly or fuel rods are replaced from the transport cask to the specified position within the pool by the pit crane. In the case of fuel assembly, the fuel rods to be tested is removed from the assembly after the top nozzle is removed by electrical discharge machining in the pool. Then the fuel rods are loaded into No.l cell for PIE. In addition, the pool is also used to store assemblies and fuel rods as well as high level waste. It is also possible that the high level waste, which is shielded by dry cask, is stored in the warehouse on the ground except the pool. The pool is also be used for appearance test and oxide film thickness measurement. Layout of the pool is shown in Fig.4. (2)No.l cell No. 1 cell is used for non-destructive test of fuel rods. About 10mm length from the upper end of the fuel rod is clamped, and inspection is carried out in the vertical direction. So three fuel rod drive units that go up or down the distance of 4m are located in this cell. The accuracy in the operating distance, when the fuel rod is raised 4m, is within 0.5 mm. An example of the appearance of this drive unit is shown in Fig5. Appearance tests have been carried out by photographing, but this is being switched over to photographing by digital camera in recent years. This makes a significant contribution to speed up. Eddy current test and dimensions measurement have been made by entering -90- JAERI-Conf 99-009 data into the computer to facilitate subsequent data processing. (3)No.2 cell No.2 cell is utilized for preparation of the sample such as cutting of fuel rods, removing of pellets and reassembling of rods, and sealing of waste into the can. Complicated machining including not only cutting but also grinding and electric discharge machining has been performed. For example, such work includes preparation of tensile samples provided with parallel section because high burnup fuels involve difficult work of removing pellets and preparation of the 3mm-diameter disk sample to measure thermal diffusivity rate for the pellet. The schematic diagram of these machining method and procedure are shown in Fig. 6. (4)No.3 cell No.3 cell is mainly used to conduct the tensile test of the fuel cladding tube and internal pressure burst test or the like. Stress corrosion cracking(SCC) test for non-fuel is also performed in this cell. To perform SCC test in the space within the limited cell, autoclaves under the worktable are installed. The sample pulling mechanisms mounted on the top of the autoclaves are removed at normal times, and is installed for test, thereby ensuring effective use of the space in the cell. Atmosphere of SCC test is the PWR primary simulate water in high pressure and high temperature and it is circulated through the water quality regulator installed out of the cell and inside cell. The appearance of the SCC devices is shown in Fig. 7. (5)No.4 cell No.4 cell is used for the sample preparation of ceramography and metallography and for chemical examination such as the pellet dissolution characteristic tests. (6)No.5 cell No.5 cell is utilized to observe the samples polished and etched in No.4 cell. There are the optical microscopes and the hardness tester in this cell. Negative film was used to take a photograph. Similarly to the case of appearance test in No.l cell, digital camera has been frequently used to photographing in recent years. The data sampled by this are stored in the computer in preparation for immediate transmission to any other places via the network line. (7)First instrumental analysis room There are the hydrogen analyzer, the thermal ionization mass spectrometer(TI-MS) and the field emission transmission electron microscopy (FE-TEM) in this room. TI-MS and FE-TEM have not shields as very small or diluted samples are handled there. TI-MS required to measure the pellet burnup was just renewed in 1998 because of dilapidation of the equipment. Except for sample setting almost all operations are made - 91 - JAERI-Conf 99-009 automatically, and this makes a significant contribution to manpower saving The hydrogen analyzer has been renewed to improve measuring accuracy and to add He analytical function in 1998. A mass analysis detector has been added to this instrument to sure that He content of metallic materials can be measured. This instrument heated up the sample to 2500°C. Therefore, contamination may be spread out the periphery of the main unit. To prevent this, the main unit is accommodated in the glove box. The appearance of this instrument is shown in Fig8. FE-TEM has been installed to allow observation in very fine area and analysis by energy dispersive X-ray spectroscopy. We have made effort to make a good sample to observe the oxide film on the fuel cladding tube and the rim region in fuel pellets. (8)Second instrumental analysis room The second instrumental analysis room is equipped with a cell having with hexahedral sides called satellite cell. This cell is connected with No.l by a pneumatic tube to ensure easy transport of the sample. 4 devices have been adhered to the sidewall of this cell. These are electron micro analyzer (EPMA), secondary ion mass spectrometry (SIMS), X-ray diffractometer (XRD) and laser flash thermal diffusivity measuring device (L/F) that Nuclear Power Engineering Corporation (NUPEC) was introduced for high burnup fuel verification test. These devices are normally accommodated inside the biological shield. Except for the L/F, they are installed on two rails, and can be pulled out whenever required. The inductively coupled plasma-mass spectrometer (ICP-MS) has been installed to perform chemical analysis related to reprocessing It will be used mainly to measure the isotope of pellets and pellet dissolution residual substance. A mockup test is currently being planned to compare interchangeability between this data and that of TI-MS. 2. Sample management system The samples to be handed in the hot laboratory are required to be placed under severe management not only for the nuclear fuel but also RI samples. In this context, we are preparing a new sample management system that consists of new software and hardware. By this system, the speedy and fine sample management will be anticipated. 3. Summary This paper described the current status of NDC fuel hot laboratory. From now on, we has made further efforts to enhance sample preparation techniques, to offer PIE data featuring a higher measurement accuracy. -92- JAERI-Conf 99-009 Tablei Specification of hot cell Wall thickness(m) Dimension of cell Radioactivity Hot cell No. Shielding material Front Rear Ceiling (WxDxH)m (max Bq) Heavy concrete 1 No.1 8X3X7.5 3.0 X1015 Ordinary concrete 1.3 1.2 Heavy concrete 1 No.2 5X3X4.5 1.5X1018 Ordinary concrete 1.3 1.1 No.3 Ordinary concrete 1.1 1 1 3X3X4.5 2.0 X1013 No.4 Ordinary concrete 1.1 1 1 3X3X4.5 2.0 X1013 Lead 0.2 No.5 Ordinary concrete 1 3.5X1.5X2.5 3.0 X1012 Iron 0.3 Satellite cell Iron 0.2 0.2 0.2 About 2.5X2.5X2.5 2.0 X1010 -93- Service area Decomtaminiion area Sample preparation room No.2 Instrumental analysis room Na n Instrumental o analysis room oI o Sample transfer tube o raiioii room LAJ Fig.1 Layout of the Fuel Hot Laboratory JAERI-Conf 99-009 Location Flow Diagram of PIE 1Storage of Spent Fuel Pool Inspection of Rods Re-assembling (Canal) Non-Destructive Test Specific Surface Visual Inspection Length & Profile Measurement Area Measurement* No.1 X-ray radiography Cell Oxide Thickness Measurement * Puncture & Gas Gamma Scanninq Eddy-Current Test Ramnlinn ' 1 No.2 Cutting & defueling Welding Electric discharge Cell Grinding machining Cladding Mechanical Test No.3 Tensile Test Creep Test Burst Test SCCTest* Cell r -• \ Density Measurement SCC Test -> \ Chemical Separation No.4 Dissolution Test Cell W Polishing & Etching • No.5 Metallography Cell Optical Microscopy Micro Hardness Test No.1 Hydrogen & Helium r r mSlrUmcnlal Analysis TI-MS Gas Analvsis Analvsis Room TEM r •1 EPMA* ICP-MS No.2 r Instrumental —^i QrtfAlljfi-* f^iall ^ ^\ OIIVIO / M^O Analysis Room * XRD* \^ • L/F Thermal Diffusivity* * : NU P EC's Devices Fig. 2 A flow of PIE in the fuel hot laboratory - 95- JAERI-Conf 99-009 Neutron No.I Cell Loading unit -Waste storag< Fuel storage rack rack 6m 4m View from A-A section - 96 - JAERI-Conf 99-009 Rod Clamp Rod Support Roller i Fig.5 Appearance of fuel rod drive unit - 97 - Cooling water bonded CKttinS a: (1) Processing procedyre of tensile test specimen > en 2 to o 00 Sample holder o A section of fuel pellet 3mm 3 O O Wax Slice cutting mmammmmM Crack conditions X confirmation by optical microscope observation Processing by diamond cylinder drill (2) Processing procedyre of fuel pellet for the laser iash thermal diffusivify rate measurement Fig.6 An example of specimen preparation JAERI-Conf 99-009 Sample pulling mechanism Autoclave device 'So.2. device No. Fig.7 Appearance of the SCC Devices -99- JAERI-Conf 99-009 Hydrogen analysis Helium analysis con- Shielding Main unit control system trol system \ I Fig.8 Appearance of hydrogen analyser - 100- JAERI-Conf 99-009 SESSION 2 PIE TECHNIQUES ELEMENTARY TECHNIQUES FOR THE STUDY OF POWER REACTOR CHAIR : Y.-S. Kim (Hanyang Univ.) MICRO STRUCTURAL AND QUANTITATIVE ANALYSIS CHAIRS : T. Matsuoka (NDC) and S.-B. Ahn (KAERI) ASSEMBLING AND UTILIZATION TECHNIQUES CHAIRS : K. Ogata (NFD) and Y.-S. Choo (KAERI) - 101- This is a blank page. JP9950631 JAERI-Conf 99-009 2.1 Development and Application of PIE Apparatuses for High-burnup LWR Fuels Katsuya HARADA, Naoaki MIT A, Yasuharu NISHINO and Hidetoshi AMANO Hot Engineering Division Department of Hot Laboratories ABSTRACT The Reactor Fuel Examination Facility (RFEF) is developing the following post irradiation examination apparatuses: •Ion Microprobe mass analyzer (IMA) •Pellet Thermal Capacity measuring apparatus (PTC) •Micro Density Measuring apparatus (MDM) •Shield-type Field Emission Scanning Electron Microscope (FE-SEM) The present paper mainly describes several technical topics of these apparatuses. 1 . INTRODUCTION Extended fuel burnup of LWR fuel is in progress, and the post irradiation examinations (PIEs) of reactor fuels have become very important to clarify the irradiation behavior of high- burnup fuel pellets and cladding tubes. In these contexts, the Reactor Fuel Examination Facility (RFEF) was developing several new apparatuses to obtain the PIE data for confirming the integrity and safety of the high-burnup fuels. Ion Microprobe mass Analyzer (IMA) is used for analysis three-dimensional and isotopes analysis on fuel pellet or micro surface of cladding tube based on the secondary ion mass analysis. Pellet Thermal Capacity measuring apparatus (PTC) is used to measure the thermal capacity of minute fuel pellet specimen based on the thermal flux type Differential Scanning Calorimetry (DSC). Micro Density Measuring apparatus (MDM) is used to measure the micro fuel pellet - 103 - JAERI-Conf 99-009 density by the immersion density method. Shield-type Field Emission Scanning Electron Microscope (FE-SEM) is used to observe and analyze the micro region. Shield-type FE-SEM uses the Field Emission type electron gun and Energy Dispersive X-ray Spectrometer (EDS) for element analysis. 2 . THE OUTLINE OF EACH APPARATUS 2. IMA Qualitative and quantitative analysis in the micro region have become very important to grasp in detail the irradiation behavior of high-burnup fuel pellets and cladding tubes. Data obtained was strongly needed to clarify the irradiation behavior of high-burnup fuels such as Fission Product (FP) gas release, Pellet-Cladding Interaction (PCI), hydrogen pickup in cladding tube, etc. IMA was developed to analyze depth profile of isotopes on fuel pellet or micro surface of cladding tube in order to clarify irradiation behavior in high-burnup fuel by Secondary Ion Mass Spectrometry (SIMS). SIMS uses a focused primary ion beam to erode atoms from a selected sample region. A fraction of the sputtered sample is emitted as positive or negative secondary ion. A secondary ion emitted from the sample is detected by secondary ion optics, then mass spectrometry is carried out. IMA consists of ion guns, sample system, vacuum pumps, secondary ion optics, secondary electron detector and so on. IMA is covered with shielding for radioactive materials, and measuring operation is remotely performd an outside control section. Schematic drawing of The IMA is shown in Figure 1. ENTRY CHAMBER x N SECONDARY ELECTRON^DETECTOFr [ELECTRON y TRON ©SAMPLE SECONDARY ION OPTICS FiQ.1 SCHEMATIC DRAWING OF THE IMA - 104 JAERI-Conf 99-009 The ion gun system has oxygen and cesium ion guns as primary ion source. The oxygen ion gun is used as effectively releasing positive secondary ions for analyzing metal FP, TRU, etc. The minimum diameter of oxygen ion beam is less than 0.6um. The cesium ion gun is applied to release negative secondary ions effectively for analyzing oxygen in the oxide films on the cladding surface and hydrogen in the cladding. The minimum diameter of cesium ion beam is less than 2|im. Objective lens was made to be the plug-in type. It was considered that the contamination by radioactive material discharged from the sample. The secondary ion optics used quadrupolar mass spectrometer, which was simple and compact size. Shielding material made of the tungsten alloy was equipped around the quadrupole in order to decrease the effect on the background by discharged gamma-ray from the sample. The secondary ion lead-in electrode was made to be the plug-in type in order to ease the maintenance. IMA is possible to detect the trace element of ppb level in solid sample, because the secondary ion optics has been equipped with the channeltron. The mass spectrometric analysis is range from 1 to 450 amu. The sample system mainly consists of analysis chamber, entry chamber and sample stage. The sample stage is used by means of the eucentric type. The sample stage can be moved straightly, rotated and tilted, and analyzed within the region of 30mm. The analysis chamber can be evacuated beyond 4x10" Pa in order to analyze Low Mass element such as hydrogen in the cladding tube. Presently, IMA is carried out mock-up examination of sample preparation with small degas, because the measurement is performed on high evacuate condition. 2. 2 PTC The thermal conductivity data is strongly needed for more detailed evaluation of high- burnup fuel behavior. The thermal conductivity is calculated from thermal diffusivity, thermal capacity and density. Until now, the data of the thermal capacity has been using literature data. PTC was developed to measure the thermal capacity of the same small specimen used by the thermal diffusivity measurement and to improve the accuracy of thermal conductivity of irradiated fuel pellet. PTC measures the thermal capacity by heat-flux type Differential Scanning Calorimetry (DSC). The heat-flux type DSC is generally accepted to collect specific heat capacity data in the high temperature region. The basic composition of heat-flux type DSC and the measuring principle of specific heat capacity are shown in Figure 2. The specific heat capacity will be measured as follows: At first, the blank pan and sample pan are put on the holder of heating furnace. Secondly, the temperature is controlled with constant heating velocity. The specific heat capacity can be calculated by the following equation: n WstxHsa _ + Cp = x Cpst WsaxHst - 105 - JAERI-Conf 99-009 Where: Wst, the standard sample weight Wsa, the measured sample weight Hsa, the base line shift quantity of the measured sample (Hsa=C-A) Hst, the base line shift quantity of the standard sample (Hst=B-A) Cpst, the standard sample specific heat capacity at the Tl temperature SAMPLE PAN(Pt) HEATING FURNACE BLANK PAN(Pt) HOLDER(Pt) SENSIBLE HOT PLATE BASELINE(STANDARD) 1 BAS ELI N E (SAM P LE) I THERMOCOUPLE OF TEMPERATURE (TYPE R) OETECTOR OF TEMPERATURE DIFFERENCE RECORDER 1 Fig.2 THE BASIC COMPOSITION OF HEAT- FLUX TYPE DSC AND THE MEASURING PRINCIPLE OF SPECIFIC HEAT CAPACITY. The PTC consists of sensor unit, base unit, sample preparation unit and shielding box. The PTC is able to be operates remotely for using the irradiated sample. Schematic drawing of the PTC is shown in Figure 3. HOT CELL 0>HEATING FURNACE J: HOLDER SAMPLE PREPARATION UNIT (PLATINUM) E CHAMBER £ AMPLE ©THERMOCOUPLE (TYPE? R) I J SENSOR UNITl PAN(P1A[ IMUM1 COVFR(PI ATINUM) ! SAMPLE Fig.3 SCHEMATIC DRAWING OF THE PTC The sensor unit is composed of heating furnace, sample pan, sample chamber and sample transfer system. The measuring temperature range is from room temperature to - 106 - JAERI-Conf 99-009 1500°C. The temperature is measured with typeR thermocouple. The sample pan and sample holder are made of platinum. The sample transfer system can automatically set up a sample to sample holder of which is very weak. The base unit is composed of the electronic circuit for the temperature control and data measuring. The base unit is installed outside shielding box to avoid radiation damage of the electronic circuit. The sample preparation unit is used to pack irradiated fuel pellet fragment for good heating contact. The packed sample can prevent dispersion of radioactive materials. The sample preparation is carried out in hot cell. The measurement data is transferred to the computer, which displays data in real-time and saves. In the data processing, it is possible to carry out specific heat capacity calculation, etc. To verify the performace of the PTC, an experiment has been carried out using platinum sample. Standard sample was A12O3 powder, and measuring sample was platinum wire. The specific heat capacity data was measured in the 100°C step from 200°C to 1400°C at one time. As an example, the measurement result of specific heat capacity of platinum is shown in Figure 4. 0.30 0.25 j: 0.20 E 0.10 0.05 Measuring data Literature data 0.00 200 400 600 800 1000 1600 TEMPERATURE (°O) Fig.4 SPECIFIC HEAT CAPACITYOF PLATINUM The measuring data was compared with literature data, the each data comparatively good linearity. It was possible to obtain the sufficient data in order to confirm the initial performance in case of platinum, but in case of the shape of the sample changes, the data seems to change. The measurement uses different shape and sample will be carried out in the future for establish preparation method and measuring technique, in order to obtain the high precise data using the high-burnup irradiated sample. 2. 3MDM - 107- JAERI-Conf 99-009 The thermal conductivity data is very important in order to carry out the detailed evaluation of the high-burnup fuel behavior. Until now, the measurement of density was used in the one of fuel pellet fragment of approximately lOg. On the other hand, the sample used in the thermal diffusivity measurement is small pellet fragment less than lOOmg. So that, MDM was developed to measure the density of the same sample and to improve the accuracy of thermal conductivity o f irradiated fuel pellet. MDM is measured by immersion density method, which is little effect for shape and open pore of the sample. The immersion density is measured by buoyancy difference of the sample between in liquid and air. The apparatus construction is simply for maintenance. MDM is composed of weighing unit, liquid tank, sample changing system and shielding box. The MDM can be operated by remote handling in order to use the irradiated sample. Schematic drawing of the MDM is shown in Figure 5. v j \ \ 1 1QU1D TANK | vlETAXYLENE r ( ^ATER 1 VIETHYLENE ] ' (,' I (ODfDE \-—(SAMPLE CATENALY v_ SAMPLE CHANGING SYSTEM V. y SAMPLE BASKET Fig.5 SCHEMATIC DRAWING OF THE MDM Sample catenary is installed in the weighing unit on the hook of the bottom suspension style. The balance readability is 1 n g, and its repeatability is ±0.9 /i g. The liquid tank unit is carried out the gravimetry in the atmosphere and liquid by a vertical motor driven movement. One of water, meta-xylene and diiodomethone is used for the immersion liquid. The sample changing system is automatically hanged for a sample on the bottom suspension hook under the weighing cell. The measuring data of weight and temperature are transferred to the computer and then is processed. The gravimetry data is collected with the presetable condition. It takes only the 10 stable data after the immersion of the sample. The mean value of 8 data except for maximum and minimum are adopted in the measuring data. The temperature data in air and liquid are collected and used for parameter of determining sample density. To verify the performance of the MDM and the weight of sample basket affects the 108 - JAERI-Conf 99-009 measuring result, an experiment has been carried out using nickel sample. The immersion liquid used meta-xylene which osmosis better than pure water. The weight of each sample basket is 318.333mg, 326.374mg and 376.567mg. The sample weight is 61.09mg. As an example, The measurement result of density of nickels shown in Figure 6. • A:31 8.333m g • 8:326.374m g 9.1 A C :376.567m g • D .Literature data o s t 8-5 T A B CD Fig.6 DENSITY OF NICKEL The measuring data was compared with literature (4) data. The measuring data is not affected by the sample basket. It was possible to obtain the sufficient data in order to confirm the initial performance of MDM in case of nickel sample. However, it is necessary to consider that the porous influences in the result, in case of the sample such as a fuel pellet. The density measurement used another immersion liquid and sample will carry out in the future for establish measuring technique, in order to obtain the good result using high-burnup irradiated sample. 2. 4 Shield-type FE-SEM Observation and analysis in the micro region are indispensable to grasp the high-burnup irradiation behavior of cladding tube and fuel pellet. Therefore, shield-type FE-SEM was developed. FE-SEM can observe sample with high magnification, since it has resolution of the submicron order in secondary electron image by using field-emission electron gun. The shield-type FE-SEM consists of electron gun, detector, sample stage, vacuum system, and is possible to operate by remote handling outside shielding box. The electron gun uses the cathode field emission type. Since the illumination of cathode emission type is very higher than the conventional thermionic emission type, so it is possible to obtain high resolution in an equal acceleration voltage. The sample stage is eucentric goniometer type, which able to tilt the sample without changing the focus of the observation position. The vacuum system consists of rotary pump, turbomolecular pump and ion pump. - 109- JAERI-Conf 99-009 FE-SEM is equiped with EDS to identify element. Schematic drawing of the EDS is shown in Figure 7. o o LNj TANK EXHAUST SYSTEM SOLENOID VALVE DETECTOR Magnet collimator PREAMPLIFIER COLLIMATOR COLUMATOR DRIVEN MECHANISM MOUNTING FLANGE • SHIELDING BLOCK Fig.7 SCHEMATIC DRAWING OF THE EDS Since sample is irradiated material, therefore, effect of radiation to detector must be reduced to the utmost. The shielding block and collimator made of tungsten alloy were installed at the detector tip in order to reduce the radiation effects and to detect characteristic X-ray with small noise. The elemental analysis can be carried out from oxygen to plutonium. Presently, shield-type FE-SEM is under design and construction. FE-SEM construction will be completed in FY 1999, and performed mock-up measurement. 3. SUMMARY The RFEF is developing IMA, PTC, MDM and Shield-type FE-SEM to clarify the irradiation behavior of high-burnup fuel pellets. It was performed the characteristic test and possible to obtsin the sufficient data in order to confirm the initial performance of apparatus. It is planned to use these apparatus for PIE of high-burnup fuel in the future. 110- JAERI-Conf 99-009 4. ACKNOWLEDGEMENT The authors acknowledge to Dr. T Kodaira, director of Department of Hot Laboratories, for his valuable advice. The authors with to express their sincere thanks to the staff of Department of Hot Laboratories for performing experiment and useful discussions, and to the staff of SEIKO EG&G CO., LTD, RIGAKU Corporation, TAIYO Corporation and JEOL LTD for developing new apparatuses. 5. REFERENCES 1) Development of examination technique; JAERI-Review 98-023, PP.51-63 2) Iwanami Shoten Publishers; The physical and chemistry dictionary. P.934. 3) Maruzen co., ltd; The chemistry handbook basic edition 2. PP.214-216 4) Saito, Y; The base of the thermal analysis for material science; P. 107 5) Kodaira. T, Kikuchi. A; Present Status of PIE Techniques in Tokai Hot Cell Facilities; The 5Th Asian symposium on research reactors; May 29-31, 1996, Taejon, Korea - Ill - JP9950632 JAERI-Conf 99-009 2.2 A TECHNIQUE OF MELTING TEMPERATURE MEASUREMENT AND ITS APPLICATION FOR IRRADIATED HIGH-BURNUP MOX FUELS Takashi NAMEKAWA and Takashi HIROSAWA Alpha-Gamma Section, Fuels and Materials Division O-arai Engineering Center Japan Nuclear Cycle Development Institute O-arai-machi, Higashiibaraki-gun, Ibaraki-ken, Japan ABSTRACT A melting temperature measurement technique for irradiated oxide fuels is described. In this technique, the melting temperature was determined from a thermal arrest on a heating curve of the specimen which was enclosed in a tungsten capsule to maintain constant chemical composition of the specimen during measurement. The measurement apparatus was installed in an alpha-tight steel box within a gamma-shielding cell and operated by remote handling. The temperature of the specimen was measured with a two-color pyrometer sighted on a black-body well at the bottom of the tungsten capsule. The diameter of the black-body well was optimized so that the uncertainties of measurement were reduced. To calibrate the measured temperature, two reference melting temperature materials, tantalum and molybdenum, were encapsulated and run before and after every oxide fuel test. The melting temperature data on fast reactor mixed oxide fuels irradiated up to 124 GWd/t were obtained. In addition, simulated high-burnup mixed oxide fuel up to 250 GWd/t by adding non-radioactive soluble fission products was examined. These data shows that the melting temperature decrease with increasing burnup and saturated at high burnup region. INTRODUCTION The Alpha-Gamma Facility (the AGF) is consisted of equipments for PIE, i.e., twenty one gamma-shielding cells and sixteen glove boxes, and has been operated successfully since 1971. Physical and chemical examination were conducted for the fuel pins irradiated in "JOYO" and overseas reactors such as the Phenix and the FFTF. In a design of fuel pin, temperature of fuel pellet centerline is limited not to exceed the melting temperature of a fuel pellet. During irradiation, the melting temperature of fuel is considered to decrease slightly with increasing burnup due to buildup of soluble fission products in fuel matrix. For irradiated and unirradiated UO2 and (U, Pu)O2 fuels, many workers have determined melting temperatures by so-called V-shaped filament method [1]. However, there have been - 112- JAERI-Conf 99-009 large variation in the results which might be the change of the oxygen to metal ratio and/or the vaporization of the specimen at high temperature. To avoid these effects, Lyon et al. applied the thermal arrest technique to determine solidus and liquidus temperatures for unirradiated UO2-PUO2 phase diagram [2]. In this technique, an oxide fuel specimen is heated within an sealed tungsten capsule, so that effects of the composition change and vaporization of specimen can be avoided during a measurement. Aitken and Evans also evaluated the melting temperature of unirradiated (U, Pu)O2-x as a function of plutonium content and oxygen stoichiometry, 2-x, using this technique [3]. At the AGF, some apparatus for this technique had been introduced in cells and melting temperature has been measured with good accuracy on UO2 and (U, Pu)O2 fuels irradiated in fast breeder reactor (FBRs), an advanced thermal reactor (ATR), and boiling water reactor (BWRs) for more than fifteen years. This paper describes the melting temperature technique and recent results. SPECIMEN PREPARATION Specimens to be examined are both irradiated and unirradiated UO2 and (U, Pu)O2 fuel pellets. At first, fuel pellets are granulated by either drilling or crushing in order to pack the fuel specimen into a tungsten capsule. For irradiated fuel pellets, an irradiated fuel pin is cut into small pieces about 40 mm in length and claddings are removed, before granulation. Then about nine grams of granular fuel is carefully introduced into capsule. The capsule containing the granular fuel is then sealed in a high vacuum of 10"2 to 10-3 Pa by an electron beam welding machine. These preparations are carried out remotely in an alpha-tight steel box(an in-cell box) within a gamma-shielding cell. On this process, the temperature of the whole capsule is liable to rise up by welding heat. It will cause that FP gas retained in the fuel is put away, the pressure in the capsule heightens, the lid of a capsule is pushed up, and then the normal welding becomes impossible. In the worst case, the fuel in the capsule scatters out. Therefore, it is necessary to keep the temperature of the capsule as much as possible low temperature. Then, cooling method of the capsule under welding was devised in this system. In an electron beam welding machine, the tungsten capsule is set by a chuck. The outline drawing of the chuck is shown in Fig. 1. The base design is a three point scroll chuck , and it was improved into the plane chucking system for fitting to the capsular shape. So that the contact area with the capsule drastically increased, and the heat radiation effect was heightened. The chuck is made of copper metal and tantalum liner so that it has also a function of heat sink. The tantalum liner of the high melting point prevents depleting the copper heat sink by heat in a high vacuum condition. In addition, the stainless plate improves the handling by the manipulator is installed in the upper surface of the chuck. The welding process was also devised that the capsule is welded in the low power condition temporary, and complete welding is carried out afterwards. - 113- JAERI-Conf 99-009 DESCRIPTION OF THE MELTING TEMPERATURE APPARATUS The main part of the apparatus is located in a gamma-shielding cell. Schematic drawing of the apparatus is shown in Fig. 2 and Fig. 3. The furnace unit is installed in an in-cell box. The vacuum chamber is evacuated to lO2 Pa. The specimen, enclosed in a tungsten capsule, is located in a tungsten crucible which is set in the center of the induction furnace to heat up the tungsten capsule uniformly. The water cooled concentrator is equipped to provide the magnetic field with the crucible efficiently and to shield the glass chamber from high temperature radiation. The temperature of the specimen is continuously monitored by two sets of two-color pyrometers. Heating power is controlled by a signal from the upper pyrometer. The temperature of the specimen is monitored with the lower pyrometer sighted on a small diameter well at the bottom of the capsule which simulates black-body. The light path traverses through the revolving protection glass disk, the vacuum enclosure window, and the shielding cell window. The revolving protection glass disk which is nearest to the furnace, prevents the vacuum enclosure window from being coated by vapor deposition. When the protection glass disk is dimmed, the disk is rotated to provide a new clean path. MAINTENANCE OF THE APPARATUS On the measurement operation for above 15 years, the capsule occasionally leaked, because of rise of the capsule internal pressure by the volatile FPs, coarsening of crystal grain by the recrystallization of the tungsten and embrittlement being generated in high temperature heating. The volatile FPs which was discharged out under heating deposited to the low temperature portion of the apparatus that is mainly the concentrator in the furnace unit and the exhaust duct of the vacuum pumping system, and the radiation dose rate of the equipment rose very much. The contaminated concentrator can be exchanged periodically, because the remoteness disassembly is possible. However, the exchange is not possible for the exhaust duct, since it is being fixed in the in-cell box. In the point maximum, the dose rate reached 320 mSv/h. Such a high radiation condition interfered with ordinary maintenance of the apparatus. Then, the equipment for the remote decontamination was devised, and the exhaust duct inside and heating furnace inside were tried to decontaminate. However, it was not possible to lower only to about 190 mSv/h, since the inside is very narrow and sufficient decontamination was impossible. Some radiation shielding devices was designed and used for ordinary maintenance such as exchange of the oil of the vacuum pump and exchange of the globe under the environment of such high dose rate. The shielding screen consisted of the lead on 1800 mm height, 1300 mm width, 30 mm thickness. By using these shielding devices, the environment of maintenance area was secured and maintenance of the apparatus was able to be carried out. However, this work is hard to be carried out frequently. Therefore, the whole equipment will be renewed in FY 1999. - 114 - JAERl-Conf 99-009 EXPERIMENTAL DATA AND DISCUSSION 1. Accuracy It is most important to obtain the accurate temperature of a specimen in the melting point measurement. It is necessary that absolute temperature is calibrated by the calibration formula in the two-color pyrometer. The temperature reproducibility is high for this pyrometer itself, and the measurement with the high accuracy becomes possible, if the calibration formula is obtained. There is the ratio of observed emissivity to rue one of measured object (i.e., tungsten) as a main error factor of the pyrometer. The emissivity of tungsten is well known, but there are some differences by the surface state (e.g., roughness ). Then, the temperature measuring well was made at the capsule bottom to simulate black-body, in which the effective emissivity was about 1. At first, the well was designed the dimension with 1 mm in diameter and a depth of 10 mm. However, adjusting condition for focusing of the pyrometer is very severe, since a distance from the tungsten capsule is about 2 m and a sight range of the pyrometer is not less than 1 mm in diameter. Then, diameter of the well was enlarged up to 2 mm with the same length that is the limit to get a black-body condition. It contributes to good accuracy of measured temperature data. The uncertainties of measured temperature on each capsule were reduced by this method. Furthermore, transmission of the glass and reflection by the mirror in the measurement optical path affect the emissivity, and it becomes a factor of the indication error. Then it must be corrected as an emissivity ratio. The correction is possible to be carried out by these condition being fixed. However, it must be evaluated at each measurement, since the transmittance of the window glass gradually changes by each measurement operation. At the AGF, melting point of reference sample (i.e., Mo metal of mp 2630*0 and Ta metal of mp 2990^) is measured before and after the measurement of the fuel as an error evaluation method. These reference sample were selected since they are metal sample which melting point is around and near to the fuel's one. As the above description, the slippage of melting point around the measurement of reference sample was adopted as an error of the pyrometer indicating temperature. The whole measurement error is added reading error of thermal arrest in the heating curve. The melting point of unirradiated UO2 was verified being 2845 ± 12*0 at the AGF in which 2847 ± 25'C is the recommendation value of the IAEA [4]. As the other report for unirradiated UO2 using with the encapsulation method, Aitken & Evansns [3], Lyon & Baily [2] and Latta & Fryxell's [5] is representative, and the measurement accuracy is ± 251C, ±2013, ± 15 *C each. These measurement error do not carry out the evaluation of the every sample like this report. For irradiated MOX fuel,Krankota &Craig [6] reported the error ±52^-+ 130'C. This large error was caused by the V-shaped filament method, since the specimen is not enclosing and the composition change is easy to be caused by unequal evaporation and the oxygen transfer. Adamson et al.[7] mentioned about the reliability of the V-shaped filament method that this method gave melting temperatures approximately 50*0 lower than true one. - 115 - JAERI-Conf 99-009 2.Melting Temperature Data At the AGF, melting temperature has been measured with good accuracy on UO2 and (U, Pu)O2 fuels irradiated in FBRs, ATR, and BWRs. Typical data are shown in Fig.4. The melting temperature of 29 wt% PUO2-UO2 obtained at the AGF are plotted,, comparing with the data of 25 wt% one reported by Krankota and Craig [6]. Their data locate at higher temperature range than the AGF's, and it is well-known that melting temperature decreases with an increase in PuO2 concentration[6,7]. In this data, the maximum burnup of specimen is 124 GWd/t which was irradiated in "JOYO". This figure also includes the data obtained from simulated burnt MOX fuels. They were prepared by adding the non-radioactive oxide powder of eight soluble FP elements to unirradiated or irradiated fuel powder, that amount corresponds to the maximum burnup of 250 GWd/t. These data shows that the melting temperature of 29 wt% PUO2-UO2 decrease almost linearly with burnup at a rate of 5 "C within the lower burnup region up to 170 GWd/t, and almost saturate above it. It may be necessary to measure the actual specimen of higher burnup in order to confirm this behavior. SUMMARY A technique and operation for melting temperature measurement of irradiated oxide fuels at the AGF is described. In this technique, the fuel is enclosed in a sealed tungsten capsule to prevent chemical change, and heated in the induction furnace. The melting temperature is determined from arrest of the heating curve. Observed melting temperature of MOX fuel decrease with increasing burnup and saturated at high burnup region. REFERENCES [1] For instance; Christensen, J. A., et. al., WCAP-6065 (1965). [2] Lyon, W. L., Baily, W. E., J. Nucl. Mater., 22 (1967) 332. [3] Aitken, E. A., Evans, S. K., GEAP-12229 (1971). [4] M.H.Rand et al., Rev. Int. Hautes Temp. Refract. 15 (1978) 355. [5] R.E.Latta & R.E.Fryxell, J. Nucl. Mater., 35 (1970) 195. [6] Krankota, J. L., Craig, C. N., GEAP-1315 (1969). [7] M.G.Adamson, etal., J. of Nucl. Mater., 130 (1985) 349. - 116- JAERI-Conf 99-009 Handling plate Copper heat sink Tantalum liner Driving screw Fig. 1 Structure of the capsule chuck in an electron beam welder '/7/////7//A cimen (Tungsten capsule) Manipulator ) Heat shielding ) Concentrator • Induction heating coil 1 Vacuum chamber (Quartz) _ Shielding 1 Cooling water window Argon gas supply In-cell box Exhaust duct Hydraulic lift " Gamma-shielding Diffusion pump Rotary pump Mirror Two-color pyrometer Window Induction cable Protecting glass Inverter Fig. 2 A schematic drawing of the melting temperature measurement apparatus - 117- JAERI-Conf 99-009 Specimen Tungsten capsule Tungsten crucible Crucible support • Concentrator • Upper radiation shield Lower radiation shield • Cooling water passage • Isolator (Alumina) • Vacuum chamber (Quartz) • Induction heating coil Black-body well Fig. 3 A schematic drawing of the induction furnace unit and a tungsten capsule Report Pu content (wtS6) Specimen Krankota 3000 O 25 irradiated fuel and Craig A 29 irradiated fuel Data at the AGF • 29 unirradiated fuel+simulated FP 2900 • 29 irradiated fuel+simulated FP | 2800 8. _|A ..A 2700 2600 2500 L I • I 0 20 40 60 80 100 120 140 160 180 200 220 240 Bunnup (GWd/t) Fig. 4 Melting temperature of MOX fuels as a function of burnup 118- JP9950633 JAERI-Conf 99-009 2.3 Development of PIE Techniques for Irradiated LWR Pressure Vessel Steels Masahiro NISHI, Minoru KIZAKI and Tomohide SUKEGAWA Research Hot Laboratory Division Department of Hot Laboratories Tokai Research Establishment, JAERI Tokai-mura, Naka-gun, Ibaraki-ken, Japan ABSTRACT For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. 1. INTRODUCTION Mechanical properties of RPV steels and fuel cladding of LWRs at the post irradiation state are the key parameter for the evaluation of safety, structural integrity and lifetime as well as the material development. The mechanical tests at the RHL have been performed for 38 years to support R&D works in this field at JAERI. The data of Charpy impact test are effectively utilized to evaluate the neutron irradiation embrittlement. Recently, the existing Charpy impact testing machine was remodeled in order to improve its accuracy and reliability. - 119- JAERI-Conf 99-009 By this remodeling, absorbed energy and other useful information can be delivered from one-time striking. In addition, the remote machining technology from actual irradiated RPV steels has been developed in order to clarify the aging behavior of LWRs at the RHL. Another new technique is developed to determine the post-irradiation fatigue characteristics of structural and fuel cladding materials as low and high-cycle fatigue tests technology with the function as tensile test equipment. The present paper describes the outline of two mechanical testing apparatuses and techniques and remote machining of mechanical test pieces for irradiated LWR-RPV steels and so on. 2. REMODELING OF CHARPY IMPACT TESTING MACHINE 2.1 Remodeling The Charpy impact testing machine was redesigned and modified in order to clarify the neutron irradiation embrittlement behavior of LWR-RPV. This machine instrumented with electronic measuring devices to detect an impact force and a displacement of specimen has an automatic specimen setting system. The block diagram of instrumented Charpy impact testing machine is shown in Fig.l. The load capacity is 300J and it is possible to test in the •DISPLACEMENT MEASURING SYSTEM- FORCE MEASURING SYSTEM HAMMER (SEMICONDUCTOR ACTIVE STRAIN GAGES) Fig.l Block diagram of instrumented Charpy impact testing machine - 120- JAERI-Conf 99-009 temperature range from -140t to 240°C by using two types of agitated liquid baths. The test specimen is transferred from the cooling (or heating) bath to an anvil of the machine using industrial robot, and struck by a hammer within 4 seconds after removal it from the medium. The test temperature accuracy is within 0.5^. The test items are V-notch Charpy impact test and KId dynamic fracture toughness test. The data from Charpy impact test are evaluated on ductile/brittle transition temperature using least squares method, referring a polynomial expression, a hyperbolic tangent and a Gaussian error function. The sensor for the load detection was composed of two semiconductor active strain gages on the tup and two dummy gages put on near the hammer. Moreover, a potentiometer for the displacement detection was inserted and fixed to the hammer shaft. These signals from sensors are recorded in the wave-memory with the capacity of 32Kwords x 2 channels, the resolution of 12bit and the sampling rate of 50~500nsec/word. Collected data are utilized for data processing and analysis. 2.2 Characteristics A calibration technique is most important in the instrumentation, because it is utilized to convert the data sampled by measuring devices to the force and displacement. The calibration must be carried out in a constant condition that is never affected by a human error and the environment. The RHL developed completely new methods for a load calibration technique as shown in Fig.2. The calibrator can be fastened with positioning and height adjusting to the specimen support on anvils without causing pre-strain to the semiconductor gages pasted on the hammer tup. Therefore, the calibration is possible by the comparison between a proof load from the calibrator and an output of the force- measuring device. In addition, the RHL found that the conversion value should be corrected every time because an output mentioned above changes with a change in ambient temperature. An equation of correction derived from proof tests offers a high accurate conversion value. On the other hand, the calibration method for the displacement measurement is shown in Fig.3. For the same reason, the calibrator with an electric micrometer is also fixed to the specimen support. Then, calibration is performed by comparing a proof displacement from the calibrator with an output of the displacement-measuring device. 2.3 Performance An example of obtained data from PIE is shown in Fig.4. A force- displacement curve shows the impact properties during the striking very clearly. The Fgy, Fm, Fu and Fa marked on the curve mean a yield impact force, a maximum impact force, an unstable crack (cleavage crack) initiation force and a crack arrest force respectively. While, the Einst is total impact energy which is - 121- JAERI-Conf 99-009 area under the complete force-displacement curve. In addition, the curve offers a ductile erack initiation energy, partial impact energy at above points, total displacement of specimen and so on. These impact properties and characteristic values of the points are defined in accordance with ISO standard ISO/DIS 14556(1). As shown in the figure, the Fgy, Fm and Fu decrease with increasing the test temperature, whereas the Fa increases. The Fa on the curve means a specimen struck in the transition range. The force-displacement curves obtained from PIE after remodeling are very useful data for the evaluation of irradiation embrittlement. SIGNAL FROM CSEMICONDUCTOR STRAIN GAGE PROOF LOAD APPLYING SYSTEM (0~3ton) Fig. 2 Force cal ibrator 0 MOVED AXIS ANGLE SIGNAL FROM POTENTIOMETER J—»\AMPLIFIERJ HAMMER ELECTRICAL MICROMETER TUP WITH SEMICONDUCTOR STRAIN GAGE' ilSPLACEMENTOnm) Fig. 3 Displacement calibrator - 122 - JAERI-Conf 99-009 Material: A533B-1 Fluonce: ~9X1O" n/cm2 Fm(Fu) Fm Fm Curves CD ® <3) ® Test Temp (°C) 50 120 125 180 Shear Area (%) 7 37 71 100 (KN) 16.028 15.891 15.823 15.469 Fgy (mm) 0.442 0.436 0.438 0.452 (kN) 18.679 18.970 18.795 18.061 (mm) 0.936 1.715 1.898 1.687 (kN) 18.679 18.579 17.103 — Pi • (mm) 0.936 1.849 3.187 — Fa (kN) — 7.246 9.763 — (mm) — 1.950 3.192 - Displacement Fig.4 An example of PIE data obtained from remodeled Charpy impact testing machine 3. DEVELOPMENT OF REMOTE MACHINING TECHNOLOGY 3.1 Design concept In case of irradiated material, since all of manipulation must be remotely handled, machining of the mechanical test specimen should be performed accurately in accordance with the material testing standards such as ISO, ASTM and JIS. However, the remote machining with high accuracy has never been done up to date because the requirement is quite difficult. Therefore, a numerically controlled machine tool is selected and developed as the most useful apparatus for hot cell work without a human error. As shown in Fig.5, a computerized numerical control (CNC) milling machine developed is composed with the main body for machining and a control system included a personal computer. There are two main bodies produced by same design, which can be operated by an identical controller. One is installed in the hot cell and the other is set up in the operating area for mock-up test prior to in-cell machining. The original machine for general use is modified according to some requirements from radiation environment, free maintenance and higher - 123- JAERI-Conf 99-009 performance. Moreover, the innovational techniques are applied to achieve the allowable machining by means of remote handling. Fig.5 Computerized numerical control (CNC) milling machine Fig.6 Charpy (type-A) impact test specimen machined by CNC milling machine 3.2 Characteristics and Specification The CNC milling machine developed is extremely compact of 850 x 800 x 1100mm and quite on a par with a general machine tool in mechanical accuracy. In addition, the main body with an automatic tool changer (ATC) included necessary six tools is highly rigid portal structure. The parts reviewed and redesigned are as follows. The motor for X/Y/Z motion was modified from a DC servomotor with rotary encoder to an AC servomotor with resolver. DC servo - 124 JAERI-Conf 99-009 spindle motor of 200w was also changed to an AC servo of 800w. Then, a magnetic device was selected as the sensors for positioning and the amplifiers were located into control box in operating area. Moreover, the innovational techniques added are as follows. 3.2.1 Clamping mechanism of the work A clamping mechanism is composed with a hydraulic vise for clamping of a four-cornered work and an electric rotating chuck with AC servo resolver motor combined a bearing which is able to rotate the vice. Therefore, it is possible to machine without re-clamping on five surfaces except bottom of the rectangular work and notch-machining to the top surface. 3.2.2 Automatic measuring mechanism of the origin in Z-motion In case of a general machine, the origin of Z-motion is manually decided by using a so-called touch sensor, standard gage, which can indicate a contact point of the mill tip and the sensor top. However, decision of the Z-origin with remote handling is not so easy because the largish sensor must be set on a central part of the smallish work. For that reason, the automatic measuring mechanism that is able to detect a contact point through an electrical signal was designed. This idea was accomplished by insulating of spindle using a ceramic coating. 3.2.3 Machining techniques In general, the tool and work during machining is cooled by plenty of oil, however, the dry machining is the best way, in case of hot cell work, if possible. CNC milling machine in the RHL has an air-cooling system and an intermittent oil-atomizer, which performs the duty as a lubricant rather than a coolant. The dry machining is achieved by finding out the suitable conditions including kind of machining, selection of tool, machinability of work, cutting and feed speed and so on. Then, an atomizer automatically and periodically sprays cutting oil, if it is necessary. One of problems in specimen machining of RPV steels is the removing technique of barr (frash) growing on edges. To resolve the problem, the unique mechanism that is utilized CNC milling function is now under development. By this idea, the barr will be removed by a whetstone with automatic operation as a series work of the machining. 3.2.4 Automatic marking technique for identification To avoid a blunder, the marking for identification should be performed during the machining or immediately after it finished. In addition, the marking must not affect mechanical and metallurgical properties of the specimen. Automatic marking of identification number is possible by combining a - 125 JAERI-Conf 99-009 pneumatic marking pen with a function of CNC milling machine. The pen is fixed by air chucking with easy handling like a machining tool on ATC magazine. Therefore, the identification number can be automatically marked on the programmed position according to command with key-in. 3.2.5 Dimension measuring system for the machined specimen The machining error that is the deviation from programmed nominal value is mainly depending on the rigidity and actual diameter of the tool. To compensate the error, the CNC operation program requires the difference between the command and as-machined condition. Therefore, a dimension measuring system utilized CNC milling function was designed. It is composed with two SONY magnescales for measuring the X/Y motions, a touch sensor for commanding the start/end of measuring, a display unit (counter), a resetting jig of the sensor and a holder for combining the sensor with CNC milling body. The sensor is able to touch automatically one side surface and the opposite face of the machined specimen and measures the wide and length. In addition, the TV monitoring system for confirming the machined notch- shape is now under investigation. 3.3 Performance The machining programs developed in the RHL are a Charpy impact test specimen type A of 10 x 10 x 55(mm) so-called V-Charpy, a plate type tensile test specimen with parallel part of 22.95Lx 3Wx 3t(mm), and a three point bending type fracture toughness test specimen with knife-edges. The CNC milling machine is possible to machine these specimens with high accuracy that satisfies the standards, ISO, ASTM and JIS. As one of examples, a machined Charpy impact specimen is shown in Fig.6. For checking of the machining condition, there are a sound sensor and an acceleration sensor on the body. Moreover, a TV monitoring system for observing of tool-edge chipping and chips of work is located nearby the machine. 4. DEVELOPMENT OF REMOTE SYSTEM TECHNOLOGY FOR FATIGUE TESTING 4.1 Fatigue testing machine One of the important research subjects on the LWR fuel cladding performance at extending burn-up is to understand the mechanical properties. The RHL developed an electro-hydraulic fatigue testing machine with two kinds of load cells and servo valves in tandem and in parallel respectively. - 126- JAERI-Conf 99-009 Main load cell with 20KN is used when the maximum load will be more than 2KN for dynamic and 3KN for static loads. Sub load cell with 2KN is utilized for the specimens with lower maximum load than the above-mentioned one. In case of servo valves, the larger valve with discharge rate of 38 1/min is utilized for the high frequency test condition with high load. It is possible to load up to lOOHz. The smaller one with discharge rate of 3.8 1/min is very useful for most of PIEs such as a low frequency test, a lower load test and all of static tests. Therefore, even the tension/compression tests bellow 200N, it is possible with high accuracy. A changing of the signal between the larger ones and the smaller ones can be carried out by manual work at in-cell for the load cells and by switching from out-cell for the servo valves. By exchanging the test fixtures with remote handling, the machine is utilized for a high-cycle fatigue test with arc-shaped specimen machined from LWR fuel cladding, a low-cycle fatigue test with round specimen from structural materials, a crack propagation test and a high-frequency test. Moreover, tensile test, plan strain fracture toughness test and the fatigue pre-cracking for fracture toughness specimen are also possible. The specification and the performance of fatigue testing machine are shown in Table.1.