JAEA-Review

Proceedings of the Specialist Meeting on Mo-99 Productionby (n, ) Method
March 9-10, 2012, Yurakucho Asahi Hall, Tokyo, Japan(Eds.) Etsuo ISHITSUKA, Masahiro ISHIHARA and Masahide SUZUKI

Neutron Irradiation and Testing Reactor Center
Oarai Research and Development Center

July 2012
Japan Atomic Energy Agency

日本原子力研究開発機構

JAEA-Review

2012-030

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JAEA-Review 2012-030
Proceedings of the Specialist Meeting on Mo-99 Production by (n, γ) Method
March 9-10, 2012, Yurakucho Asahi Hall, Tokyo, Japan(Eds.) Etsuo ISHITSUKA, Masahiro ISHIHARA and Masahide SUZUKI

Neutron Irradiation and Testing Reactor Center,
Oarai Research and Development Center,
Japan Atomic Energy Agency
Oarai-machi, Higashiibaraki-gun, Ibaraki-ken(Received May 11, 2012)

This report summarizes the documents presented in the Specialist Meeting on 99Mo Production by(n, γ) Method, which was held on March 9 to 10, 2012, at the Yurakucho Asahi Hall in Tokyo, hosted by
Japan Atomic Energy Agency. The objective of the meeting is to exchange the information of current
status, future plan for the 99Mo production, and to make a discussion of “How to cooperate” in eachresearch and test reactor. There were 27 participants from Poland, Kazakhstan, Indonesia, Thailand,
Malaysia, Netherlands, Korea and Japan.
As a result of this meeting, it was recognized that to push forward the development of 99Mo
production by (n, γ) method is necessary for the future steady supply of 99Mo. Moreover, an irradiation
test using the high density MoO3 pellet developed by the Japan Atomic Energy Agency was proposedfrom a viewpoint of a merit for the common irradiation target in each research and test reactor.

Keywords : 99Mo Production, （n,γ）Method, Research and Test Reactors, Irradiation Target, MoO3i

JAEA-Review 2012-030「(n,γ)法による Mo-99 製造に係る専門家会議」 講演集
2012 年 3 月 9-10 日、有楽町朝日ホール
日本原子力研究開発機構

大洗研究開発センター

照射試験炉センター（編） 石塚 悦男、石原 正博、鈴木 雅秀(2012 年 5 月 11 日 受理)

本講演集は、（独）日本原子力研究開発機構主催の「(n,γ)法による Mo-99 製造に係る専門家会議」に提出された発表資料をまとめたものである。本会議は、各国の 99Mo 製造の現状と計画に関する情報交換及び 99Mo 製造に係る今後の試験研究炉の協力について議論することを目的として、2012 年 3 月
9 日から 10 日に東京の有楽町朝日ホールで開催された。会議には、ポーランド、カザフスタン、インドネ
シア、タイ、マレーシア、オランダ、韓国、日本の 8 カ国から合計 27 名が出席した。
会議の結果、将来の 99Mo の安定供給のために、(n,γ)法による 99Mo 製造技術開発を進めることで一致するとともに、照射ターゲットの共通化によるメリットを考慮して、原子力機構で開発した高密度 MoO3
ペレットを共通ターゲットとした照射試験が提案された。

大洗研究開発センター：〒311-1393 茨城県東茨城郡大洗町成田町 4002

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Contents
1.

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

2.

Approach of 99Mo production in Japan ............................................................................... 1
2.1 Status of Japan ................................................................................................................ 1
2.2 R&D in Japan .................................................................................................................. 3

3.

Status and plan in each country ........................................................................................... 4

4.

Discussion for cooperation ................................................................................................ 10

5.

Conclusion ........................................................................................................................ 11

Appendix ................................................................................................................................. 15
A.1 List of participants ....................................................................................................... 15
A.2 Program ....................................................................................................................... 17
A.3 Presentation documents for approach of 99Mo production in Japan ............................. 19
A.4 Presentation documents for status and plan in each countries ...................................... 67目

次

1. 序 論 .................................................................................................................................. 1
2. 日本における 99Mo 製造の取組 ......................................................................................... 1
2.1 日本の状況 .................................................................................................................... 1
2.2 日本における技術開発 .................................................................................................. 3
3. 各国の現状と計画 ................................................................................................................. 4
4. 協力に向けた議論 ............................................................................................................ 10
5. 結 論 ................................................................................................................................. 11付 録 ...................................................................................................................................... 15
A.1 参加者リスト ................................................................................................................. 15
A.2 プログラム .................................................................................................................... 17
A.3 日本における 99Mo 製造の取組に関する発表資料 ................................................... 19
A.4 各国の現状と計画に関する発表資料 ......................................................................... 67

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1.

Introduction
Virtually all 99Mo used over the world is produced by fission method and four companies haveshared more than 95% of the world 99Mo supply using five research reactors and four post processing
facilities. On the other hand, there are many other research reactors over the world which are capable ofproducing the 99Mo. Thus, the risk of the shortage for the 99Mo supply can be significantly reduced if
99

Mo production is shared by many reactors. However, the production of the fission 99Mo has

disadvantage in terms the nuclear proliferation resistance, waste management etc.
The 99Mo production by (n, γ) method had been used in the 1970~1980 until the fission 99Mo
became available. In the regional supplies of 99Mo, the (n, γ) 99Mo has several advantages comparedwith the fission 99Mo, however specific activity at about three order lower than that by the fission 99Mo
makes less commercialization than the fission 99Mo.
In this situation, the Japan Atomic Energy Agency (JAEA) held the “Specialist Meeting on 99Mo
Production by (n, γ) Method” on March 9 to 10, 2012, at the Yurakucho Asahi Hall in Tokyo, Japan.
The objective of this meeting is to exchange the information of current status, future plan for the99

Mo production, and to make a discussion of “How to cooperate” in each research and test reactors.

There were 27 participants from Poland, Kazakhstan, Indonesia, Korea, Malaysia, Thailand,
Netherlands and Japan. Participants and program are shown in Appendix A.1 and A.2. In this report,
documents presented in the meeting are summarized, and also the results of discussion in this meetingare summarized.

2.

Approach of 99Mo production in Japan
Status and R&D in Japan were presented in this meeting. Presented documents are shown inAppendix A.3. Summaries presented as the “Approach of 99Mo production in Japan” are described asfollows.
2.1 Status of Japan(1) Affects of earthquake and re-operation of JMTR
M.Kaminaga (JAEA)The Japan Materials Testing Reactor (JMTR) in JAEA is a light water cooled tank type reactor with
first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by acanal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiatedsamples.
The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion
reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check& review on the reoperation had been conducted by internal as well as external committees. As a result-1-

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of the discussion, the JMTR reoperation was determined, and refurbishment works started from the
beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from
JFY 2007.
Unfortunately, on March 11 2011 (JFY 2010), the Great-Eastern-Japan-Earthquake occurred, andfunctional tests before the JMTR restart, such as cooling system, reactor control system and so on, were
delayed by the earthquake. Moreover, a detail inspection found some damages such as slight
deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of
the JMTR will be delayed from June to next October, 2012. Now, the seismic influence evaluation after
the earthquake disaster is being carried out aiming at the restart of the JMTR.
The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until
around JFY 2030.
(2) Needs and current status of 99Mo/99m Tc production in Japan
T.Genka (JAEA)
99m

Tc, the decay product of

99

Mo, is used to diagnose a variety of diseases for approximately nine

hundred thousand patients in Japan each year. However, 99Mo/99mTc has not been produced in Japannevertheless its consumption ratio of global supply is the second largest (14 %) after the United States(44 %).

Accordingly resent shutdown of Canada’s NRU reactor and Netherlands’ HFR has brought

increased attention to Japan’s reliance on foreign production of vital medical radioisotopes. A primaryconcern is the reliability of 99Mo supply. Once the NRU reactor or HFR shuts down, it results increating serious domestic shortages of 99Mo. There are also concerns over the age of the foreign reactors,
as many of them are reaching the end of their operation life. Another concern is the proliferation of
HEU, which is weapons-grade material.
As the past achievements in Japan, production method of 99Mo based on nuclear fission processwith UO2 (235U 2.6%) pellet was developed in 1977 by JAERI (presently JAEA). Moreover neutron
capture reaction process with MoO3 pellet for production and local distribution of 99Mo was alsoestablished in 1980. Distributed quantities of fission and neutron capture method were 20 Ci/week(1977~79) and 200Ci/week (1984~85) respectively. However, due to budgetary cutback policy in 1995,
the then Cabinet decided to discard the development of such radioisotopes that were able to import fromoverseas. Meanwhile, from 2001 to 2006, a portable type 99mTc generator (up to 300 mCi 99Mo loaded)based on PZC adsorbent was developed under the framework of FNCA supported by MEXT, Japan.
The fission method using enriched uranium target (either HEU or LEU) has not been considered inJapan from the view point of non-proliferation of nuclear weapons nevertheless it enables obtaining
high specific activity 99Mo suited to large scale production of quality radiopharmaceuticals.
Currently ongoing trials for 99Mo domestic production by various sectors in Japan are;1) Production of 99Mo by thermal neutron capture reaction as 98Mo(n, γ) 99Mo using 98MoO3
pellet target along with the JMTR renewal program of JAEA,
2)

99

Mo production in a commercial power plant reactor (BWR) with (n, γ) reaction using 98Mo

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metal target,
3) Production based on 100Mo(n, 2n)99Mo reaction by fast neutron irradiation in the accelerator of
JAEA’s FNS facility,
4) Production based on reactions of 100Mo(p, pn)99Mo in a medium scale cyclotron and
100Mo(p, 2n)99mTc with a mini cyclotron in nuclear medicine facility, and

5) Separation, extraction and concentration of 99mTc from the solution of neutron activated 99Mo
using activated charcoal.
This presentation summarizes efforts to create a reliable and complementary domestic supply of99

Mo/99mTc in Japan.

2.2 R&D in Japan(1) Status of 99Mo/99mTc Production Development by (n, γ) Reaction in Japan
K.Tsuchiya (JAEA)
The renewed JMTR will be started from the later half of JFY2012, and it is expected to contributeto many fields. As one of effective uses of the JMTR, JAEA has a plan to produce 99Mo by (n, γ)
method, a parent nuclide of 99mTc. 99mTc is most commonly used as a radiopharmaceutical in the field of
nuclear medicine. Currently, in case of Japan, the supplying of 99Mo depends only on import fromforeign countries, therefore JAEA has performed the R&D on production method of 99Mo/99mTc in
JMTR with foreign organizations and relevant Japanese manufacturers under the cooperation programs.
The status of R&D and construction of facilities for demonstration tests are as follows;
1) Development of the sintered MoO3 pellet with high density as irradiated target material,
2) Extraction and concentration of 99mTc solution by solvent extraction from Mo solution,3) Extraction of 99mTc solution with Mo absorbent from Mo solution,4) Mo recycling from the used Mo generators and solutions,5) Construction of new hydraulic rabbit irradiation facility in JMTR.
The status of the R&Ds in Japan is introduced for the production of 99Mo/99mTc. The constructionsof the irradiation and PIE facilities are also introduced for the production of 99Mo/99mTc in JMTR.
(2) R&D in Japan on Extraction and Concentration of 99mTc,- Preliminary Study using Re instead of 99mTc M.Tanase (Chiyoda Technol Corporation)
A method is discussed for producing

99m

Tc from99

Mo by extraction with MEK and concentration

with alumina column. The purpose of this study is to get high specific-volume solution of 99mTc fromneutron-irradiated MoO3 in large quantities. The experiments were carried out using Re instead of 99mTc.
The process is as follows:
1) Dissolution of MoO3,
2) Addition of Re,
3) Extraction of Re with MEK,

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4) Passage of MEK containing Re to tandem-column system; the upper column for removing Mo andH2O and the bottom one for adsorbing Re,
5) Elution of Re from the bottom column by saline, and
6) Inspection of the product. The Re product was obtained successfully with yields of over 95%, andradioactive impurity of 99Mo was estimated to be less than 0.015% when 200 g of MoO3 is
irradiated in JMTR and 13.2 TBq(357 Ci) of 99mTc is used, and also the impurity of MEK in the
product to be about 600 ppm. These values of impurities will be below the limits of tentative
regulations in Japan. The concentration of the product obtained was also estimated to becorresponding to 728 GBq(19.7Ci)/ml

when 200 g of MoO3 and 13.2 TBq(357 Ci) of 99mTc areused. The method proposed would be applicable to a practical production of 99mTc obtained from (n,γ) 99Mo in large quantities.
(3) Technical Request/Standard for 99mTc Solution for Medical Use in Japan
M.Kobayashi (FUJIFILM RI Pharma Co. Ltd.)
Some statistics related to the supply/sales of radiopharmaceuticals in Japan are introduced. They
include the quantity of in vivo procedures in Japan and number of hospitals and clinical laboratories
where the nuclear medicine is practicing. Standard specifications or criteria for 99Mo production byeither reaction of (n,f) or (n, γ) are provisionally decided to be identical but those of 99mTc solution as
sodium pertechnetate are compared among the pharmacopoeias of Japan, USA and Europe. Due to thenature of short half-life materials, strict time control and process control with carefully scheduled
management are essential in the production and delivery. And high enough radioactive concentration of
the final solution of 99mTc, more than 1 Ci/ml at the start of processing, is requested for the productionof quality radiopharmaceuticals. Detailed description of the flow of processing is to be introduced froma stand point of radiopharmaceutical manufacturer.
Finally, the requirement of high quality 99mTc solution as raw materials is briefly introduced basedon the guideline for sterilized medicine obtained by aseptic process which was issued by the Ministry of
Health, Labor and Welfare (MHLW) of Japan on 20th April 2011.

3.

Status and plan in each country
Status and future plan of Poland, Kazakhstan, Indonesia, Republic of Korea, Malaysia, Thailand

and Netherland was presented in this meeting. The irradiation capability for the (n, γ) method in eachreactor is summarized in Table 1 to Table 3, and presented documents are attached in Appendix A.4.

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JAEA-Review 2012-030(1) Poland
Radioisotope and 99Mo Production in Research Reactor MARIA. Experience and Plans for Future
J.Jaroszewicz (National Center for Nuclear Research [NCBJ])NCBJ is willing to provide solution ensuring sustainable production and distribution of 99Mo inboth mid and long terms based on HEU & LEU irradiation targets in MARIA Research Reactor using
the new Laboratory for High Activity Radionuclides (LHAR) and a new processing facility.
As a first step, the commercial irradiation of the HEU targets was started from March 2010. As asecond step, the development of new innovative production technology for LEU-based 99Mo production
and the feasibility study of the LHAR are ongoing. Production capacity is expected at about 1000 Ci (6days 99Mo) using 2 irradiation positions (rigs) per cycle, and maximum 30 cycles/y is expected. As analternative 99Mo production, MARIA Research Reactor has an experience of 99Mo production by the (n,γ) method in 1980’s. Therefore, MARIA Research Reactor is ready to restart the irradiation of 98Mo
targets, and to participate the program of 99Mo production by the (n, γ) method.
(2) Kazakhstan

Production of 99Mo as well as other radioactive isotopes for medical, scientific and industrial
purposes in Kazakhstan is performed by Institute of Nuclear Physics belonging to the National
Nuclear Center Republic of Kazakhstan.
From 1998 year the natural molybdenum oxide is used for production of (n, γ) 99Mo. For
effective utilisation of activated 99Mo with low specific activity (0,3-1,2 Ci/g) and production ofcentral and portable types of 99mTc gel-generator the gel-technology developed by Australianscientists was improved and original technological equipment was developed. The main steps of theInstitute’s activity in 99Mo and 99mTc gel-generators production are:
 2000-2001 - we had preclinical and clinical testing, and registration of “Sodium pertechnetate99m

Tc, injection” from central 99Mo/99mTc gel-generator.

 2002 - we started commercial supply of ready to use 99mTc solution to Almaty hospitals.
 2009-2011 - we passed preclinical, clinical trials and registration of “Sodium pertechnetate99m

Tc injection” from portable gel-generator.

From 2012 year we started the commercial production 18 GBq portable Tc gel-generators to the
local clinics. The quality of 99mTc solution is meeting to pharmacopoeia requirement, thus content
of 99Mo is not more than 0,02%, other gamma-emitting impurities - not more than 0,002%,
radiochemical purity is not 99,9%.
To increase 99Mo specific activity and optimize the irradiation rig for large quantity ofmolybdenum the research activity is continued. Besides the three joint experiments with JAEA
scientists were done at INP with using of 99Mo produced at WWR-K nuclear reactor from natural
molybdenum oxide to compare PZC and 99Mo-Zr gel methods and to demonstrate the possibility touse both methods for utilization of Mo with low specific activity for production of Tc solution withhigh quality.

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The INP is interested in international cooperation in the field of activated 99Mo production,
irradiation of Mo targets and technology transfer for production of 99mTc gel-generator.

Experience of Institute of Nuclear Physics on 99mTc Gel-generator Production
E.T.Chakrova (Institute of Nuclear Physics [INP NNC])
Institute of Nuclear Physics has more than 10 years experience on production of both central
and portable types of 99mTc gel-generators. Natural molybdenum oxide is irradiated in central
channels of WWR-K with thermal flax 0.8-2.1 n·s-1·cm-2 during 72-120 hours for activation of99

Mo. According to neutron spectrum of in central channels of WWR-K reactor more that 60% of

total 99Mo activity produced due to neutrons with energy higher then 0.5 eV. For production 99mTc
with high volume activity the gel-technology is used for the direct conversion of neutron irradiatedMo into insoluble zirconium polymolybdate gel that contains 25-30 wt.% of molybdenum. The gel
is predominantly a cation exchanger with an open structure which allows free diffusion ofpertechnetate anion under simple elution by saline. For prevention of molybdenum break-throughadditional Al2O3 column is needed.
High quality of 99mTc solution, small quantity of radioactive waste, no gas waste, no high level
radioactive waste are advantages of using (n, γ) 99Mo and gel-technology.
(3) Indonesia
Current Status on Production of 99Mo in Indonesia
A.Mutalib (National Nuclear Energy Agency [BATAN])
Production of99

Mo for commercial purposes in Indonesia is performed by PT Batan Teknologi, astate-owned company under supervision of Ministry of State Enterprises. This production is mainly
carried out for fission product 99Mo using HEU targets and at present PT Batan Teknologi plans to carryout conversion HEU target to metal foil LEU target combined with the modified Cintichem process.
This technology was developed through a joint-research between CRR (Center for Radioisotopes andRadiopharmaceuticals - BATAN) and ANL, USA. Most fission product 99Mo produced is shipped forexport and some of which are used for preparing 99Mo/99mTc generators distributed for domestic
hospitals. Although BATAN is a nuclear technology R&D institute, which is responsible directly to the
President of Republic of Indonesia and most of BATAN’s activities are under coordination of Ministryof Research and Technology, it has been assigned to provide its facility as a back-up if the production
facility of PT Batan Teknologi is closed for maintenance. Regarding the (n, γ) method for producing
99

Mo, even though since last 20 years ago BATAN has developed 99Mo/99mTc gel generator, from acouple of years ago until now the CRR has focused on the development and implementation of neutron
irradiated natural molybdenum for preparing 99mTc through two methods, i.e. PZC-based 99Mo/99mTc
generator and MEK extractor combined with two alumina columns, respectively. The first method wasdeveloped through a joint cooperation between CRR and Kaken Co., Japan and the generators were

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used satisfactorily for clinical trials in several hospitals in Jakarta and Bandung. Recent
joint-experiment between CRR and JAEA/Chiyoda Co., Japan has successfully developed PZC-based
99

Mo/99mTc generator loading with 10 Ci 99Mo, in which PZC material was supplied by Kaken Co.

Currently, the CRR has developed its own zirconium polymer for production of PZC-based 99Mo/99mTc
generators to be distributed to the hospitals outside Jakarta.

The extraction method combined with two

alumina columns, i.e base and acid alumina columns, respectively, is being developed through a
joint-cooperation between CRR and JAEA/Chiyoda Co., Japan. CRR prefers implementing this method
for preparing 99mTc to be labeled with several radiopharmaceutical kits produced by CRR itself as
injectable radiopharmaceuticals needed by several hospitals both in Jakarta and near the CRR’s
production facility in Serpong.
(4) Republic of Korea
Current Status of Research on (n,γ) Mo/99m Tc Generator as Mid-term Approach
U.J.Park (Korea Atomic Energy Research Institute [KAERI])In HANARO, radioisotopes for industrial & medical purpose are produced, and research &development for various radioisotopes are carried out. Major products include 192Ir for NDT, 131I fortreatment and diagnosis of thyroid cancer and 99Mo/99mTc Generator for imaging diagnosis of cancer.
Due to world-wide unstable supply of 99Mo, a technology to produce (n, γ) 99Mo generator in HANARO
had been developed as a short term countermeasure.
In case of irradiation for 60g MoO3, 99Mo activity after six days Ci was estimated by Monte Carlosimulation as follows;
Irradiation position

IR-2

OR-3

IP-15

Natural Mo

33 Ci/target

18 Ci/target

8 Ci/target

98% enriched Mo

91 Ci/target

58 Ci/target

30 Ci/target

However, routine supply of 99Mo for the abroad is not realistic because the irradiation position isalways limited for domestic RI supply. KAERI would be pleased to supply small quantities of 99Mo forresearch purpose for the limited time period.
As a long term countermeasure, KAERI is trying to build a new fully dedicated isotope reactor inwhich Fission 99Mo will be produced.
(5) Malaysia
Strategies for 99Mo Supply Security in Malaysia
R.Dahalan (Malaysian Nuclear Agency)
The medical technology division started producing

99m

Tc since 1986 by the neutron activation with98

the irradiation of natural molybdenum trioxide ( Mo) to produce 99Mo. The 99mTc was extracted dailyusing the methyl ethyl ketone extraction technique and distributed to the main nuclear medicine facility-7-

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in the Kuala Lumpur hospital. Those, such as limitation in yield, logistic challenges and expansion of
nuclear medicine applications and centers in Malaysia lead to the establishment of the highly portableweekly 99Mo/99mTc chromatographic generator production facility in the Malaysian Nuclear Agency
utilizing imported fission 99Mo. This 99Mo/99mTc generator production facility is currently capable ofsupplying the total weekly demand for 99mTc at all the nuclear medicine facilities in Malaysia. In
addition, there are 3 other alternative 99Mo/99mTc generators imported by various local companies asalternative to those production in Nuclear Malaysia. However, the current world shortage in fission
99

Mo supply has also affected the routine supply of 99mTc in Malaysia and to date we are experiencing

frequent disruptions and delays in our 99Mo supply due to logistics and various other reasons. It is
expected that the local demand for 99Mo/99mTc generators will also increase in the future which may
aggravate the current difficulties thus prompting alternative strategies in 99Mo/99mTc security of supply.
By considering the plan for the Malaysian Nuclear Agency research reactor upgrade in this situation, we
initiated the review of the various modality of 99Mo production, including fission 99Mo production using
low enriched uranium and the possibility of neutron activation 99Mo with its alternative 99Mo/99mTc
generators. There are currently, medium energy cyclotron in Malaysia and reported the possibility of
producing 99mTc directly using these cyclotrons. This is also being considered as viable options.
(6) Thailand99

Mo/99mTc-generator and 99mTc-Radiopharmaceutical Service in Thailand
J.Sangsuriyan (Thailand Institute Of Nuclear Technology [TINT])

99m

Tc radiopharmaceuticals had been used for diagnostic procedure in Thailand for more than 30

years. At least 6 nuclear medicine centers, 5 in Bangkok city and 1 out side, had used 99mTc
radiopharmaceuticals for routine application by using imported Tc-genertor since 1978, and the numberof nuclear medical centers that routinely used 99mTc was increased into 13 centers in 1986 and morethan 20 centers in 2011, most of them were public or university hospitals. During 1985 to 1996 ourinstitute had involved in the supply of 99mTc to some nuclear medicine centers in Bangkok city area by
our MEK extraction process after irradiation of MoO3 in research reactor. The process gave production
yield of 2.2Ci/batch from 100 g MoO3 target and 48 h irradiation time with neutron flux of 2x1013 and99m

Tc (TcO4) was distributed to nuclear medicine center in the morning every day. However, after thattime,the 99mTc production was stopped due to the limit capability of research reactor and the user
intended

to

use

99m

Tc

from

Tc-generator(imported).

From

rapid

increasing

demandofTc-radiopharmaceuticals, we have also developed our facilities for Tc-kits production laboratory since
1984 .Our Tc-kit products were covered almost diagnostic procedures currently used, eg., MDP, MIBI,
ECD, DMSA. MAG3, DTPA, DISIDA etc. More than 50% of kits used in all nuclear medicine centerswere supplied by our institute, and the kits will be compounded on site by using 99mTc from theirTc-generators. All Tc-generator supply in Thailand now are imported product and the quantity of
Tc-generator is approximately 500-625 Ci 99Mo/year (base on 2010, informal data) and most of
Tc-genertor size used is 0.5Ci and some centers (< 5 center) use 0.9Ci Tc-generator for 1or 2 week. In-8-

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the last 10 year, nearly 50% of Tc-radiopharmaceutical service in Bangkok city had changed from onsite compounding of Tc-kit into centralized radiopharmacy that supply 99mTc radiopharmaceuticalfinished product as unit dose(ready to inject) directly to nuclear medicine center on time. There are twocentralized radiopharmacy currently services in Bangkok city; one is our institute that provide thisservice since the last two years, the other one is the private company that shares greater than 80% of
total 99mTc unit does services in central Bangkok area.
Under the shortage of world 99Mo supply from 2007 while the country demand of 99mTc go up year
by year, our institute, by isotope group, has now paid much attention in this issue and is looking forpotential technology that can serve stable availability of 99mTc radiopharmaceutical for local use. A
project for production and supply of 99mTc isotope or 99Mo/99mTc generator may be initiated in the next
fiscal year. Project proposal will be submitted to our government for approval and funding.
(7) Netherlands
Current status of 99Mo production in HFR
F.Moet (Nuclear Research and Consultancy Group [NRG])
The Nuclear Research and Consultancy Group (NRG) is the Dutch centre of expertise for Nuclear
Technology. NRG has a staff of around 360 and is located in Petten and Arnhem in the Netherlands. The
Petten site is the most significant and is occupied by 4 main organizations; NRG, its mother company
ECN (the Dutch Energy Research Centre), the European Commissions Institute for Energy (EC IE) and
Covidien the global Radiopharmaceutical Company.
The Petten site hosts a unique set of nuclear infrastructure including the HFR Research Reactor
owned by the EC IE and operated by NRG, a low flux Reactor, extensive Hot Cell Laboratories,
Radiochemical Processing Laboratories, Radiological Laboratories and an Alpha Laboratory owned andoperated by NRG. Additionally, the Molybdenum Production Facility, is owned by ECN, managed by
NRG and operated by Covidien. These facilities jointly produce a large percentage of the world demandfor isotopes for medical and industrial applications.
Aside from the production of radioisotopes and the provision of irradiation services, the main
activities of NRG are R&D in Nuclear Energy for commercial and governmental customers, thisincluding Qualification Testing for Nuclear Fuels and Structural Materials. NRG also provides a rangeof teaching and consultancy services.
The most important medical isotope produced at Petten by NRG is 99Mo, the mother isotope of
99m

Tc, which is produced from the irradiation of Uranium targets. NRG sits at the heart of the process

managing the Uranium Cycle for 99Mo production, by purchasing the Uranium targets, performing thetarget irradiations in the HFR, transporting the active targets, managing the facilities where Target
processing is performed and finally managing the waste produced from the process. The HFR isrecognized as being one of the largest producers of 99Mo in the world, producing in recent years around
33% of world demand. The HFR irradiates targets for 2 99Mo processors, Covidien and IRE (Belgium)
and is at the heart of European 99Mo supply network.

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In addition to 99Mo, NRG produces a wide range of other important isotopes, including 192Ir, 89Sr,
153

Sm, 131I, 90Y, 198Au, 186Re, 169Er, 32P, 125I, 169Yb, 210Ho, 170Tm, 195m Pt. These isotopes are supplied to awide range of customers located around the world and many of these isotopes, including 99Mo aresupplied to the Japanese market; although many are delivered by indirect routes through otherorganizations.
The large scale and extensive range of isotopes supplied by NRG from the HFR are due to the highlevel of availability of the HFR (290 days/year), which routinely operates with a power of 45MW and atypical maximum Thermal Flux of 2.6 x 1018n/m2/s. The HFR Core has a fueled height of 60 cm and a
total of 30 usable irradiation channels. In normal operation, 7 channels are used for 99Mo production,
but at peak and during periods of high market demand, 11 channels can be run in parallel. An importantfeature of the HFR is the ability to load and unload target material during the running cycle. This allows
irradiations to be started and stopped at any point in time, and also this allows product to be delivered at
the exact time requested by the customers.
While the HFR has no recent history in the neutron activation of 98Mo to produce 99Mo, theexisting production facilities in the HFR are well suited to 98Mo irradiation, with re-loadable channels
available with a typical peak thermal flux of 1.6 x 1018n/m2/s. A typical facility of this type has 18
irradiation positions in the channel, each for carrying an irradiation can of 25mm diameter x 80mm
height (can volume capacity: 30cm3 per can). NRG is interested to discuss the technical possibilities of(test) irradiating 98Mo including radio chemistry and transportation issues.
While NRG has no present experience in the shipping and distribution of 99Mo produced from a
98

Mo target, NRG has extensive experience in shipping active Uranium targets and other high activity

shipments worldwide, including to Japan. NRG is planning the medium term replacement of the HFRwith a new high power research reactor with extensive isotope production capabilities called PALLAS.

4.

Discussion for cooperation
Based on the information exchange, an importance of 99Mo production by (n, γ) method wasrecognized by each country, and discussion for the cooperation was carried out on the proposed four
points that were target, transportation, certification and refinement.
(1) Target
Irradiation target should be standardized in pellet type or powder type to irradiate easily in many
research reactors for enhancement of the (n, γ) method. Here, the high purity requirement is not acommon interest, because it is different from country by country. Also, the enriched 98Mo pellet is
not addressed as a standard one owing to the much higher price than natural one. Through thediscussion in this meeting, following items were agreed;- As the first step in cooperation, irradiation test and evaluation by each country is indispensable by- 10 -

JAEA-Review
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use of a common target.
- From the discussion of capsule design in each research reactor, maximum five high density
pellets (18 mm diameter and 10 mm high) were selected as a common target.
- JAEA will distribute the common target to each country by the around October 2012.
(2) Transportation
Transportation from reactor facility to processing facility should be solid condition (MoO3),
because the transportation by the MoO3 solution is not easy and necessary for additional
reprocessing facility in the reactor side. More details will be discussed after the cooperative
irradiation test using the common target.
(3) Certification
For the certification of validity of the 99Mo product, the best way is to conduct the target irradiationin each reactor, and return it back to Japan to confirm. This plan will be discussed in the next
meeting.
(4) Refinement
Commercialization of the (n, γ) method will contain newly developed technology, therefore issuesof intellectual property or patent should be considered. These points will be discussed with
continually.

5.

Conclusion
This meeting was held aiming at the information exchange of current status and future plan for the99

Mo production by the (n, γ) method, and also aiming at the discussion on cooperation among research

reactors to promote the (n, γ) method. Through the information exchange and discussion, every country
had an interest to produce the 99Mo by the (n, γ) method, and it was recognized that to push forward thedevelopment of 99Mo production by (n, γ) method is necessary for the future steady supply of 99Mo.
Moreover, an irradiation test using the high density MoO3 pellet developed by the JAEA was proposedfrom a view point of the merit for the common irradiation target in each research and test reactor, andthe MoO3 pellet irradiation is to be done in each reactor after JAEA distributes irradiation targets. JAEA
proposed the next meeting, and it will be held on February or March in 2013 in Japan.

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Table 1

Irradiation capability for (n, γ) method in each reactor(1)

CountryJapan

PolandKazakhstan

Korea

Reactor

JMTR

MARIA

WWR-K

HANARO

50

30

6

30

30

5-10

7-10

29

Available

Available

Available- Thermal power (MW)- Operating days/cycle(day)
Irradiation facility- Status

HR-1

Available

HR-2
Under
construction

Neutron Flux of Irradiation Facility- Irradiation hole

D-5

M-9

- Thermal neutron

1.1 (max.)

3.5 (max.)

2.5 (max.)

0.81 (av.)

2.6 (av.)

1.8 (av.)- Fast neutron

0.088 (max.)

0.93 (max.)

0.022 (max.)

18

0.067 (av.)

0.70 (av.)

0.017 (av.)

32, 15032, 150

26, 120

26, 120

Al, SUSAl, SUS

Al

Al

Al

6

7

7-10

7

6.3*

1.6

0.8-1.2

2.2

18

2

x10 (n/m /s)

2

x10 (n/m /s)

IP-15

1.1 (max.)

1.8 (max.)
1.5 (av.)

Capsule- Outer diameter, length(mm)- Inner diameter, length(mm)- Capsule material- Irradiation time (days)
99

25, 100
25, 87?, 132

38,13031, 85

Mo production rate of natural MoO3

- 99Mo production rate
(Ci/g-Mo)

1.3*(at end of irradiation)*: Under estimation (tentative value)- 12 -

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Table 2

Irradiation capability for (n, γ) method in each reactor(2)

Country

Indonesia

Malaysia

Thailand

Reactor

RSG-GAS

TRIGA-PUSPATI

TRR-1/M130

0.75-14 and 12 days

6 h/day , 75 h/week- Thermal power (MW)- Operating days/cycle(day)
Irradiation facility

1.2(Max 2 MW)
46 h/week

Central Thimble- Status

Available

Available

Almost occupied

CT

Central hold (CT)

1.3 (av.)

0.18

0. 24

25.4, 5028, 127

22, 120

23, 21024, 10118, 118

Al

Al

Al

4 or 12

75 h

0.5

0.02

Neutron Flux of Irradiation Facility- Irradiation hole
- Thermal neutronx1018 (n/m2/s)

Central Irradiation
Position (CIP)- Fast neutronx1018 (n/m2/s)
Capsule- Outer diameter, length(mm)- Inner diameter, length(mm)- Capsule material- Irradiation time (days)
99

Mo production rate of natural MoO3

99

- Mo production rate
(Ci/g-Mo)(at end of irradiation)- 13 -

0.03

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Table 3

Irradiation capability for (n, γ) method in each reactor(3)

Country

Netherlands

Belgium

Reactor

HFR

BR-2*

- Thermal power (MW)

45

100

28-29

21-28

- Operating days/cycle(day)
Irradiation facility- Status

DG thimble

RA basket

Available

Available

Available

Neutron Flux of Irradiation Facility- Irradiation hole
- Thermal neutronx1018 (n/m2/s)

18 positions

1.6(max.)- Fast neutronx1018 (n/m2/s)

4.0(max.)

10(max.)

2.9

(av.)

7.1

(av.)

1.8

(max.)

2.0(max.)

1.3

(av.)

1.4

(av.)

Capsule- Outer diameter, length22, 6922, 69- Capsule material

AlAl- Irradiation time (days)

7

21 or 28

2.5

20

(mm)

25, 80- Inner diameter, length
(mm)

99

Mo production rate of natural MoO3

- 99Mo production rate
(Ci/g-Mo)

2.6(at end of irradiation)*: Information through e-mail- 14 -

JAEA-Review
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Appendix
A.1 List of participants
Poland,National Center for Nuclear Research (NCBJ)
Grzegorz Krzysztoszek, Janusz Jaroszewicz

Kazakhstan, National Nuclear Center (NNC)
Erlan G. Batyrbekov, Yelena Chakrova
Indonesia,

National Nuclear Energy Agency (BATAN)
Djarot S. Wisnubroto, Abdul Mutalib

Korea,

Korea Atomic Energy Research Institute (KAERI)
Ul Jae Park, Cheol Park

Malaysia,

Malaysian Nuclear Agency
Rehir Dahalan

Thailand,Thailand Institute Of Nuclear Technology (TINT)
Somporn Chongkum, Jatupol Sangsuriyan

Netherlands, Nuclear Research and Consultancy Group (NRG)
Frits MoetJapan,

Chiyoda Technol Co.
Nobuhiro Takeuchi, Norihito Sato, Masakazu Tanase, Akio Ohta

Japan,

FUJIFILM RI Pharma Co. Ltd.
Masaaki Kobayashi, Yasumasa Morikawa, Asaki Yamamoto

Japan,

Japan Atomic Energy Agency (JAEA)
Hiroshi Kawamura, Masahide Suzuki, Masahiro Ishihara, Masanori Kaminaga
Tsuguo Genka, Etsuo Ishitsuka, Kunihiko Tsuchiya, Takashi Sekiya<Observer>
Japan,

Japan Radioisotope Association (JRIA)
Yoshihide Nakamura- 15 -

JAEA-Review
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Fig A-1

Photograph of participants.

- 16 -

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A.2 Program
Specialist Meeting on Mo-99 Production by (n,gamma) Method, March 9-10, 2012, Yurakucho Asahi Hall, Tokyo, JapanPROGRAM
Friday 9th Mar., 2012
9:30 - 9:40

Purpose of the specialist meeting

M.Ishihara

JAEA

9:40 - 10:00

Affects of earthquake and re-operation of JMTR

N.Hori, M.Kaminaga, T.Kusunoki, M.Ishihara, Y.Komori,
JAEA
M.Suzuki, H.Kawamura

Chairman: M.Ishihara

10:00 - 10:25

Needs and current status of Mo-99/Tc-99m production in
Japan

T.Genka

JAEA

10:25 - 10:50

Status of 99Mo-99mTc Production Development by (n,g)
Reaction in Japan

K.Tsuchiya

JAEA

M.Tanase

Chiyoda Technol
Co.

10:50 - 11:15

R&D in Japan on Extraction and Concentration of 99mTc- Preliminary Study using Re Instead of 99mTc -11:15 - 11:35

Technical Request/Standard for 99mTc Solution for Medical
Use in Japan

M.Kobayashi

FUJIFILM RI
Pharma Co. Ltd.

11:35 - 12:00

Experience of Institute of Nuclear Physicson Tc-99m gelgenerator production

E.T.Chakrova

National Nuclear
Center in
Kazakhstan

12:00 - 13:00

LUNCH

Session 2 : Comparison with each country

Chairman: M.Kaminaga

13:00 - 14:00

Brief introduction of reactor and irradiation facilities for RI
production

Poland, Indonesia, Kazakhstan, Korea, Malaysia,
Thailand(about 10 min each)

14:00 - 14:50

Target and capsule (or rig) for neutron irradiation

Poland, Indonesia, Kazakhstan, Korea, Malaysia,
Thailand(about 5 min each)

14:50 - 15:30

Neutron irradiation capability

Poland, Indonesia, Kazakhstan, Korea, Malaysia,
Thailand(about 5 min each)

15:30 - 15:50

BREAK

15:50 - 16:30

Cask, Transportaion

Poland, Indonesia, Kazakhstan, Korea, Malaysia,
Thailand(about 5 min each)

16:30 - 17:30

Summary of first day(M.Ishihara)- 17 -

JAEA-Review
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Saturday 10th Mar., 2012

9:30 - 9:45

Introduction of first day's summary(M.Ishihara)

Chairman: M.Suzuki

Session 3 : Current status of Mo-99 production (1)9:45 - 10:10

Radioisotope and Mo-99 Production in Research Reactor
MARIA. Experience and Plans for Future

J.Jaroszewicz

Poland National
Center for Nuclear
Research

10:10 - 10:35

Current Status on Production of 99Mo in Indonesia

A.Mutalib

BATAN, Indonesia

E.G.Batyrbekov

National Nuclear
Center in
Kazakhstan

Ul Jae ParkKAERI, Korea

10:35 - 11:00

11:00 - 11:25

11:25 - 13:00

Mid-term Approach

COMMEMORATIVE PHOTOGRAPH and LUNCH

Chairman: M.Kaminaga

Session 4 : Current status of Mo-99 production (2)13:00 - 13:25

Strategies for Mo-99 Supply Security in Malaysia

R.Dahalan

Malaysian Nuclear
Agency13:25 - 13:50

99Mo/99mTc-generator and 99mTc-Radiopharmaceutical
Service in Thailand

J.Sangsuriyan

TINT, Thailand

13:50 - 14:15

Current status of Mo-99 production in HFR

F.Moet

NRG/Petten,
Netherlands

14:15 - 14:40

BREAK

Chairman: M.Ishihara

Session 5 : Discussion of cooperation

14:40 - 16:20

How to cooperate steady supply of (n,gamma) Mo-99

16:20 - 16:40

BREAK

16:40 - 17:30

Summary (Confirmation of meeting minutes)

All

All

- 18 -

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A.3 Presentation documents for approach of 99Mo production in Japan- 19 -

This is a blank page.

JAEA-Review 2012-030

Specialist Meeting on Mo-99 Production by (n,gamma) Method,
March 9-10, 2012, Tokyo, Japan

Affects of earthquake andre-operation of JMTR
N.Hori, M.Kaminaga, T.Kusunoki, M.Ishihara, Y.Komori,
M.Suzuki, H.Kawamura
March 9, 2012
Neutron Irradiation and Testing Reactor Center
Oarai Research and Development Center
Japan Atomic Energy Agency (JAEA)

Contents
1. Outline of JMTR
2. JMTR re-operation schedule in June, 2011
3. Affects of the 2011 off the Pacific coast of
Tohoku earthquake
4. JMTR re-operation schedule
5. Installation of new irradiation and postirradiation facilities
6. Summary- 21 1JAEA-Review 2012-030

1. Outline of JMTR
PurposeAncillaryfacilities room

The JMTR was constructed toperform irradiation tests for LWR
fuels, materials to establish
domestic technology for developing
nuclear power plants, and also to
produce radio isotopes, and for the
purpose of education & training.

Reactorbuilding

Canal (Waterway)• Reactor thermal : 50 MW
power• Fast neutron flux : 4 x 1018 (n/m2/s)(Maximum)• Thermal neutron : 4 x 1018 (n/m2/s)
flux (Maximum)

Reactor and
Irradiation Facilities

Hot Laboratory
Concrete Cell
: 8
Microscope Lead Cell : 4
Lead Cell: 7
Steel Cell
: 5
X-ray Microscopic
analyzer : 1

Construction began : 1965 Apr.
First Criticality: 1968 Mar.
For user operation : 1970 Sep.
to 2006 Aug.

Major feature• One of the high neutron flux Materials Testing Reactor in the world• Large irradiation area in the core region for various irradiation tests• Flexible reactor core configuration allows various irradiation facilities
installation to the reactor core
• The reactor building is connected to the hot laboratory by a canal for
PIEs for fuels and materials.

16

2. JMTR re-operation schedule in June, 2011
JMTR re-operation schedule was changed to early in November 2011, in order to carry out reactor
components inspection and seismic influence analysis. The change report of the JMTR operation plan was
submitted to the Ministry of Education, Culture, Sports, Science and Technology (MEXT), based on the
Nuclear Reactor Regulation Law (June 21, 2011).
2011

JFY

2012

Jun. Jul. Aug. Sep. Oct. Nov. Dec. Jan. Feb. Mar. Apr. May Jun. Jul. Aug. Sep. Oct. Nov. Dec.

Items

Oarai R&D center synthesis training(JMTR station black out was assumed.)

JMTR operation plan
Periodical independent inspection
etc.
based on maintenance plan.

Operator training assuming
JMTR station black out.

Seismic influence
evaluation
Seismic
influence
analysis

Reactorcomponents
inspection
Response
analysis

167

166

168

The operation plan in JFY 2012 is omitted.

IASCC capsule loading, characteristic test and aging .
Fuel loading andcharacteristic measurement

Reactor components inspection
Time history response analysis for reactor
components, Stress analysis for reactor
building
Presentation of report
to MEXT
Seismic response analysis of JMTR reactor building basedon observation wave and decision of floor response
spectrum (FRS)26- 22 2

JAEA-Review 2012-030

3.1 The restoration situation of the majordisaster parts related to JMTR
Disaster part

Before repair

After repair

JMTR attachedmanagementbuilding wall

October 27, 2011Land subsidenceof JMTR reactor
building outside

October 27, 2011

Floor crack of
exhaust blowerroom
Hot lab. concrete
cell shielding
door electric lock
distortion

Completion of repair

October 7, 2011

December 22, 2011
Distortion part36

3.2 Seismic influence evaluation of JMTR after
the 2011 off the Pacific coast of Tohoku Earthquake
Based on the reactor components inspection and the seismic response analysis, integrity of the JMTR willbe identified for re-operation.

Seismic response analysis

Reactor components inspection
Facility or components

Inspection results

Main part of the reactor
Pressure vessel,
Reactor pool etc.

No damage was observed byvisual inspection.

Cooling systems
Primary cooling system
Emergency cooling system

No damage was observed byvisual inspection, pressurized test,
and operation test at rated flow.

Instrumentation system
No abnormality was observed byProcess instrumentation
operation test.
Nuclear instrumentation etc.
Reactor control system
Control rod drive
mechanism

No abnormality was observed byoperation test after readjustmentof the drive mechanism.

Reactor building air supply
No damage was observed byand exhaust system
visual inspection and operation
Exhaust stuck, Blowers etc. test.
Emergency power supply
Diesel generators

No abnormality was observed byoperation test.

Irradiation facilities

No abnormality was observed bypower supply inspection andleakage check etc.

Buildings

Truss beams bend of the roof wasobserved during visual inspection
on September 27, 2011.

【Observation record of the earthquake in the JMTR
reactor building】

The maximum acceleration of the observed earthquake
motion was 218 to 249 gal on the 2nd and 3rd basementlevel of the building. A part of observed acceleration had
exceeded the static seismic force for a design(Horizontal:0.3G (294 gal), Vertical:0.15G (147 gal)).
Since the acceleration observed had exceeded the
maximum response acceleration assumed at the time ofthe design, the report of the following matters was
requested from MEXT on September 20, 2011.
1. Detailed verification result of the observed earthquake
data for seismic response analysis.
2. Seismic influence evaluation results by the seismic
response analysis etc.

【Current status of seismic response analysis】

Based on observed record at the 2011 off the Pacificcoast of Tohoku Earthquake, the earthquake-proof
evaluation (check of generating stress being in an elastic
region) is carried out for buildings and reactor components
by the seismic response analysis.
As for the reactor building, while seismic response
analysis including re-examination of the analytical model
was carried out so that 3.11 seismic observation records
could be reproduced, and obtained the floor response
spectrum (FRS) etc. by the mass system model. As a
46
result, the building was in the elastic range in general.

- 23 3

JAEA-Review 2012-030

3.3 Truss beams bend of the JMTR
reactor building roof
Truss beams bend (including deformation of lattice, cracks of concrete) of the reactor building roof were
found on September 27, 2011, during detailed visual inspection of the JMTR reactor building.
排気筒

約44m

原子炉建家

炉
建
家
真

30t/5t クレーン

ホットラボ建家
炉
真

原子炉
制御室
キャットウォーク

コンクリートセル

サービスエリア
カナル室

GL

炉 プール
カナル

カナル

カナル

圧力容器

B３FL

JMTR reactor
Building
cross-section

制御棒駆動装置室
B４FL

 In order to carry out detailed investigation of the JMTR reactor
building roof, re-operation schedule was changed. (The changereport was submitted to MEXT on October 26, 2011.)
 From November, 2011 to January, 2012, detailed investigation of the
JMTR reactor building roof was carried out.
 Position measurement by a laser displacement meter, size
measurement, visual investigation of the components surface etc.
were performed.
 As a result of the investigation, truss beams bend, deformation oflattices, cracks of concrete were considered to have been caused by
the on-site adjustment at the time of construction, the influence of theearthquake, the influences of cyclic stress and these complex factors.
 Repairing method is under consideration based on the detailed
investigation result and stress analysis results of the roof.
5

Truss beam bend found at the JMTR
reactor building roof

6

4.1 JMTR re-operation, seismic influence
evaluation and roof repair schedule
JMTR re-operation schedule was changed to October 2012, in order to carry out detailed inspection andrepair of the JMTR reactor building roof. The change report of the JMTR operation plan was submitted
to MEXT, based on the Nuclear Reactor Regulation Law (October 26, 2011).
JFY

2012

2011

Jun. Jul. Aug. Sep. Oct. Nov. Dec. Jan. Feb. Mar. Apr. May Jun. Jul. Aug. Sep. Oct. Nov. Dec. Jan. Feb. Mar.

Items

JMTR operation plan etc.

Fuel loading andcharacteristic measurement

Oarai R&D center synthesis training(JMTR station black out was assumed.)

166

167

168

169

Periodical independent inspection based on maintenance plan.

Reactor components
inspection, Repair
Seismic
influence
analysis

Seismic influence evaluation

Operator training assuming
JMTR station black out.

Inspection

RepairResponse
analysis

Periodical
inspection

Reactor components
Detailed inspection and investigation
inspection
Examination of the roof repair method
Time history response analysis for reactor Repair
Repaircomponents, Stress analysis for reactor
preparationbuilding
work
Presentation of report
to MEXT
Stress analysis of the roof
Seismic response analysis of JMTR reactor building basedon observation wave and decision of floor response
spectrum (FRS)

Validity check of the seismic
influence evaluation

66- 24 4

JAEA-Review 2012-030

4.2 Schedule for JMTR re-operation
‘05

‘06

‘07

‘08

‘09

’10

1st

Periods of JAEA

‘11

‘12

‘13

‘14

‘15

‘16

‘17

2nd

‘18

‘19

3rd

Operation of JMTR
Seismic influence evaluation
after the 2011 off the Pacificcoast of Tohoku Earthquake

Truss beams bend of the JMTR reactor
building roof were found in September, 2011.
In order to investigate and repair the roof ofthe building, re-operation of the JMTR was
postponed to October 2012.

The 2011 off the Pacificcoast of Tohoku Earthquake(including analysis and repair)

Installation of new irradiation
facilities

Refurbishment of JMTR

(Refurbishment budget isabout 6 billion yen.)

Project for Nuclear andIndustrial Safety Agency

The manufacture and installation of components
regarding refurbishment of JMTR had been completed.
Only characteristic tests were delayed by influence ofthe earthquake disaster about two months.

Acquisition of the project- LWR material (for BWR)
- LWR fuel power rump
- LWR fuel loop
- Improvement of Hot laboratory(Conceptual design)

The irradiation tests for irradiation embrittlement, IASCC,
etc. will be carried out . These tests will contribute tosafety research of the present light water reactor material.

（Under consideration)

Project for Advanced research
infrastructure

Irradiation facilities are
installed according to useneeds and users fund.

Acquisition of the project- Hydraulic rabbit system- Water loop for LWR condition- Advanced PIE device
installation to Hot laboratory

R&D for 99Mo production by (n,)method will be carried out by using
the hydraulic rabbit system.

76

5.1 Installation of new irradiation facilities
Concentrated control panel(Installation completed)

Load control device etc.
(Installation completed)
LWRs fuel irradiation
facility for operationaltransient
Water environment
control system(Installation completed)

Fuel irradiation facility for operationaltransient (Installation completed)

LWRs Material irradiation
facility (for PWR)
Water environment
control system for waterchemistry tests(Under installation）
Basement of thereactor building

JMTR
Pressure
vessel

LWRs Material irradiation
facility (for BWR)
Water environment
control system(2 units installation
completed）

Reactorpool
BOCA capsule(Manufacture completed)

JMTR
Reactor hallBasement of the reactor building
Green: Fuel s irradiation facility
Yellow: Materials irradiation facility

Heating unit, surge tank,etc.
Cooling tube forpreheating pipe etc.

86

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JAEA-Review 2012-030

5.2 Installation of new post-irradiation facilities
Microstructure observation
device (SEM） installation

Improvement of concrete cellshielding capability

Power manipulator for large size
capsule installation(Installation completed in JFY2009)

Concrete cell No.1 to No.3, High burn-up
fuels (Maximum 1１0GWd/t) can be treated in
the cell. (Improvement completed in JFY2009)(Installation completed in JFY2009)

Improved capsuleassembling device
installation
This device will be used mainly for
BOCA capsule assembling.
(Installation completed in JFY2010)

Master-slave manipulator
installation
Master-slave manipulators which have
improved shielding capability were installedin concrete cell No.1 and No.2.
(Installation completed in JFY 2009)

JMTR Hot Laboratory

JMTR hot laboratory cooperates with NFD (Nippon Nuclear Fuel Development Co., Ltd.) and
Tohoku university PIE facilities and can provide wide variety of PIEs requested from users.

96

5.3 Installation of new irradiation facilities by
the project of advanced research infrastructure
Materials Testing Reactor Simulator

Installation JFY 2010 to 2012

This simulator can simulate normal operation,
operational transients and accident conditionsof MTR including irradiation facilities.

This facility is designed for safety
research of LWRs. Water environment
can be control as same as BWR or
PWR.

Complex type fine texture
analyzer

Transfer

Tohoku Univ.

Hot Laboratory

Irradiatedspecimentransfer

Hot Lab.
outside JAEA
LWR

SR
1

SH
1
O
S
F

SH
2

SH
3

SR
2

Un-irradiatedspecimen

Transfer
irradiatedspecimen

RI production
Transfer
facility

Beryllium H
frame
HR

1

Installation in JFY 2010 to 2011
High precision observation :the highest magnification 20
million times

shielding
plate

LWRs water-environment
demonstration test facility

Installation JFY 2010 to 2011

JMTR
Water canalHigh grade manipulators with visual function
Installation in JFY 2010 to 2011

Mini. handling capability：Specimen size 1mm x 1mm
Max. handling capability : Large size capsule (Weight Aprox.100kg)：Al
ｱﾙ ﾐﾆｳﾑ反射体
reflector

reflector：Be
ﾍﾞﾘﾘｳﾑ反射体

High accuracy timecontrol irradiation facility
Installation JFY 2010 to 2012

Short half-life RI for medical
treatments, etc. will be manufactured.
Control of irradiation time is enabledwith high accuracy using the state-ofthe-art control technology.

High accuracy capsule temperature
control unit
Installation JFY 2010 to 2011

Irradiation environment to super-high temperature(2000 deg.C) and high precision temperature control(500 deg.C±0.5% temperature control ) are realized.

10
6- 26 6

JAEA-Review 2012-030

6. Summary
Refurbishment of JMTR was conducted from JFY2007 for 4 years and completed in end of
JFY2010.
At the time of the 2011 off the Pacific coast of Tohoku Earthquake on March 11, 2011, an
official inspection was carried out at JMTR and the reactor was maintained under stopped
condition.
Some damages due to the earthquake were found at building walls etc., just after theearthquake. There was no effect from the view point of maintaining safety of the reactor,
damages were repaired until the end of December, 2011. Integrity of the reactor components
were investigated by visual inspection, operation test, pressurized test, etc. The seismic
influence evaluation was carried out for buildings and reactor components by the seismic
response analysis. At present, no damage or abnormality was observed except the reactor
building roof.
Truss beams bend (including deformation of lattice, cracks of concrete) of the reactor building
roof were found on September 27, 2011. Therefore, JMTR re-operation schedule waschanged to October 2012, in order to carry out detailed inspection and repair of the reactor
building roof. The change report of the JMTR operation plan was submitted to MEXT (October
26, 2011).
Repairing method is under consideration based on the detailed investigation result. JMTR will
be re-started promptly after the end of the seismic influence evaluation and the repairing
work of the reactor building roof.
11

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Specialist Meeting on Mo-99 Production by (n,g) Method
March 9–10,2012 Tokyo, Japan

Needs and current status
of Mo-99/Tc-99m productionin Japan
T. Genka
JAEA
4002 Narita, Oarai, Ibaraki 311-1393, Japan

Introduction

Necessity of domestic Mo-99/Tc-99m in Japan• Tc-99m, decay product of Mo-99, is used in more than 30
kind of radiopharmaceuticals and performs about 900,000
Tc-99m-based diagnostic procedures in Japan each year
• Mo-99 is not produced in Japan although the amount ofconsumption is the second largest (14%) in the world after
the United States (44~50%)• Japan totally relies on the foreign producers, primarily on thereactors of Canada’s NRU, Netherlands’ HFR and South
Africa’s SAFARI-1• Repeated shutdown of NRU and HFR has brought
increased attention to Japan’s reliance on foreign productionof vital medical radioisotopes- 29 -

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Global Market Share by Reactor

Source: A.M. Versteegh: European Nuclear Conference Barcelona 31 May - 2 June 2010

Repeated shutdown leading to the shortage
2002 : Shutdown of the HFR reactor (42 days)
2006 : Shutdown of the NRU reactor (6 days)
2007 : Shutdown of the HFR reactor (1 month)
2008 : Cancellation of the MAPLE reactors due to technicalinconsistency in power coefficient of reactivity
2008 : Shutdown of the NRU reactor (11 days)2008 - 2009 : Shutdown of the HFR reactor (6 months
2009 - 2010 : Shutdown of the NRU reactor (11 months)
reactor vessel welding repairs2010 : Shutdown of the HFR reactor (6 months)
primary cooling pipe system repairs
2016 : Shutdown of the NRU reactor (Permanently)
operation license expire- 30 -

2

2012/6/20
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Brief history of Mo-99 production in Japan
1965 Beginning of medical application of Tc-99m in Japan
1977 Fission process using UO2 (U-235 2.6%) pellet target wasdeveloped by JAERI (presently JAEA).
Production/Trial distribution: 20 Ci/week for 20 month (1977-79)
1980 Neutron capture process using target of MoO3 pellet wasestablished by JAERI.
Production/Regular supply: 200 Ci/week for one year (1984-85)
1995 Due to budgetary cutback policy, the Cabinet decided todiscard development of such radioisotopes that were able toimport from overseas
1997 For PZC US and Japanese Patent granted to Kaken-JAERI
2001- 2006 A portable type Tc-99m generator based on PZC
adsorbent was developed under the framework of FNCA
supported by MEXT,Japan
2008 JAEA set up a subcommittee for domestic Mo-99
production (Final Report released on March 9, 2011)

Specification of JAERI Produced (n,γ) Mo-99
Target material
Irradiation

MoO3 pellet, 200 g

Thermal neutron flux
Irradiation time
Cooling time
Irradiation capsule>10E14 cm-2 s-1 in JRR-2
4 – 7 days
30 minAl

Radioactivity at processing
Facility
Materials of equipment
Chemical process
Processing time
Chemical yield

200 Ci Mo-99
Pb-cell (10 – 15 cm thick)
Quartz and glass
Dissolution in aq. NaOH
～2 h
>99%

Specific activity
Radioactivity concentration
Chemical form
Radionuclidic purity
Quantity of shipment
Major impurities> 2 Ci /g
0.5 Ci/ml
Na2MoO4 in NaOH (pH >7)>99%
200 CiNb-92m, W-188/Re-188

Extraction

Product

Source: H.Kudo et.al. IAEA-TECDOC-515 (1989)- 31 -

3

2012/6/20
JAEA-Review 2012-030

Research Reactors in FNCA Member Countries
CHINA
CARR


60 MW

Planning
Mo from LEU target

MALAYSIAINDONESIAKOREA
HANARO

RSG-GAS Pool type 24 MW (30 MW)

 15 MW 4.5E14(30 MW)
2.4E14

 Pool typeJAPAN

 1 MW 2.0E13

VIET NAM
DALAT RR
 Pooltype
 500 kW 2.1E13

2 MW
5.1E13THAILAND
TRR-1/M1

TRIGA-RTP
 TRIGA MKIITRIGA-200

 1.2 MW (2 MW)

JRR-3MJMTR Pool type

50 MW

 2.7E14(n,γ)Mo-99

 20 MWBANGLADESHTRIGA MARK II
3 MW
5.6E13

 3.1E13

TRIGA Mark III(ONRC)
 Planning 10～ 20 MW

AUSTRALIA

OPAL Mo from LEU target

Pool type

20 MW
5.0E14
D2O Refl.

Governmental initiative for the stable supply
Responding to fragile situation of Mo-99 supply, the Cabinet
Office set up a joint study group meeting* consists of relevantorganizations of governmental, public, academic, medical andindustrial sectors (Action Plan released on July 7, 2011)
•
•
•
•
•
•
•
•
•
•
•
•

Members:
Cabinet Office
Ministry of Health, Labor and Welfare (MHLW)
Ministry of Education, Culture, Sports, Science and Technology (MEXT)
Japan Atomic Energy Agency (JAEA)
National Institute of Radiological Sciences (NIRS)The Japanese Association of Nuclear Medicine
Japan Radioisotope Association (JRIA)
Nihon Medi-physics Co. Ltd.
Fujifilm RI Pharma Co. Ltd.
Hitachi-GE Nuclear Energy, Ltd.
＊ モリブデン-99m/ テクネチウム-99m のChiyoda Technol Corporation
安定供給のための官民検討会
Kaken Inc.
Observers: Atomic Energy Commission (AEC), Japan Atomic Industrial Forum, Inc. (JAIF)- 32 -

42012/6/20
JAEA-Review 2012-030

Proposed Domestic Mo-99 Production Concepts• Fission method using enriched uranium target (either HEU
or LEU) is not adopted considering the non-proliferation of
nuclear weapons in addition to facility cost and radioactive
waste management• Concepts selected by the Joint Study Group Meeting(1) Production in the research reactor (JMTR)(2) Production in a power reactor (BWR)(3) Production in a large-size accelerator (fast neutron)(4) Production in medium and mini-size cyclotrons (proton)(5) Process of refinement using activated charcoal(1) Production in the research reactor (JMTR)
・Production based on 98Mo(n, γ)99Mo reaction using MoO3
pellet target is ongoing along with the JMTR renewal program
・New hydraulic rabbit irradiation facility was installed in JMTR・Sintering procedures to fabricate high density MoO3 pellet isunder investigation
・Various extraction and concentration methods for Tc-99m
recovery using PZC, MEK and alumina are investigated undercollaboration with overseas laboratories
・ Recycling of materials of target and adsorbent is accounted
・ Estimated production capacity: 230 6day-Ci/week (～20% of
domestic consumption)
・ Part of this project is the subject of grant from MEXT・Bodies involved: JAEA-Oarai, Chiyoda Technol Corp.,
Taiyo Koko Corporation, BATAN Indonesia, INP Kazakhustan- 33 -

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2012/6/20
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Facilities for 99Mo Production in JMTR

Loading toContainer

Dismantle

EnclosureTransfer

PIE Facility

Irradiation Facility

Hydraulic rabbit system

Source: K.Tsuchiya et al ISMTR-4 2011/12/6-8, Oarai Park Hotel, Oarai, Japan(2) Production in a power reactor (BWR)
Feasibility study of Mo-99 production by 98Mo(n,γ)99Mo
reaction in a boiling water reactor (BWR) is going on
Mo-98 metallic target is inserted into the bottom of existing GE
built commercial reactors in place of neutron density
measuring instrument
After about 7days insertion for irradiation, target is sent toprocessing plant to recover Mo-99 with Ti adsorbent
Estimated production capacity: 1,000 6day-Ci/week
Clearance of various regulatory criteria on reactor control,
safety, security, health and environment
Safety measures derived from Fukushima nuclear accident
should be taken into consideration
Enterprise in charge: Hitachi-GE Nuclear Energy, Ltd.
This concept is owned jointly with GE-Hitachi N.E. in U.S.A.

- 34 -

6

2012/6/20
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Outline of BWR Power Plant

Housing for neutron-measuring instrument at the bottom

Mo98 metallic target(3) Production in a large-size accelerator
Production of Mo99 from 100Mo(n; 2n)99Mo reaction using
fast neutrons produced by the 3H(d.n)4He reaction at theFNS (Fusion Neutronics Source) of JAEA
Irradiation of enriched Mo-100 target for 198 h with neutronsof 10E13 n/(cm2 s) at En 14 MeV, produce 2 Ci/g specificactivity of Mo-99
Labeling efficiency of 99mTc-MDP was higher than 99%,
enough met the US Pharmacopeia
Expected production capacity: 126 6day-Ci/week (13% ofdomestic consumption)
This work is financially supported in part by JSTA and MEXT
Collaboration: JAEA-Tokai, Chiyoda Technol Corp.,
Fujifilm RI Pharma Co.,Ltd.

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7

2012/6/20
JAEA-Review 2012-030

Tc-99m was separated from Mo-99 by sublimation andits radionuclide purity was confirmed to be higher than
99.99%

Schematic view of the apparatus for sublimation separation of Tc-99m(F1, F2: furnaces; H1, H2, H3, H4: heaters; HI1, HI2: heat insulators)
Source: Y. Nagai et.al. Letters, J. Phys. Soc. Jpn. 80 (2011) 083201-3(4) Production
in medium and mini size cyclotrons
Medium size cyclotron (～40 MeV)
Production of Mo-99 based on 100Mo(p,pn)99Mo
reaction using water soluble molybdenum target which
can be remotely solidified within a target container. The
irradiation and labeling can be done semi-automatically.
Expected production capacity: 40 6day-Ci/week
Mini size cyclotron (～18 MeV)
On-site production of Tc-99m by 100Mo(p,2n)99mTc
reaction with a mini size cyclotron existing in the facilities
of nuclear medicine centers (>130 PET sites in Japan)
Expected production capacity: 16 6day-Ci/week
Collaboration: National Institute of Radiological Sciences(NIRS), Nihon Medi-physics Co.,Ltd.

- 36 -

8

2012/6/20
JAEA-Review 2012-030(5) Processing with activated charcoal• Investigation of separation, extraction and concentrationof Tc-99m from (n,γ) Mo-99 using activated charcoal isproceeding in pursuit of obtaining quality Tc-99m
solution for pharmaceutical manufacturing• Expected handling capacity: 500 Ci/batch/day, >1Ci/ml• Verification test using 50 Ci Mo-99 for effective
manufacturing procedure is in progress• Part of this study is supported by JSTA
• Collaboration: Kaken Inc., Chiba University

Summary
Considerations to cope with a possible future supply crisis:• Currently there are no facilities that are dedicated to
producing Mo-99 in Japan. Domestic capabilities ofsupplemental production must be secured• Priority was set on the production by (n,γ) process in thereactors taking account of production capacity,
economical viability and NTP concern
Additional countermeasures:• Utilizing alternative diagnoses, such as Tl-201 forcardiac imaging, and F-18 for bone imaging etc.
• Diversifying the route of supply and transportation forforeign imports• Encouraging and fostering regional/international
partnerships - collaborations (FNCA, CJK Congress,
Asia-Oceania Network etc.)- 37 -

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Specialist Meeting on Mo-99 Production by (n,) Method
March 9-10, 2012, Tokyo, Japan

Status of 99Mo-99mTc
Production Developmentby (n,) Reaction in Japan
K. Tsuchiya
JAEA
4002 Narita, Oarai, Ibaraki 311-1393, Japan

1. Introduction- 39 -

1

2012/6/20
JAEA-Review 2012-030

Background
The renewed JMTR will be started from the later half ofJFY2012, and it is expected to contribute to various
fields. As one of effective uses of the JMTR, JAEA has aplan to produce 99Mo by (n, ) method, a parent nuclideof 99mTc. 99mTc is most commonly used as aradiopharmaceutical in the field of nuclear medicine.
Currently, in case of Japan, supply of 99Mo depends
only on imports from foreign countries, therefore JAEA
has been performed the R&D on production method of
99Mo-99mTc in JMTR cooperating with foreign countries
and Japanese manufactures under the cooperation
programs.
1

Concept of 99Mo Production in JMTR
Advantage of JMTR
- High neutron flux
- Connection to hot laboratory

Production

Rate of consumption

Target : 20% of Domestic consumption
100%[Status][Future]ImportImport20%

100TBq/w JMTR(n, f) method(n, γ) method

235U (n,f) FP→ 99Mo →99mTc

98Mo (n,γ) 99Mo →99mTc

Cost

High

Low

Nonproliferation

Unsuitable

Suitable

Specific activity

High

Low

Impurity

High

Low

Waste

Large

Small

Japan Materials Testing Reactor (JMTR) has a plan to produce a medical
isotope of 99Mo, the parent nuclide of 99mTc by the (n, ) method, as a part of theindustrial use expansion after JMTR re-start.
2- 40 -

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2012/6/20
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R&D Items and Objective
The main R&D items for the production are as follows;(1) Fabrication of irradiation target such as thesintered MoO3 pellets,(2) Separation and concentration of 99mTc from Mo
solution,(3) Examination of 99mTc solution for a medical use,(4) Mo recycling from used Mo-Tc generator andsolution.
The status of the R&D for the production of 99Mo-99mTc
is introduced. Constructions of the irradiation and PIE
facilities in JMTR are also introduced.
3

2. R&D Items- 41 -

3

2012/6/20
JAEA-Review 2012-030

Flow Chart for 99Mo Production in JMTR
Target R&D on Irradiation Target production
Preparation

Φ32ｍｍ

Neutron
Irradiation

120~150mm

Un-irradiatedpellets

98Mo enriched

MoO3 powderΦ20mm

Inner capsule (Setting of MoO3 pellets)

Irradiatedpellets

MoO3：about 100g

(Pellet production)(Enrichment : 98%)

Recovery of
PZC

Mo solution

Recovery of Mo

Sodium molybdate(Na2MoO4)(PZC+HCl)
Transport

Transport
Used
Mo generator

R&D on Mo
recycleSolvent
99mTc
separation

Transport

Solution: 99mTc

Establishment ofradiopharmaceutical: Mo

Domestic
production of 99Mo

Shipment of 99mTc

99mTc extraction

R&D on separation, extraction andconcentration from99Mo-99mTc

Separation/Extraction/
Concentration

4

Fabrication Development for Irradiation Target
It is necessary to fabricate the high density MoO3 pellets as an irradiation target
for increasing of 99Mo production for stable supply. Then, fabrication method andcharacteristics of MoO3 have been investigated.
Investigation of Fabrication Method

Properties of MoO3

Density
Melting Point

Temperature (C)

2500
2000
1500
1000

100

Method

1atmG

818±7

MoO3
Mo9O24
Mo8O23
Mo4O11

40

MoO3+G60

O2（wt%）

80

100
O

Ref.: L.Brewer and R.H. Lamoreaux, 1980

CS

Plasma Sintering

SPS

Ed-Pas

F

P

P

F

P

P

E

P

P

Density Control

P

E

E

Product Impurity

F

E

E

Items

Sublimation:
Over 700C

L

MoO2

20

Cold press & Sintering (CS)
Spark Plasma Sintering (SPS)
Plasma Activated Sintering (Ed-Pass)

L+G

(Mo)

0
Mo

90

80

2300±1500C
2150±100OC

1300±150 ０C500
0O2 （at%）0 40 60
L: 4.692 g/cm3: 795C

Mass Fabrication
Fabrication Time
Fabrication Cost

E : Excellent, P : Possible, F : Failure

It is promising to fabricate the high density MoO3 pellets
by plasma sintering method.

5- 42 -

42012/6/20
JAEA-Review 2012-030

Preliminary Fabrication Test of MoO3 Pellets
Fabrication by SPS Method

Sintering properties- High pressure- Short fabrication
time
Load
Pulse
current

Thermocouple

MoO3
powderGraphite-dieSintering density (% T.D.)

Graphite-punch

100
Target Values

95
90

20m

85
80

20m

75

: This work (21kNx10min): Patent value (10kNx10min)

70
500

550

600

650

Sintering temperature (C)

700

Ref. : Application No. 2009-018878, 30, JAN., 2009.

The high density MoO3 pellets canbe produced by SPS method.

Load
MoO3 Pellet : 20x10 mm

Evaluation of MoO3 pellets : Chemical properties, Solution property of 6M-NaOH, etc.

6

Extraction and Concentration of Tc-99m
It is necessary for utilization by the (n, γ)99Mo to develop the extraction andconcentration methods of 99mTc through Mo-Tc generators or Mo-Tc extraction
devices.
Subject of (n, )99Mo

The (n, γ)99Mo has several
advantages compared to
the fission Mo, but theextremely low specificactivity makes its uses
less convenient than thefission Mo.

R&D items- Separation from Mo solution- Extraction of 99mTc- Concentration of 99mTc

Investigation of Extraction Methods
MEK : Methyl Ethyl Ketone, PZC : Poly Zirconium Compound

MethodItems

Solvent Column Method
Extraction

PZC

Al2O3

Separation of Mo

P

E

P

Extraction of 99mTc

E

E

P

Concentration of 99mTc

E

F

F

Purity of Solution

P

P

E

Operation Time

E

P

P

Operation Control

E

P

P

E : Excellent, P : Possible, F : Failure

Solvent extraction
Column method: Master-Milker: Mo-Tc generator

7- 43 -

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2012/6/20
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R&D of Solvent Extraction

with BATAN and CTC

The solvent extraction is still used for small local supply of 99mTc at several
research reactor centers. Up to now, there has been no experience of a large
scale 99mTc extraction by this method.
Experimental Flow

Scheme of Cooperation

Separation and Extraction
Water phase : Mo solution or Saline
Organic phase : Ketone type (MEK)

Organic

Organic

JAEA

BATAN
International Cooperation

Cooperation

JPN Company
(CTC)
Water

Water

: Target material (99mTc)

Development and improvementof solvent extraction

Concentration
Development of concentration
method in extracted 99mTc

Between JAEA and JPN company
Development of method in out-pile tests
Between BATAN and JAEA
Demonstration tests for developed methods
The experiments were carried out 8 times in
BATAN and the useful data have been obtainedfor development of solvent extraction.

8

R&D as 99Mo-99mTc Generators with INP-NNC-RK
The 99Mo-99mTc generators loaded with (n, γ)99Mo adsorbed in zirconium
molybdate gel (Ze gel) have been used in Kazakhstan. Comparison with Zr gel
and PZC has been carried out between INP-NNC-RK and JAEA.
Experimental Flow

Construction of Database

Irradiation in WWR-K- 99Mo generation- Preparation of Mo(99Mo) solution
Extraction experiments[Zr-gel][PZC]

Preparation of Zr gel(Mo Sol. + HNO3 + ZrOCl2-8H2O)

Elution of 99mTc (99mTcO4-)
Evaluation Items- Properties(amount of Mo, purity, etc.)- Process time and control

Production method

Zr gel (NNC-RK)PZC (JAEA)<Irradiation material>- Chemical type

Natural Mo- MoO3, Metal Mo

Natural Mo- MoO3, Metal Mo<Irradiation>- Facility- Capsule volume- Irradiation time
- Adsorption of 99Mo

- Rabbit- About 30 cm3- 5-7 days
- Batch type

- Rabbit- About 30 cm3- 5-7 days
- Batch type

<Treatment as PIE>
- Treatment time

Dissolution in cell
~7 hDissolution in cell
~4h<Characteristics >- Radiation fraction- Activation by-product- 37-74 GBq/g- 92mNb- 37-74 GBq/g- 92mNb<Radioactive waste>- Waste solution- A little  nuclide- A little  nuclide

The experiments were carried out 4 times in
INP-NNC-RK and database between Zr gel and
PZC have been constructed.
9- 44 -

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2012/6/20
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Development of New Mo Generator

with Art-Kagaku Co.

It is necessary to develop new Mo generator in terms of effective use of resources
and reduction of radioactive wastes.
Mo Generator (PZC)

Development Target
Study of chemical construction of Mo generator
and development of new-type Mo generator.

Complex devices[Chemical construction of PZC and PTC]

Necessary to atmosphere control(Cl2 gas)

PZC

Inert gas
Difficulty of keeping

Cl Cl

Zr

Zr

O

(Case of PZC)

Mo generator

Mo adsorption by the replacement
between Cl and MoO4-

99Mo solution and
99mTc extruction

NaOH
Mo recovery by the replacement
between MoO4- and OH-

Used generator

Difficulty of the replacement
between OH- and MoO4-

Mo recovery

PTC

Cl Cl

Difficulty of PZC re-use

O

O

Cl Cl
Zr

O

O

Zr

Zr

Cl Cl

Cl Cl

O

OR

O

OR

OR

－Ti－O－Ti－O－Ti－O－Ti－O－
O

O

OR

O

[Mo adsorption site]
PZC : Zr-Cl site
PTC : Ti-organic compound site

Mo Adsorption：> 200mg/g
99mTc elution Property ：better than PZC
Re-used number：> 5 times
Fabrication cost
：1/10 of PZC
Demonstration test in JMTR

99mTc solution for Medicine

10with FRIThe specifications of 99mTc solution are determined in pharmacopoeia in eachcountry. It is important for domestic production to determine detail specificationsof 99mTc solution.
Items
99mTc content (%)

Radiochemical purity (%)
Radionuclidic impurity(μCi/mCi 99mTc)
99Mo
131I
103Ru
89Sr
90Sr
OtherspH
Chemical impurityAl
MEK

JPN Pharmacopeia(n, fiss) by USP(n, γ) by USP

90 ~ 11090 ~11090 ~110

≥ 95

≥ 95

≥ 95

≤ 0.15－－－－－
4.5 ~ 7.0
≤ 10 ppm－≤ 0.15≤ 0.15≤ 0.05－－≤ 0.05≤ 0.006－≤ 0.00006－
& emitters≤0.1≤ 1x10-5
 Emitter≤1x10-6
4.5 ~ 7.5
≤ 10 μg/ml－≤ 10 μg/ml≤ 0.1%

USP : United States Pharmacopeia

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Investigation of 98Mo Enriched MoO3
It is necessary to fabricate the high density MoO3 pellets with 98Mo enriched MoO3
for increasing of 99Mo production. However, it is difficult to obtain t 98Mo enriched
MoO3 powder.
Evaluation of 99Mo Activity

Activity of 99Mo (Bq/g)

2.00E+11

99Mo Enriched Powder

Irradiation datein JMTR (7 days)

Now, the 98Mo enriched MoO3 powder
is produced only in Russia.
Expensive
Unstable supply

98%

1.60E+1170%
1.20E+1150%
8.00E+10

Use of 98Mo enriched MoO3 powder
in an emergency such as stop oftransportation, etc.

Natural
4.00E+10[R&D for 98Mo enriched powder](1) 98Mo recycle system(see next viewgraph)

0.00E+00
0

2

4

6

8

10

Irradiation Time (day)The amount of 99Mo generation increases
greatly if the enriched powder is used.

(2) Fabrication of 98Mo enriched powder
12

Development of Mo Recycling with Taiyo Koko Co.
MoO3 is used as the irradiation target for the production by (n, )99Mo. Recycling
technology development is proposed to recover molybdenum (Mo) for the effective
use of resources and reduction of radioactive wastes.
Subjects for Mo Recycling(a) Recovery of Mo resource
Mo recovery from the Mo
solution and used 99Mo-99mTc
generators (PZC, etc.) andutilization of recovered Mo.
(b) Reduction of Wastes
Treatment
and
waste
reduction of used 99Mo-99mTc
generators (PZC, etc.) andMo solution(c) Recovery of PZC
Development of reusable
99Mo-99mTc generators

Recycling System with PZC (for example)
99mTc extraction

Recovery of
PZC

(c)

PZC

99Mo production(Neutron irradiation)

99Mo recovery(a)
Recovery of Mo

used
PZC
Waste

(b)

MoO3 Powder

Reduction of Wastes

Reduction ofradioactive wastes(only ZrO2)

13- 46 -

8

2012/6/20
JAEA-Review 2012-030

Proposal of Mo Recycling Methods

with Taiyo Koko Co.

Two kinds of recycling methods are proposed to recover molybdenum (Mo)from the used Mo generator.
Elute method

Sublimate method(1) The adsorbed Mo is eluted from the used
Mo-adsorbent.
(2) After the elute, Mo is recovered as MoO3.
(3) Reused of Mo-adsorbent is also possible.

(1) Mo is recovered from the used Moadsorbent by the sublimate of MoO3.
(2) The used adsorbent is reduced.
(Radioactive waste is reducible)
100

PZC
：5.0g
Dipping velocity ：1.2mℓ/min
Dipping temp. ：RT30

Mo recovery rate（%）

Mo concentration（g/ℓ）

40： 1N-NaOH： 0.5N-NaOH： 0.1N-NaOH

20

10

Test condition：950℃×2h

9080
Heater70

Air

Electric furnace

0

0

20

40

60

80

100

Dipping（mℓ）

600cooling

Used PZC
MoO3

0.1
0.2
0.3
Content of Na in Used-PZC（%）

0.4

14

3. Constructions ofIrradiation and PIE Facilities- 47 -

9

2012/6/20
JAEA-Review 2012-030

Flow Chart of 99Mo Production in JMTR
The 99Mo production process has been developing in JMTR. The performance tests
are also carried out in each part. Now, the irradiation and PIE facilities are prepared.

Loading toContainer

Dismantle

EnclosureTransfer

PIE Facility
Maintenance of hot cells

Irradiation Facility
Hydraulic rabbit Irradiation Facility

15

JMTR Core and 99Mo Production Facility
New hydraulic rabbit irradiation facility (irradiation hole : M-9) will be prepared in2012. The structure of new hydraulic rabbit is revised.
Irradiation Holes

Structure of hydraulic rabbit tube

Existent 99Mo production facility
New 99Moproduction facility

5 rabbits

36.6 TBq/week(~ 1000 Ci/week)

Fuel element

Be frame

Control rod with
fuel follower
Be reflector

Al reflector
Gamma ray shield
plate

16

- 48 -

10

2012/6/20
JAEA-Review 2012-030

4. Conclusion

Conclusions
99Mo

production by (n, ) method has been determined fromviewpoints of nuclear nonproliferation, reduction of radioactive
wastes and utilization of Mo resource in JMTR.
The sintered MoO3 pellet with high density will be used as
irradiated target material for 99Mo production and fabrication
development of high density MoO3 pellet is carried out.
Development of 99Mo-99mTc extraction and concentration
methods is carried out and useful data are accumulated under
the international cooperation.
New hydraulic rabbit irradiation facility is constructed as
irradiation facility of 99Mo production. JMTR will be able toprovide about 20% (about 37 TBq/week (1,000Ci/week)) of theamount of 99Mo imported from overseas.

17- 49 -

11

2012/6/20
JAEA-Review 2012-030

R&D Schedule on 99Mo Production in Japan
2010 2011 2012 2013 2014 2015
Schedule

Planning

R&D

Production

Fabrication Development of
Irradiation Targets
Development of Separation,
Extraction and Concentrationof 99Mo/99mTc*1

Examination of 99mTc solution
for radiopharmaceutical
Mo Recycling
Sharing of Irradiation Facility
for 99Mo production

Preparation of Facilities
Consideration for 99Mo Production, etc.

*1 : Approval for the Pharmaceutical Affairs Law

18- 50 -

12

JAEA-Review 2012-030

R&D in Japan on extraction andconcentration of 99mTc-A Preliminary study using Re in stead of 99mTcM. TANASE *1, T. SHIINA*1, S. FUJISAKI *1, A. OHTA*1,
Y.KAWAUCHI*1, A. KIMURA*2, K. NISHIKATA*2,
M. YONEKAWA*2, Y. KATO *2, K. TSUCHIYA *2,
Y. MORIKAWA*3, M. KOBAYASHI *3, A. YAMAMOTO *3*1 Chiyoda Technol Corporation*2 Japan Atomic Energy Agency*3 FUJIFILM RI Pharma Co., Ltd.

Contents
1. Introduction
2. In the last study (1)3. In the last study (2)4. Purposes in the present study
5. Flow of Experimental Procedures
6. Procedures (1)7. Procedures (2)8. Goals of the tests
9. Inspection of product10. Result 1 Recovery of Re
11. Result 2-1 Impurity of Mo in Product12. Result 2-2 Impurity of MEK in Product13. Conclusion- 51 -

1

JAEA-Review 2012-030

Introduction•

99mTc is used as a radiopharmaceutical in the medical field

for the diagnosis, and manufactured from 99Mo, the parent
nuclide.
• Chiyoda Technol Corporation (CTC), Japan Atomic Energy
Agency (JAEA) and FUJIFILM RI Pharma Co., Ltd. (FRI)
have developed a (n, γ)99Mo method to produce 99Mo, as apart of the industrial use expansion after JMTR will restart.
• However, the radioactive concentration of 99mTc in thesolution by the (n, γ) 99Mo method is greatly lower thanthat by the (n, f) 99Mo method.
To overcome the difficulty, the solvent extraction of 99mTc
produced from 99Mo were done and followed by an alumina
column method.
1

In the last study (1)Dissolution of MoO3
Extraction of
Re with MEK
Purification with
Basic-alumina column

A method shown in left side of this slide
was proposed and, by using Re* instead of
99mTc, Re recovery tests were tried in order
to ascertain that the method is applicableto the commercial production of 99mTc
created from (n,γ)99Mo (ISMTR-4, Oarai
Japan (Dec.6-9, 2011). (* Re and Tc arehomologous element）
The purposes of the tests were as follows:

Adsorption to acidicalumina column
Elution of Rewith Saline(i)Re recovery yield is more than 80%.
(ii)TheReconcentration
factor,concentration of Re in saline to that of Re
in Re-Mo solution, is more than 15.

2- 52 -

2

JAEA-Review 2012-030

In the last study (2)We got the following results:
・ To dissolve MoO3 , 6M NaOH solution was chosen.
・ Re was recovered with a yield of more than 90%, andcleared our goal of 80%.
・ By column method using basic and acid alumina
columns, Re was recovered successfully with theconcentration factor of 33, and cleared our goal of 15.
These results show that it is possible toobtain the high concentration of 99mTc by thesolvent extraction with MEK, followed byalumina column method.
3

Purposes in the present study
In the present study, the process fromextraction of Re to elution of Re wasrepeated in order to simulate milking using
the same Mo-Re solution, and(i) Effect of milking to recovery yields was
examined. And also,(ii) Impurities of Mo and MEK in the productwere determined and compared with thetentative limits of regulations in Japan,
respectively.
4- 53 -

3

JAEA-Review 2012-030

Flow of Experimental Procedures
Extraction of Re

Dissolutionof MoO3

Milking (Extraction) process:The aqueous phase of Mo-solution were subjected to theextraction process from “Addition of Re to Mo-solution” to“Inspection of product” five times.

Addition of Re
to Mo-solution

Extraction of
Re with MEK

Aqueousphase

Separation ofOrganic andAqueous phases

Organic
phase

Purification and concentration of Re
Organic
phase

Purification with
Basic-alumina column

Adsorption toacidic-alumina column

Elution of Rewith Saline

Inspection of
product

5

Procedures(1)Extraction of Re

Mo/Re solution

Milking was repeated five times: 1st day, 2ndday, 3rd day, 8th day and 9th day.
Extraction of Re with MEK

After separation,
aqueous phase is
kept at 30°C inwater bath.

6M-NaOH :400ml
MoO3:200g
Re : 450μg(99mTc estimated : 357Ci)

MEK : 90ml
Mixing time: 5minStanding time : 30min

：Re
：Mo

MEK after separation

Purification
andconcentration of Re

Next step6- 54 -

4

JAEA-Review 2012-030

Procedures(2)Purification and Concentration of Re
MEKContainingRe

1st column
Basic Alumina:7g 1st column is used
12 mm I.D.
for removing Mo
2nd column
Acid Alumina:7g
12 mm I.D.

and H2O from MEK,
and 2nd column for
adsorbing Re.

Basic
alumina

Acidicalumina
Product ofRe

MEKDistilled
water

30ml

2nd column
Acid Alumina
12 mm I.D.

In order to rinse
MEK remaining
in 2nd column,
distilledwater
was passed.

6ml (discarded)+
16ml (Re was collected)

Saline

2nd column
Acid Alumina
12 mm I.D.

Distilled water

7

Goals of the tests
In order to make sure adaptable to the practicalproduction, effect of milking to recovery yields, Re
concentration factors, and impurities of product were
examined.
The goals of the tests were as follows:(i) Re recovery yields are more than 80%.
(ii)
The
Re
concentration
factor,concentration of Re in saline to that of Re in
Re-Mo solution, is more than 15.
(iii) The impurities of 99Mo and MEK in
product are below the tentative limits ofregulations in Japan.
9- 55 -

5

JAEA-Review 2012-030

Inspection of product
Impurities of Mo and MEK in the product were
determined by ICP-MS and HPLC. From the impurity ofMo, the radioactivity of 99Mo was calculated under theconditions of practical production and the radioactive
impurity percent of 99Mo/99mTc was estimated.
And the values determined were compared withtentative ones regulated in radiopharmaceuticals andshown below.
Quality levels
Radioactive impurity
99Mo/99mTc

Ratio(estimated) ≦0.015%

Chemical impurity

Permissible amount of MEK：5000ppm

9

Result 1 Recovery of Re
Experimental result on recovery yields is shown.
Recovery yields

Re recovery yield (%)

Effect of milking to recovery yields was examined.
100

After the first 6 ml of saline
was discarded, Re was collectedin the next 16 ml of saline
containing as a product at aflow rate of about 10 ml/min.

80
60
40
20
0
1st
(1 Day)2nd(2 Day)3rd(3 Day)Milking

4th（ 8 Day）

5th（ 9 Day）

The average recovery yield was very high to be
98.1%, and the concentration factor 29.
10- 56 -

6

JAEA-Review 2012-030

Results 2-1 Impurity of Mo in productThe impurity of Mo is 0.021 ppm(20 ml) and the percent ratio of Mo
in the product to Mo in the original
Re-Mo solution is 3.2x10-7%. From
this value, the impurity of
99Mo/99mTc was estimated below.

0.07：Amount of 99Mo in the product

0.06
0.05
0.04
0.03
0.02
0.01
0
1st
(1 Day)99Mo/99mTc

2nd(2 Day)3rd(3 Day)Milking

4th（ 8 Day）

5th（ 9 Day）

in 99mTc product And ratio

9.E-07

6.E-03：impurity of 99Mo in the product(left scale)(%)

7.E-07

99Mo/99mTc

8.E-07

5.E-07：ratio to 0.015%(right scale)

6.E-07

5.E-03
4.E-03
3.E-03

4.E-07
3.E-07

2.E-03

2.E-07

1.E-03

1.E-070.E+00The impurity of 99Mo/99mTc
was calculated to be
4.6x10-7% which is muchless than 0.015% and theratio to the regulation limit
is about 3x10-3.

Ratio to 0.015%

Amount of 99Mo (ppm)

Amount of Mo in 99mTc product

0.E+00
1st(1 Day)2nd(2 Day)3rd(3 Day)Milking

4th（ 8 Day）

5th（ 9 Day）

11

Results 2-2 Impurity of MEK in productAmount of MEK (ppm)

1000

20
：amount of MEK (ppm) (left scale)：percent ratio to 5000 ppm (right scale)

900
800

18
16

700

14

600

12

500

10

400

8

300

6

200

4

100

2

0

0
1st
(1 Day)2nd(2 Day)3rd(3 Day)Milking

4th（ 8 Day）

5th（ 9 Day）

Percent ratio to 5000 ppm

Amount of MEKAs shown in the figure,the amount of MEK in
the product is increasing
with elapsed day.

Permissibleamount of MEK：
5000ppm

Average impurity of MEK was 572 ppm and about 1 order lower
than 5000 ppm of regulation in Japan.
12- 57 -

7

JAEA-Review 2012-030

Conclusion
The results are concluded below;(i) Recovery yield・The average recovery yield was 98.1%, and the concentration factor 29,
which is beyond our goal of 15.
・ The concentration of 99mTc was estimated to be 728 GBq (19.7 Ci) /ml
when 200 g of MoO3 to be irradiated in JMTR and 13.2 TBq(357 Ci) of
99mTc are handled.
(ii) Impurities of Mo and MEK・Impurity of Mo is 0.021 ppm and the percent ratio of 99Mo/99mTc
in the product was calculated to be 4.6x10-7%, when 200 g of MoO3 and13.2 TBq(357 Ci) of 99mTc are handled. The value of 4.6x10-7%
was 3 orders lower than that of tentative regulation limit in Japan.
・ The average impurity of MEK was 572 ppm, and about 1 order lower
than that of tentative regulation limit in Japan.
Following these results, we have been developing the production
method using 99Mo / 99mTc with BATAN in Indonesia.
13

Thank you for your attention
ありがとうございました。

- 58 -

8

JAEA-Review 2012-030

Specialist Meeting on Mo-99 Production by (n,γ) Method

Technical request/standard for 99mTc
solution for medical use in Japan

March 9th, 2012
FUJIFILM RI Pharma Co.,Ltd.
1

Radiopharmaceuticals supplied for Sale
Unit : million yen
70,00060,00050,00040,000in VIVOin VITRO
Total

30,00020,00010,00002000 2001 2002 2003 2004 2005 2006 2007 2008 2009 2010
Statistics on the Distribution of Radioisotopes 20102

- 59 -

1

JAEA-Review 2012-030

Number of Hospitals and/or Clinical Laboratories(ｗhere radiopharmaceuticals were supplied for clinical purpose)

1400
1200
1000
800
600
400
200
0

2000

2001

2002

2003

2004

2005

2006

2007

2008

2009

2010

Total

1273

1262

1265

1268

1266

1280

1301

1297

1293

1278

1281

In Vivo

1225

1221

1229

1237

1239

1255

1279

1275

1274

1263

1267

In Vitro

182

160

133

110

90

74

61

53

45

34

31

Statistics on the Distribution of Radioisotopes 2010

3

Amount of Radiopharmaceuticals (for in vivo use) Supplied
99Mo-99mTc Generator＆

99mTc Products

450,000

Unit：GBq

400,000
350,000
Ｍｏ-Ｔｃ（Ｇ）
Ｔｃ-99ｍ
Ｘｅ-133
Ｇａ-67
Ｔｌ-201
Ｉ-131
Ｉ-123
Ｆ-18

300,000
250,000
200,000
150,000
100,000
50,000
0
2006

2007

2008

2009

2010

Statistics on the Distribution of Radioisotopes 20104

- 60 -

2

JAEA-Review 2012-030

Standard Specification of 99Mo solution
This is Standard Specification of our company which was decided on the basisof that there is no difference from (n,f) Mo-99 and (n,γ) Mo-99.

Ｉｔｅｍs

Specification

Chemical Form

Sodium Molybdate

Mo-99 content (％）

90～100Description

Clear and ColorlesspH

9.0～12.0

Identification

Photopeak : 0.14MeV(Tc-99m)(Gamma –Ray Measurement)

0.739MeV(Mo-99)

Radiochemical impurity (％）≦105Standard Specification of 99mTc solution
Comparison for Sodium Pertechnetate solution with JP,USP and EP
Items

JP

USP

EP

〔from （n,γ）〕〔from （n,γ）〕

Tc-99m content (％）

90～11090～11090～110Description

Clear and Colorless

Clear and Colorless

Clear and ColorlessｐH

4.5～7.0

4.5～7.5

4.0～8.0

Radionuclidic
impurity

Mo-99≦0.015％

Mo-99≦0.015％

Mo-99≦0.1％

Others≦0.01％

Others≦0.01％

Radiochemical

≧95

≧95

≧95

Aｌ≦10μｇ/mL

Al ≦5ppm

Purity(％）
Ｃｈｅｍｉｃａｌ impurity

Al ≦10ppm

MEK ≦:0.1％
JP standard for Sodium Pertechnetate is estimated by alumina column method without MEK. But as weare now investigating MEK extraction method, we need new standard of MEK for chemical impurity.
6- 61 -

3

JAEA-Review 2012-030

Main Operation and Delivery
Chemical Separation
facility

ReactorFacility

Main
Operation

99Mo
Production
By JMTR

Transportation
Container

Radiopharmaceuticalfacility

99Mo/99mTc

Hospital

99mTc-Compounds

Separation
Extraction
Concentration

production

Mo-99

Tc-99mType B(U)Type A

Tc-99mType A

7

Weekly Schedule for 99mTc compounds
We are now importing (n,f) Mo-99 from foreign countries every week. But in case of (n,γ)Mo-99,
we will need to get Tc-99m solution from Sunday to Friday morning.
In case of Mo-99 acquisition

Fri

Tue

Wed

Thu

Fri

○

○

○

○

○

○

○

○

○

○

○

Sun

Mon

Tue

Wed

Thu

Fri

Sat

Tc-99m solution arrival

●

●

●

●

●

●

Production/shipment

○

○

○

○

○

○

○

○

○

○

○

Mo-99 arrival

Sat

Sun

◎

Tc-99m generator

Mon

Sat

（◎）
●

Production/shipment●○

Delivery to hospitals

In case of Tc-99m acquisition

Fri

Delivery to hospitals

Sat

○
8

- 62 -

4

JAEA-Review 2012-030

Radioactive Conc. for 99mTc solution at production time
We deliver Tc-99m compounds every weekday to hospitals. We needto produce all of our products one day before.
Pharmaceutical Company

Hospital

Production day

Next dayElution/Acquisition

Production

9:00

9：00-15：00

ArrivalCalibration time

8：00-9：00*12:00:00

Due to labeling to the compound, Tc-99m solution
is diluted by 2-5 times to adjust bulk solution.

≧1Ci/mL

our
Requestproduction

225 ～450mCi/mL

deliveryApproval Standard
Main Compounds
10,20mCi/mL

Administration

Decay
Tc-99m after 27 hours =0.0444
9

Production flow for Techne MDP Injection
Sodium Pertechnetate Injection
Adjustment for Radioactive concentration
Mixing with MDP solution

Reaction
Filtration

Dispense /Closing

Pre-filled Syringe

Insert into W and Lead shield
Techne MDP Injection
10- 63 -

5

JAEA-Review 2012-030

Techne MDP injection (pre-filled syringe)

Diagnosis for Bone & Malignant tumor
11

Requirement of the quality of 99mTc solution
Guideline for Sterilized medicine obtained by aseptic process（MHLW issued this notification on April 20th,2011 )The raw materials that will put into critical area must fulfill one ofitems followed.
・the sterilized materials
・the materials would be performed adequate sterilization,
as to characteristic and bio-burden level (according toperiodical validation in advance, the bio- burden should be
confirmed to be within standard)
The raw materials should be controlled below the quantity ofendotoxin permitted.
・If it does not remove the pyrogen, it must be guaranteed
lower the standard of endotoxin.
・If there is a process to remove the pyrogen, you should
apply proper de-pyrogen method.
12- 64 -

6

JAEA-Review 2012-030

Application process for Pharmaceutical affairs

・To establish the Solvent Extraction method
for Sodium Pertechnetate solution↓
・To collect physics and chemistry (include non-toxic test) data↓・Application of Solvent Extraction method for Sodium
Pertechnetate Generator to MHLW
・To apply the revision of JP(The Minimum Requirements for Radiopharmaceuticals)↓
・Approval by MHLW and Revision of JP↓
・Start of Production
13

Conclusions
 The sodium pertechnetate solution is needed
every morning until around 9:00 AM from
Sunday to Friday.
 The concentration of sodium pertechnetate
solution that is supplied is over 1Ci/mL
at around 9:00 AM
 The quality of Tc-99m solution may be guaranteedby the management of bio-burden and endotoxin
quantity.
 We need application of the sodium pertechnetate
generator to MHLW and revision of JP
14- 65 -

7

This is a blank page.

JAEA-Review
2012-030
JAEA-Review 2012-030

A.4 Presentation documents for status and plan in each countries- 67 -

This is a blank page.

JAEA-Review 2012-030

National Center for Nuclear
Research
RADIOISOTOPE AND
Mo-99 PRODUCTION
in Research Reactor

MARIA

Experience & Plans
for Future
Janusz Jaroszewicz
National Centre for Nuclear Research(former: Institute of Atomic Energy POLATOM)j.jaroszewicz@ncbj.gov.pl

March 2012

THE MAIN NUCLEAR FACILITIESSWIERK - OTWOCK

Specialist Meeting
Tokyo,09 Mar 2012

POLATOM
Research Reactor

MARIA

Material ResearchLaboratory

General view

- 69 -

1

JAEA-Review 2012-030

National Center for Nuclear
Research

Specialist Meeting
Tokyo,09 Mar 2012 The first Polish institute in the field of nuclear energy was
founded in 1955 as the Institute of Nuclear Research; The Institute of Atomic Energy (IAE) was emerged in 1983
as one of three independent research institutes formedon the basis of the Institute of Nuclear Research; On the 1st January, 2007 the Research Developing Center
for Radioisotopes POLATOM joined the IAE structure at
the right of Research Enterprise; On the September 1st, 2011 the IEA POLATOM & Institutefor Nuclear Studies have been merged as a National
Center for Nuclear Research (NCBJ).

3

THE MAIN NUCLEAR FACILITIES (NCBJ)
Specialist Meeting
Tokyo,09 Mar 2012

Radioisotope Centre POLATOM

Material
Research
Laboratory(MRL)

MARIA Research Reactor- 70 -

2

JAEA-Review 2012-030

Specialist Meeting
Tokyo,09 Mar 2012MARIA

RESEARCH REACTOR

MAIN NUCLEAR FACILITYMARIA RESEARCH REACTOR
Otwock - Świerk

General view

 Nominal power: 30 MW

Operated since

December 16, 1974

 Type: pool type reactor

 Thermal flux: 2.5*1014 n/cm2s

Upgraded 1992

- 71 -

3

JAEA-Review 2012-030

MARIA RESEARCH REACTOR
Facility description

Specialist Meeting
Tokyo,09 Mar 2012 The high flux research reactor MARIA is a water andberyllium moderated reactor of 30 MW power level; Pool type reactor with pressurized fuel channels
containing concentric tube assemblies of fuel elements;
 Fuel channels are situated in matrix containing
beryllium blocks surrounded by graphite reflector:• nominal power

30 MW(th)• thermal neutron flux density

2.5*1014 n/cm2s
H2O, beryllium
graphite in Al
channel type2025-30• moderator• reflector• cooling system• expected operation

MARIA RESEARCH REACTOR
Facility opportunities

Specialist Meeting
Tokyo,09 Mar 2012

 Expected operation time of reactor: 2025-30
 Experienced reactor operation staff;
 Guaranteed delivery of fuel for the next years:- Russian fabricated high enrichment fuel (36%)
will assure operation for the next 3 years;- MARIA has two potential suppliers of LEU fuel:(a) AREVA (CERCA), France (tested)(b) TVEL, Russian Federation (under design);

 MARIA RR fuel conversion program HEU to LEU
according to RERTR Initiative is in progress
19.75% / 485 g U-235 per FE / density of 4.8 g/cm
(2 silicide type FE made by CERCA were tested since August 2009).

- 72 -

4

JAEA-Review 2012-030

MARIA RESEARCH REACTOR
Vertical cross-section

Specialist Meeting
Tokyo,09 Mar 2012

HOT CELLSPENT FUELS INTECHNOLOGYCHANNELSCONTROL RODSDRIVE MECHANISMFLOWGATETECHNOLOGY POOL+.6.33
R 3250+4.1

12250

SPENT FUEL

+2.25

HORIZONTALCHANNELR 2250
REACTOR

+1.1

-2.85
-2.85

MARIA

REACTOR POOL

Specialist Meeting
Tokyo,09 Mar 2012

Configuration of reactor core
22 FE / P=22MW- 73 -

5

JAEA-Review 2012-030

MARIA RESEARCH REACTOR

Specialist Meeting
Tokyo,09 Mar 2012

operation
4259 h3912 h4279 hMARIA RESEARCH REACTOR
Operation schedule for 2011

Specialist Meeting
Tokyo,09 Mar 2012

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6

JAEA-Review 2012-030

MARIA RESEARCH REACTOR
Main reactor applications

RI Production

Fuel Management

Education & Training

Neutron Beam Applications

Specialist Meeting
Tokyo,09 Mar 2012

MARIA RESEARCH REACTOR
Existing & Potential Applications

Specialist Meeting
Tokyo,09 Mar 2012

 Irradiation of target materials for production of radioisotopes
for medicine and industry purposes:

 Radioisotope production: 600 TBq (annually in 2011 year), Irradiation HEU targets for Mo-99 production: 6 000 TBq (2011y)
 Laboratory for High Activity Radionuclides: irradiation of LEU
targets for production radionuclides based on FPs (perspective); Neutron modification of crystals (NTD) and minerals;
 Research in neutron physics with utilization of neutron
beams on horizontal channels including BNCT facility; Testing of fuel and structural materials for nuclear powerengineering focused on HTR program (in the plans); Education and Training on atomic energy, nuclear physics
and nuclear chemistry.

- 75 -

7

JAEA-Review 2012-030

Specialist Meeting
Tokyo,09 Mar 2012MARIA

RESEARCH REACTORIRRADIATION SERVICESMARIA RESEARCH REACTOR

Existing & Potential Applications

Reactor Facility for

Neutron Transmutation Dopingof Silicon Single Crystals (NTD)

Specialist Meeting
Tokyo,09 Mar 2012

Radioisotope
production

Target irradiation in
MARIA Research Reactor& radiochemical
reprocessing in POLATOM


28  0.2
 0.16
 37 

100
85

78

100

78

100

78

100

145

The project presented& awardedon the Editionof the Polish Productof the Futurer5

Typ 37mXe

 0.1
 35 

 0.1
 25 

 0.1
 20 

 0.1
 15 

Typ 35m

Typ 25m

Typ 20m

Typ 15m- 76 -

8

JAEA-Review 2012-030

MARIA RESEARCH REACTOR
In-core irradiation facilities

Specialist Meeting
Tokyo,09 Mar 2012

Irradiation channels
Channels with hydraulic transport system
NTD irradiation facility
Irradiation position for minerals modification

Reactor MARIA pool

Configuration of reactor core

MARIA RESEARCH REACTORRadioisotope Production

Specialist Meeting
Tokyo,09 Mar 2012

Neutroncharacteristicof theirradiation
channels*) – activation
measurements
recalculated for
P=18 MW- 77 -

9

JAEA-Review 2012-030

MARIA RESEARCH REACTORRadioisotope Irradiation Facilities

Specialist Meeting
Tokyo,09 Mar 2012

Radioisotope production

Radioisotope Processing Facility & Infrastructure

MARIA RESEARCH REACTORRadioisotope Irradiation Facilities

Specialist Meeting
Tokyo,09 Mar 2012

Irradiation standard cans

- 78 -

10

JAEA-Review 2012-030

MARIA RESEARCH REACTORRadioisotope Irradiation Facilities

Irradiation standard cans

(a)

1.
2.
3.
4.

Specialist Meeting
Tokyo,09 Mar 2012(b)

1. Upper Al layer
2. Graphite holder
with iridium pellets
3. Graphite separator
4. Lower Al layer

Tighten belt
Aluminum holder
Iridium pellets
Target separators

Special irradiation cans: with aluminum holder (a) & graphite holder (b)MARIA RESEARCH REACTORRadioisotope Irradiation Facilities

1.
2.
3.
4.
5.

Specialist Meeting
Tokyo,09 Mar 20121.

Lower station in hot cell
Drainer
Transfer station in reactor pool
ForcerIrradiation channels

Irradiation channels with hydraulic transfer system- 79 -

11

JAEA-Review 2012-030

MARIA RESEARCH REACTORRadioisotope Irradiation Facilities

Specialist Meeting
Tokyo,09 Mar 2012

Reactor pool

Hydrauliccatcher

Irradiation
channels
CameraMounting slab

Reloading Systemof irradiated cans

TransferstationReactor on operation
regimeCore matrix

MARIA RESEARCH REACTORRadioisotope Irradiation Facilities

Specialist Meeting
Tokyo,09 Mar 2012

Hot cell for radioisotope
production
1. Loading hot cell
2. Expedition hot cell
3. Transport car
4. Storage channels- 80 -

12

JAEA-Review 2012-030

MARIA RESEARCH REACTOR
Facility capabilities

Specialist Meeting
Tokyo,09 Mar 2012

Radioisotope production (main activity):• Major target material are irradiated: TeO2, Lu2O3,
LuCl3, Yb2O3, S, , KCl, Cu, Se, SmCl3, Co, NiO• Vertical channel and hydraulic transfer system wereconsidered for high activity radioisotopes;• Ir-192 seeds for Intravascular Therapy and low

activity Ir-192 source ribbon for Oncology Application
have been irradiated;• Production of high activity Ir-192 sources: specificactivity of iridium discs ca 550Ci/g.

MARIA RESEARCH REACTORRadioisotope Production

Specialist Meeting
Tokyo,09 Mar 2012

- 81 -

13

JAEA-Review 2012-030

MARIA RESEARCH REACTORRadioisotope Production

Specialist Meeting
Tokyo,09 Mar 2012

Specification of irradiated targets & produced activity in 2011

MARIA RESEARCH REACTORRadiopharmaceuticals Production

Specialist Meeting
Tokyo,09 Mar 2012

Specification of POLATOM production• 99mTc generator• Cold kits for 99mTc
labeling• 131I in gel capsules• 131I for injection• MIBG – 131I
• MIBG – 123I
• Hipurate – 131I
• SrCl2-89Sr• Sodium ortho –Phosphate 32P

POLATOM
By courtesy of OR POLATOM- 82 -

14

JAEA-Review 2012-030

MARIA RESEARCH REACTORRadiopharmaceuticals Production•EDTA- 51Cr for injection•Sodium Chromate-51Cr•CuCl2-64Cu•Iron (III) 59Fe citrate•90Y as yttrium chloride•177Lu as lutetium chloride•188W/188Re generator•Kits for labelling with 99mTc

Specialist Meeting
Tokyo,09 Mar 2012

POLATOM

By courtesy of OR POLATOM

Specialist Meeting
Tokyo,09 Mar 2012MARIA

RESEARCH REACTORMOLYBDENUM PRODUCTION- 83 -

15

JAEA-Review 2012-030

BACKGROUND INFORMATION:
99Mo Production

Specialist Meeting
Tokyo,09 Mar 2012

• European production was based up to 2010 on HEU target
irradiation in three reactors: HFR (Petten, NL), BR2 (Mol, B)
and OSIRIS (Sacley, FR);
• Reactors were operated as a supply network within Europe
in conjunction with two Mo-99 processing facilities operated
by COVIDIEN (Petten, NL) and IRE (Fleurus, B);• Supply elsewhere in the world is not-worked and exist as anumber of single supply lines of geographically isolated
reactor processing facilities (Canada, South Africa andAustralia);
 single break in the supply chain completely stops Mo-99/Tc-99m supply• With the ageing of the European supply network it isnecessary to consider new reliable supply sources:

31

Reactors and Mo-99 Processing
Facilities around the world
MARIA RR in the network
Nordion

NRU

BR2

HFR

Covidien

IRE

OsirisNTPMARIARR in operation

SafariRRs temporary out of service
Mo-99 Processing Facilities• BR2, Belgian Nuclear Radiopharmacy Centre, Belgium• OSIRIS, France, French Atomic Energy Commission (CEA)• NRU, National Research Universal Reactor, Canada• HFR, Nuclear Research & Consultancy Group (NRG), The Netherlands• SAFARI-1, South African Nuclear Energy Corporation (NESCA), S.Africa• COVIDIEN, USA / The Netherlands• MDS Nordion, Canada• Nuclear Technology Products (NTP), S. Africa• The Institute for Radio Elements (IRE), Belgium- 84 -

16

JAEA-Review 2012-030

BACKGROUND INFORMATION:
99Mo Production

Specialist Meeting
Tokyo,09 Mar 2012

• Processing facilities should be located close to the reactors:
 the primary reasons for this is minimize the transport of high activity
fissile nuclear materials;
 reduce as much as possible the decay loss of Mo-99 (the quantityof Mo-99 produced decay by 1% per hour from the EOB);
 the processing site should be licensed for waste management.

• The main European Mo-99 producers: COVIDIEN and IRE
plants are located at reactor site or on reasonable distance:
 COVIDIEN plant is located at the HFR site; PALLAS location will be same
 IRE plant is located close to BR-2; MYRRAHA location will be same, Plans for new plant located in Munich (TUM site next to FRM II)
and / or Cadarache (CEA site next to JHR).
33

MARIA RESEARCH REACTOR
Molybdenum Program

Specialist Meeting
Tokyo,09 Mar 2012

Number of medical procedure around the world basedon 99Mo/ 99mTc generators (2009 y)

Source: Prelminary report on supply of radioisotopes for medical use andcurrent developments in nuclear medicine, EC 200934

- 85 -

17

JAEA-Review 2012-030

MARIA RESEARCH REACTOR
Molybdenum Program IAE POLATOM and now NCBJ is willing to provide
solutionensuring
sustainableproduction
anddistribution Mo-99 in both mid and long terms based onirradiation HEU & LEU targets in MARIA RR using the
new LHAR Laboratory and a new Processing Facility
 Molybdenum Program in IAE POLATOM includes:
1)
2)
3)
4)
5)

Technology of irradiation targets in MRR: HEU-based
Laboratory for High Activity Radionuclides: LEU-based & (n,γ)
Production Processing Facility located on NCBJ site: LEU-based
Collaboration with global Mo-99 producers & RRs Coalitions
Participation on existing programs as CRP on “Developing

techniques for small scale indigenous Molybdenum-99 production
using LEU fission or neutron activation”

First Step of Molybdenum Program
Technology of irradiation uranium targets(Collaboration with COVIDIEN, Petten, NL)

Safety analyses& Design

Safety Commission
Approval

National Regulatory Body
Approval

Specialist Meeting
Tokyo,09 Mar 2012

Domestic &Foreign Licenses

Schedule of Molly Program in IAE POLATOM
Tests and reactor out-core experiments: Dec. 2009
Test with dummy plates: Jan. 2010
Test irradiation uranium plates: Feb. 2010
Irradiation facility
manufacture 2009/10

Commercial irradiation (HEU): Mar 2010

- 86 -

18

JAEA-Review 2012-030

Radioisotope production: 99Mo
Specialist Meeting
Tokyo,09 Mar 2012

Irradiation in MRR Rig(Reactor MARIA)

Uranium target assembling(CERCA, France)

Reloading in Hot Cell
(Reactor MARIA)

Dissolution process
99Mo / 99mTc

Mo-99 recovery

Generators

Purification process

FDA & Health
CanadaApprovals

Uranium plates reprocessing(COVIDIEN, The Netherlands)(March 2010)

MARIA RESEARCH REACTOR
Reactor pool

Specialist Meeting
Tokyo,09 Mar 2012

Configuration of reactor core
24 FE / P=22MW- 87 -

19

JAEA-Review 2012-030

Molybdenum Program
3

6

18

+24.705+20.30+19.30+13.707

17
11+10.10

Irradiation& transport targets
to Hot Cell

4
9+15.50

+7.60

8

+3.60
+1.10
-0.25
-1.70

-2.00

-1.70
-3.20

10

HOT CELLSPENT FUELS INTECHNOLOGICALCHANNELSCONTROL RODSDRIVE MECHANISMSLUICE

3. basen przechowawczy
4. komory izotopowe
5. komora demontażowa
6. komora D2O
7. hala reaktora
8. hala eksperymentów

9. dukt instalacyjny

10. dukt kablowy operation
Reloading
11. śluza czysta
17. kondensator
wodny
SBM-EJin HotCell
18. suwnica o udźwigu 16/3.2 T

TECHNOLOGICAL POOL+.6.33
R 3250+4.1+2.25

HORIZONTALCHANNEL

Irradiation
Rig

R 2250
REACTOR

12250

SPENT FUEL
+1.1

Transportation
container

-2.85
-2.85

MOLYBDENUM PROGRAM
Transportation technology

Reactor craneTransportation devices

Specialist Meeting
Tokyo,09 Mar 2012

Hot cell craneManipulators

HOT CELLTECHNOLOGYCAL
POOLUnderwatergate to hot cell

Under watertransportation irradiatedtargets from reactor
pool to hot cell- 88 -

20

JAEA-Review 2012-030

MOLYBDENUM PROGRAM
Transportation technology

B

JAEA-NCBJ Meeting
Oarai,06 Mar 2011

A – irradiation holder with targets
B – transportation tool for holders
C – reactor pool transportation container
D – Hot cell reloading deviceCA

1 – transportation basket
2 – support of irradiation holder
3 – irradiated uranium target
4 – transportation container
5 – support of transportation container
3
1

D

2
4
5

HOT CELLMOLYBDENUM PROGRAM
Transportation technology

Specialist Meeting
Tokyo,09 Mar 2012

A

Transportation way container
with irradiated uranium targets
from reactor Hot Cell (A) to
expedition hall (B)B- 89 -

21

JAEA-Review 2012-030

MARIA RESEARCH REACTOR
Molybdenum Program

Specialist Meeting
Tokyo,09 Mar 2012

• General assumption: production rate of 99Mo should be
proportional to fission power generated from uranium
targets in irradiation rig:A – 99Mo activity, P – fission power,  - decay constant 99Mo,
 = 0.0611 – production efficiency of 99Mo per one fission act;
Ef = 3.24410-11 J – generated energy per one fission act:dA P  Adt   / E f• Alfa constant based on physical properties and ORIGEN code
calculation:  5.47 E3 Bq / Ws

• For time irradiation ca 140 h, fission power 185 kW and for 8
LEU targets it can be assumed ca 6500 Ci of 99Mo on the endof cycle (EOI).

MARIA RESEARCH REACTOR
Molybdenum Program

Specialist Meeting
Tokyo,09 Mar 2012

Moly channel f-7

Screen of SAREMA system – reactor MARIA- 90 -

22

JAEA-Review 2012-030

MARIA RESEARCH REACTOR
Molybdenum Program

Specialist Meeting
Tokyo,09 Mar 2012

8000

Test irradiation

7000

Activity [Ci]Aktywnoœæ99 Mo [Ci]6000

 Irradiation time on reactor
nominal power: t= 137.6 h
 Avarage fission power
generated on moly
irradiation channel:
P=199 kW Activity on EOI A= 7665 Ci

5000

4000

3000

2000

1000

0

(based on SAREMA system)

0

20

40

60

80

Czas [h]

100

120

140

160

Time [h]MARIA RESEARCH REACTOR
Molybdenum Program

Specialist Meeting
Tokyo,09 Mar 2012

Current status of uranium targets irradiation- 91 -

23

JAEA-Review 2012-030

Institute of Atomic Energy POLATOM
Specialist Meeting
Tokyo,09 Mar 2012

Photo from the left: V-ce President of Pharmaceuticals COVIDIEN Stephen
Littlejohn, President of EMEA Pharmaceuticals COVIDIEN Christian Wirth, USAambassador Lee Feinstein and Leader of Project Janusz Jaroszewicz of IAE

Second Step of Molybdenum Program
Laboratory for High Activity Radionuclides

Specialist Meeting
Tokyo,09 Mar 2012

• Feasibility Study of the Laboratory for High Activity Radionuclides(LHAR) – the new innovative production technology LEU-basedfor Mo-99 and other radionuclides for nuclear medicine purposes;• Project is presented in scope of Programme Innovative Economy
Development of Centers with High Research Potential - Polish
Ministry of Science and High Education;• The LHAR is destined for the investigation new ways ofpossessive radionuclides extracted from fission products
useful for nuclear medicine: reactor irradiation rig for LEU uranium targets irradiation;
 irradiation and chemical processing targets based on (n,γ) method
 LEU annular target cladded in nickel (cooperation on CRP scope); infrastructure includes research laboratories with 3 hot cells line.

- 92 -

24

JAEA-Review 2012-030

Second Step of Molybdenum Program
Laboratory for High Activity Radionuclides

Specialist Meeting
Tokyo,09 Mar 2012

Laboratoryfor High Activity
Radionuclides (LHAR)
IAE - Swierk

LEU uranium target irradiatedin reactor MARIA in special
molybdenum channel will bedisassembled in reactor hot celland the next transported totechnological hot cells in LHAR

1st Level of LHAR

Participation IAE POLATOM in CoordinatedResearch Project (CRP):„Developing techniques for small-scale indigenous production of Mo-99
using LEU or neutron activation”

Specialist Meeting
Tokyo,09 Mar 2012

Base activity under CRP Program:(Radioisotope Center POLATOM participates on the Program)• Nickel plating of uranium rod and foil experiments in order toobtain fission recoil barriers on LEU metal foils used as targets;• Manufacture of equipment for assembling and perform welding
trials using Al-Mg alloy tubes;• Performing helium leak tests of welded Al tubes;• Performing series of experiments to measure the effect of acidconcentration (HNO3 and also H2SO4) on the recovery 99Mo.

- 93 -

25

JAEA-Review 2012-030

Participation IAE POLATOM in CoordinatedResearch Project (CRP):„Developing techniques for small-scale indigenous production of Mo-99
using LEU or neutron activation”

Specialist Meeting
Tokyo,09 Mar 2012

Welding equipment:
Test welding process
reactor MARIA / POLATOM

Test target after welding
Assembling: Al-Mg alloy tubes,
Ni+Cu+Ni foils

Heat receiver usedduring tubes welding

Participation IAE POLATOM in CoordinatedResearch Project (CRP):„Developing techniques for small-scale indigenous production of Mo-99
using LEU or neutron activation”

Specialist Meeting
Tokyo,09 Mar 2012

Assembling dummy target
Al-Mg alloy tubes,
Ni+Cu+Ni foils
before assembling

Al -Mg alloy tubes after die drawing

By courtesy of OR POLATOM- 94 -

26

JAEA-Review 2012-030

Participation IAE POLATOM in CoordinatedResearch Project (CRP):„Developing techniques for small-scale indigenous production of Mo-99
using LEU or neutron activation”

Leak test of first weld

Specialist Meeting
Tokyo,09 Mar 2012

By courtesy of OR POLATOM

Participation IAE POLATOM in CoordinatedResearch Project (CRP):„Developing techniques for small-scale indigenous production of Mo-99
using LEU or neutron activation”

Specialist Meeting
Tokyo,09 Mar 2012

Results of leak tightness testing of both sides
Second side
Welding parameters

Leak tightness test

Target No
Current[A]

Rotation[rpm]

Heliumpressure
[MPa]

Time
impregnationof helium[min]

Flow ofhelium[μPa m3 sec1]

1

47

3

0,4

30

4x10-3

2

53

3,3

0,35

40

3,9x10-2

3

55

3

0,35

40

5x10-3

According to the Standard ISO 9978 limit of helium flow is:10-2 μPa m3 sec-1 .
By courtesy of OR POLATOM- 95 -

27

JAEA-Review 2012-030

Participation IAE POLATOM in CoordinatedResearch Project (CRP):„Developing techniques for small-scale indigenous production of Mo-99
using LEU or neutron activation”

Specialist Meeting
Tokyo,09 Mar 2012

Upper
side

Microscopic examination of welds
Lower
side

By courtesy of OR POLATOM

Third Step of Molybdenum Program
Program „MOLIBDEN 2010”
based on LEU Targets

Specialist Meeting
Tokyo,09 Mar 2012

General production parameters:• Irradiation source

Research Reactor MARIA 30MW;• Targets:plate or tubular targets LEU / specification has
been made by IEA / NCBJ Świerk (potential producers: CERCA
AREVA, TVEL, INVAP or others);• Irradiation capacity:

2 irradiation rigs (channels), each is loadedwith 8 targets (16 targets irradiated per reactor
production cycle ca 120 h);• Reactor operation:

for discharge of targets reactor needs to be
shut down; reloading targets from irradiation
rigs possible after about 14 hours; resulting in
one discharge procedure per week only;- 96 -

28

JAEA-Review 2012-030

Third Step of Molybdenum Program
Program „MOLIBDEN 2010”
based on LEU Targets

Specialist Meeting
Tokyo,09 Mar 2012

General production parameters:• Production capacity: expected about 1000 Ci (6 days Mo-99) using
2 irradiation position (rigs);• Capability of reprocessing facility : irradiated targets should be
reprocessed by 2 production lines in parallel
each followed by a second run shortly after(limitation of Mo-99 activity decreasing);• Irradiation cycles/y:

expected maximum 30 cycles;• Localization:“turn key” plant, which will be compose a twomain parts: building and production technology
for Mo-99 and technetium generators.

MARIA Research
Reactor
Irradiation facility

Irradiation rigs: (1) - 2 containers x 4 targets
and 2 containers x 4 LEU targets- 97 -

29

JAEA-Review 2012-030

Third Step of Molybdenum Program
Program „MOLIBDEN 2010”
based on LEU Targets

Alternative Molybdenum-99
production process

Specialist Meeting
Tokyo,09 Mar 2012

Specialist Meeting
Tokyo,09 Mar 2012

Today are two primary process for producing 99Mo
 fission of uranium 235U Neutron capture of 98Mo

Comparison of Fission and Neutron Produced 99Mo

- 98 -

30

JAEA-Review 2012-030

Alternative Molybdenum-99
production process

Specialist Meeting
Tokyo,09 Mar 2012

Comparison of Methods – Reactor Position

Method (n,f): cross section for thermal fission of 235U is equal
σ=600b & 6.1% results in 99Mo production (37b);



Methods (n,γ): cross section for thermal flux is equal σ=0.13b
results a factor 300 less than fission process;

Ability to produce large amount of 99Mo from direct reaction
depends upon ability of a high flux reactors that could
compensate for the lower cross section;Additional neutrons are not free and would add costs ofproducing 99Mo by this method;Specific activity for fission-produced 99Mo is 2-4 orders ofmagnitude higher than from the neutron capture process

Alternative Molybdenum-99
production process

Specialist Meeting
Tokyo,09 Mar 2012

Comparison of Methods – Generators Producers Position
Implications for using neutron capture methods on medical
isotope procedure:•necessity redesign generators: large technetium
generators column, sixe and shielding,•large volume of liquid required to eluate 99mTc from thecolumn,•Useful lifetime of generators would be reduced: users
need to order and buy in the same time addition
generators for securing their needs.

- 99 -

31

JAEA-Review 2012-030

Alternative Molybdenum-99
IAE / NCBJ experience

Specialist Meeting
Tokyo,09 Mar 2012

Irradiation molybdenum targets using neutron capture method

•
•
•
•
•
•
•
•
•

Irradiation facility:
Target:Batch:
Irradiation container:
Container closing:
Mass:
Neutron flux density:•

Time of irradiation:

Yield at EOI:EWA & MARIA RRs
MoO3 (powder)3 x vials inside Al container
Θ 25 mm / L=100 mm
cold weld able
3x2,5g1x1014 [n/cm2 s] (MARIA)
8x1013 [n/cm2 s] (EWA)
2.0 Ci/g (MARIA)
1.2 Ci/g (EWA)
87 h (MARIA)
1 week (EWA)

Alternative Molybdenum-99
IAE NCBJ experience

Specialist Meeting
Tokyo,09 Mar 2012

Irradiation procedureQuartz vials with target material was tightly closed inflame, covered in Al foil for brake protection and placedin Al irradiation cans;Irradiation process performed inside channels locatedin beryllium blocks of moderator;
Transport irradiated cans using shielding container
K90 to Radioisotope Center POLATOM in Świerk forchemical processing.

- 100 -

32

JAEA-Review 2012-030

Alternative Molybdenum-99
IAE / NCBJ experience

Specialist Meeting
Tokyo,09 Mar 2012

Processing procedureIrradiated cans are opened using special cutters;Targets has been transferred to a solution flack containing 4N
NaOH solution;
Solution has been heated up to 60-70 oC to fascilate dissolution;
Solution was stirred by passing compressed air;Solution was cooled and passed through a filter to remove any
impurities;
Result: sodium molibdate in sodium hydroxide solution;
99Mo was used as a row material for conversion into medical
product for extraction 99mTc.

Alternative Molybdenum-99
IAE / NCBJ current status

Specialist Meeting
Tokyo,09 Mar 2012

 Most of 99Mo is produce using fission method: 95% of theworld demand is covered by this method; The extremely low specific activity of 99Mo from neutron
capture method makes its use less convenient; Difficulties in terms of safety & waste management arelimitation for production fission 99Mo;

Current status:
NCBJ plans continuation of irradiation of targets HEU & LEU
NCBJ plans design, construct & operate own production
facility based on irradiated in the MARIA RR LEU targets
RR MARIA is ready to restart irradiation 98Mo targets &participate on program neutron of (n,γ) method- 101 -

33

JAEA-Review 2012-030

REAKTOR MARIADLA MEDYCYNY

THANK YOUfor Your Attention- 102 -

34

JAEA-Review 2012-030

Institute of Nuclear Physics
National Nuclear Center
Republic of Kazakhstan

CURRENT STATUS OF (N, GAMMA)
MO-99 PRODUCTION IN KAZAKHSTAN
Batyrbekov E.
Specialist Meetingon Mo-99 production by (n, γ) Method,Tokyo, Japan, March 9-10, 2012

1, Ibragimov St.
050032, Almaty
Republic of Kazakhstan

Tel.: +7 727 3865261
Fax: +7 727 3865260
e-mail: info@inp.kz

Structure
RK Ministry of Industry and New Technologies
RSE “NATIONAL NUCLEAR CENTER OF THE REPUBLIC OF KAZAKHSTAN”
Kurchatov

Institute of Atomic Energy

Kurchatov (827)

Institute of Nuclear Physics(634)
AlmatyAstanaInstitute of Radiation Safety and Ecology

Aksai

Kurchatov (220)Institute of Geophysical Research

Almaty(2471)*

Kurchatov Aktyubinsk(436)

Borovoye Makanchi

Karatau

Kazakh State Center for Blasting Operations

Almaty“Baikal” Enterprise

Kurchatov (185)(22)* – personnel- 103 -

JAEA-Review 2012-030

NNC RK Basic Information
Our mission - scientific and technical supportof the Republic of Kazakhstan peaceful use of
atomic energy policy realization
Strategic directions of activity:






Development of Nuclear Power Engineering
Nuclear Science and Technologies
Radiation Ecology and STS
Support of Non-proliferation Regime
Professional Training for Nuclear Branch

Geography
Borovoye
Aksai

Kurchatov

ASTANA

Aktyubinsk

MakanchiKaratau

IGR Reactor Complex“Baikal-1” Research Reactor Complex
WWR-K Reactor Complex- 104 -ALMATY

IAE
IRSE
INP
IGR
KCBO
“Baikal”

JAEA-Review 2012-030

Institute of Nuclear Physics
in Kazakhstan Nuclear Infrastructure
Ministry of Industry and New Technologiesof the Republic of Kazakhstan

Kazakhstan Atomic
Energy Committee

Instituteof Nuclear
Physics

National Nuclear Centerof the Republic of Kazakhstan

Instituteof Atomic
EnergyAlmatyInstituteof Radiation
Safetyand Ecology

Institute ofGeophysicalResearch

State Holding
Samruk-Kazyna

National JS Company
Kazatomprom

Centerfor blasting
operations

BaikalEnterprise

Kurchatov, East Kazakhstan Region

PRIMARY ACTIVITY AREAS
OF RESEARCH AND DEVELOPMENTS
Nuclear physics
Solid state physics
Applied nuclear physics
and innovative technologies
Accelerator technologies

Reactorinvestigations- 105 -

JAEA-Review 2012-030

RADIOISOTOPE PRODUCTS FOR INDUSTRY

in 1991 – 0
in 2011:
SEALED GAMMA-SOURCES 192Ir, 124Sb, 60Co, 204Tl, 109Cd• defectoscopy• control of technological processes• instruments calibration
ISOTOPIC MARKERS 60Co
• depth marking during wells logging
BULK-SOLUTIONS 57Co, 109Cd• export

RADIOISOTOPE PRODUCTS FOR MEDICINE
in 1991 – only 201Tl
in 2011:DIAGNOSTICS- Medicines with 201Tl, 99mTc, 131I, 67Ga
- Transported 99mТс generator- Diagnostic kits (under tests)- 68Ge (under development)

THERAPY
 Medicines with 131I, 153Sm, 177Lu (under development)

RADIOISOTOPIC TAGS Solutions 85Sr, 134Cs, 60Co, 131I
for ecological and biomedical investigations- 106 -

JAEA-Review 2012-030

INP Basic Facilities

Research Nuclear Reactor WWR-K

Heavy Ion Accelerator UKP-2-1

Heavy Ion Accelerator DC-60

Isochronous Cyclotron U-150M

Commercial Electron Accelerator ELV-4

Research Nuclear Reactor WWR-K
 Commissioned in 1967
 Continuous operation till 1988
 1988-1998 - safety upgrade, licensing Re-started in 1998
 Power: 6 MW, Neutron flux: 1.01014 n/cm2s
 Fuel enrichment: 36% 235U, (under conversion to 19,7%)- 107 -

JAEA-Review 2012-030

WWR-K Reactor Conversion
to LEU 2003…2014
 Reactor operation with low-enriched uranium Improvement of reactor application capabilities
1-4

U-235
36%

1-3
1-2
1-1

3-5

2-3
АР
3-3

2-1
3-2

5-3

6-2
1А
З

6-1
7-2

2А
З
9-3

9-2

8-1
9-1

10-7

9-6
3K
O

10-6
10-5
11-4

10-4
10-3

10-2
10-1

9-8
9-7

8 -6

8-4

8-9
8-8

8-7

8 -5

3К
О

7-10

U-235
19.7%

7-9
2К
О

7-7

7-5

8-2

7-1

6-8

7-6

7-4

6-9

6-7
6-6

6-5

2К
О

5-2
5-1

5-6

6-4

5-10
5-9

3А
З
5-7

4-5

5- 5
5-4

4-2
4-1

4-9
4-8

4-7
1К
О

4-4
1К
О

3-1

3-8
3-7

3-6

2-4

2-2

2-7
2-6

2-5

11-3
11-2

11-1

5-tube fuel assemblies UAl4–Al
Water reflector
1·1014 n/(cm2s)

8-tube fuel assemblies UO2–Al
Beryllium reflector
2·1014 n/(cm2s)

Tc-99m 80% of diagnostic nuclear medicine
 Daughter product of Mo-99, pertechnetate solution
produced in Mo-99/Tc-99m generators
 Most of global production is based on high specificactivity Mo-99 produced by uranium fission andcompact chromatographic column generators

Global crisis in supply of fission product Mo-99
 Unprecedented interest to alternative sourcesof Mo-99 and/or Tc-99m
 Various options
 Special attention to low specific activity “(n,)”Mo-99 produced by neutron activation of Mo-98(natural or enriched)- 108 -

JAEA-Review 2012-030

Use of (n,) Mo-99 to Produce Tc-99m
 Variable options (solvent extraction, ion exchange,
sublimation …)
 Attractive solution: compact generator column
containing much more molybdenum than usual
chromatographic sorption column, but usedin the same way (pertechnetate is simply eluted
with normal saline) High efficiency inorganic sorbents
of molybdenum Poly-Zirconium compound (PZC), Poly-Titanium
compound (PTC)…
 Molybdenum compounds

 …, Zirconium-Polymolybdate Gel

The Main Steps in Mo-99/Tc-99m Production
 1998 - recommission of the nuclear reactor, startof Mo-99 and I-131 production for developmentof technologies for local productionof radiopharmaceuticals with Tc-99m and I-131
 2000-2001 preclinical and clinical trails, registrationof “Sodium pertechnetate 99mTc, injection” from
central Mo-99/Tc-99m gel-generator
 2002-2011 - commercial supply of ready to use 99mTc
solution to Almaty hospitals
 2005 - approval of feasibility study for Establishmentof a Center of Nuclear Medicine and Biophysics- 109 -

JAEA-Review 2012-030

The Main Steps in Mo-99/Tc-99m Production
 2005 - start of development of portable 99mTc gel-generator
production technology under IAEA support (CRP “Developing
techniques for small scale indigenous Mo-99 production Using LEU
fission or NA”)

 2006 - start of the project “Establishment of a Centerof Nuclear Medicine and Biophysics”, reconstructionof radiochemistry facilities in INP NNC RK
 2009 - start o construction of new radiopharmaceutical building
in frame of Center of Nuclear Medicine and Biophysics
 2009-2011 - preclinical, clinical trails and registration of “Sodium
pertechnetate 99mTc injection” from portable gel-generator 2012 - commercial supply of portable 99mTc gel-generator
to the hospitals

Status of Gel Technology
Developed in Australia and USA (Evans, Boyd …), 1970-s…1990-s
Routine use of gel generators:




China
India
Kazakhstan

New production facility:BrasilKazakhstan:Central

Portable(India design)- 110 -

since 2001 – routine use of central
gel generator
2010 - switching to portable gel
generators

JAEA-Review 2012-030

99Mo/99mTc Gel Generator Concept
Direct conversion of neutron
irradiated Mo into insoluble
zirconium polymolybdate gelthat contains 25-30 wt.%of molybdenum.

x(ZrO)2+· y(99MoO4)2-· z(H2O)
24 hours

The gel is predominantly
a cation exchanger with
an open structure which allows
free diffusion of pertechnetateanion under simple elution
by saline.

99MoO )24x(ZrO)2+· y ( 99m· z(H2O)( TcO4)-NaCl
9924) · z(H O)
x(ZrO)2+· y ( MoO
2
Cl-

+

Additional Al2O3 column
is needed to prevent
molybdenum break-throughNa+(99mTcO4)-

99mTc Gel Generator Production

Generator column matrix - Zirconium molybdenum gel contains ~25% Mo with (n,)99Mo
 99mTc eluted in normal saline
 additional alumina column is needed to prevent
molybdenum break-through
0,9 % NaClx(ZrO)2+ y(99MoO4)2(99mTcO4)-· z(H2O)

Al2O3Central generator(Almaty)Na+(99mTcO4)-- 111 -

Portable generator

JAEA-Review 2012-030

99mTc Gel Generator Production

Quality of Tc-99m Solution from Portable
Gel-Generator
99mTc fromgel generator

European
Pharmacopeia

Content of radionuclide impurities, % 99Mo other radionuclide impurities≤ 0,02≤ 0,002≤ 0,1≤ 0,01

Content of inactive impurities, µg/ml
 Al
 Mo
 Zr

≤ 2,0
≤ 5,0
≤ 5,0

≤ 2,0

Radiochemical purity,%≥ 99,0≥ 95,0рН

4,0 - 7,04,0 - 8,0Parameter

Sterility
Pyrogenity

SterileSterilePyrogen freePyrogen free

99mTc solution activity on the 2-nd day after generator production (3-d after irr.):

0.5-0.6 Ci with natural molybdenum irradiated 4-5 days in 1·1014 n/(cm2s)
The same generator can easily accommodate larger column → 1 Ci activity
85% of 99mTc activity in first 5 ml → 2-3 GBq/ml- 112 -

JAEA-Review 2012-030

Status of National Project “Establishment ofa Center of Nuclear Medicine and Biophysics”
 2009-2012 – Construction of new building forradiopharmaceutical production ( construction
is done; finishing, equipment installation, on site
testing and put into operation will be done in 2012) 2008-2012 - Design, production and put into
operation of the equipment for industrial
production of 50 portable 99mTc gel-generator
in new building (design, production and FATof hot cells and equipment was passed)
 From 2014 - Production of 99mTc gel-generators
in GMP condition

2009
Existent Radiochemical Building in INP-Almaty- 113 -

JAEA-Review 2012-030

2010
New Radiopharmaceutical Production
Building in INP-Almaty

2011
New Radiopharmaceutical Production Buildingin INP-Almaty (to be completed in 2012)- 114 -

JAEA-Review 2012-030

99mTc Gel Generator (INP design)
ED 08

Production from 2014
glazingglazing

1.18a
Mo-99 target
receivingroom, D

D1/20

glazing

D1/21

D1/22

Air-lock of Tc-99m
generator
production area,
D/Ñ

Lead container
disinfection, Dglazing

D1/25

Tc-99m generator production
area, C

1.18

Waste exit,
NC

1.31

Quarantine, storage
and packing areafor products, NC

1.17
ED 01

ED 02

Lead container
cleaning, NC

1.11

Dosimetry control

glazingglazing

D1/26

ED 03

Status of Mo-99 Production
for International Research
3 joint experiments with JAEA scientists were done at INP with usingof 99Mo produced at WWR-K from natural molybdenum oxide,
1·1014 n·cm-2·c-1,100 h.
Objective: Compare performance of PZC generator column,
characteristics of obtained 99mTc eluate with both Gel column inmodel conditions and study recovery of Mo from used PZC and Gel.
PZC generator column:
 PZC was prepared in Japan, grain size 300 - 710 um.
 Model molybdenum solution prepared in Kazakhstan(molybdenum concentration 25 mg/ml; activity concentrationof 99Mo 600 kBq/ml, рН =6,37).
Gel generator columns:
 99Mo-Zr Gel was taken from real production batches after decay.
Grain size of gel particles (from previous experiments):≤200 µm - 65%; 200-500 µm - 32%; 500-700 µm - 2%; ≥700 µm - 1%.

- 115 -

JAEA-Review 2012-030

Results of Joint Experiments
1.

PZC-based generator can be utilized for production of 99mТс
solution with high radioactive concentration using low specificactivity 99Mo ( 0,8-1,2 Ci/g from natural molybdenum
or 3,2- 4,8 Ci/g from enriched molybdenum-98).
Procedure of 99mTc elution is “customer friendly” like in classic
chromatographic 99mTc generator and Gel generator.

2.

Less hot cell operations, less requirements to hot cell equipment
in case of PZC compared to Gel.

3.

Both 99mТс yield and impurities content are higher in solution from
PZC-based generator in model conditions – need to be provedin practice. Alumina “safety” column removes the impurities
efficiently. Alumina column is necessary also for pH adjustment.

4.

Molybdenum content is twice lower in PZC column than in Gelcolumn of the same volume, elution profile is a bit broader.
Change of grain size distribution would possibly improve this.

5.

Organic solvent is used in synthesis of PZC. Experimental proofof absence of organic components in the 99mTc solution
is needed.

Further Activities
to support nuclear medicine in Kazakhstan: Continue of commercial productionof portable gel-generators 10 gens/bath,18-25 GBq Put into operation the Center of Nuclear
Medicine and Biophysics
 Put into operation the new line for
production of 50 gel-generators per batchin GMP condition- 116 -

JAEA-Review 2012-030

Further Activities
to increase Mo-99 production:
 Calculations and experiments to increase Mo-99
specific activity (neutron spectrum, self-shielding) Optimization of irradiation rig for large quantities
of molybdenum (hundreds of grams)
 Development technique for Mo recovery fromused gel-generator columns

 Metal molybdenum option for more efficienttransportation

Thank you for yourattention!

- 117 -

This is a blank page.

JAEA-Review 2012-030

EXPERIENCE OF THE INSTITUTE OF
NUCLEAR PHYSICS ON Tc-99m GELGENERATOR PRODUCTION
Chakrova E.
Specialist Meeting on
Mo-99 production by (n, γ) Method,Tokyo, Japan, March 9-10, 2012
1, Ibragimov st.
050032, Almaty
Republic of Kazakhstan

Tel.: +7 727 3865261
Fax: +7 727 3865260
e-mail: e.chakrova@inp.kz,
echakrova@gmail.com

Content:
1. Irradiation facilities, target for irradiation
2. 99Mo target processing
3. Technical standard for 99mTc solution
4. Quality control of 99mTc solution
5. Waste
6. Recovery of molybdenum- 119 -

1

JAEA-Review 2012-030

1. Irradiation facilities
 Commissioned in 1967

 Continuous operation till 1988
 1988-1998 - safety upgrade, licensing Re-started in 1998
 Power: 6 MW, Neutron flux: 1.01014 n/cm2s
 Fuel enrichment: 36% 235U, ( under conversion to 19,7%)

Irradiation channels in the WWR-K research reactor
Number ofchannels

Diameter,mm

Max. thermal
neutron flux,ncm-2s-1

3

62(1.0 ÷ 1.1)1014

variable62(3 ÷ 6)1013

12

70 ÷ 200

1012 ÷ 1013

Radial5100

21013

Tangential

1

193

61012

Thermal column cavity

1

1150

11013

Vertical channels
Core centerCore periphery
Reflector (light water)
Horizontal channels

Attached facilities:• Critical assembly,• Gas-vacuum loop
• Neutron radiography• Hydraulic and pneumatictransfer systems• Hot cells- 120 -

2

JAEA-Review 2012-030

Reactor WWR-K

Vertical in-core
irradiation positions
Ø 62 mm, H 600 mm

Reactor WWR-K
Vertical in-core channels- Thermal neutron flux

132

Core center
11.11014 n ּ◌cm-2 ּ◌s-1.81012 n ּ◌cm-2 ּ◌s-1

REACTOR CORE CENTERPLANE

Core periphery8.51013 n ּ◌cm-2 ּ◌s-1

3.71013 n ּ◌cm-2 ּ◌s-1

6.71013 n ּ◌cm-2 ּ◌s-1

3.01013 n ּ◌cm-2 ּ◌s-1- 121 -

3

JAEA-Review 2012-030

Target for irradiation
Natural molybdenum oxide, MoO3,
Isotopic abundance of Mo-98 in natural Molybdenum:
24.13%
Quartz ampoule → alumina foil
→ alumina container
98Mo (n, γ) 99Mo

Thermal neutron capture crosssection: 0.13 barns
Epithermal neutron capture crosssection: 6,9 barns
E ≤ 0,5 eV - 40 % of total 99Mo activity

WWR-K reactor conversion to LEU

2003…2014

Reactor operation with low-enriched uraniumImprovement of reactor application capabilities
1-4
1-3
1-2
1-1

U-235
36%

2-4
2-3

3-5

3-3

2-1
3-2

6-2

7-2

8-1

9-3
9-2

9-1

5-tube fuel assemblies
UAl4–Al
Water reflector

1·1014 n/(cm2s)

10-7

9-6
3K
O

10-6
10-5
11-4

10-4
10-3

10-2
10-1

9-8
9-7

8 -6

2А
З

8-9

U-235
19.7%

8-8
8-7

8 -5
8-4

8-2

7-1

7-7
7-6

7-5

7-10
7-9

2К
О

6-6

3К
О

6-9
6-8

6-7

6-5

7-4
1А
З

6-1

5-7

6-4
2К
О

5-2
5-1

3А
З

5-6
5- 5

5-3

5-10
5-9

4-7

4-5

5-4

4-2
4-1

4-9
4-8

1К
О

4-4
1К
О

3-1

3-8
3-7

3-6

АР

2-2

2-7
2-6

2-5

11-3
11-2

11-1

8-tube fuel assemblies
UO2–AlBeryllium reflector

2·1014 n/(cm2s)- 122 -

4

JAEA-Review 2012-030

Irradiation Test of New Core
1 -4

2 -7

1 -3
1 -2

2 -3
2 -2

3 -6

4 -1

6 -7

1

2

7 - 107 -8Beryllium carrier with 3 LTAs

8-6
7 -4

7 -3

8 -5
8 -4

3РР

7 -2
8 -2
8 -1
9 -1

10

10

-1

9-7

9-5

10

-3

11

-1

-2

-7

7-6

FA-1 (5-tube)1РР

FA-2 (3-tube) with control
rod inside

-6

6

1 2

1 3

5 4

-5
11-4

11

11

10

11

10

2АЗ

9 -3

9 -8

9 -6

9 -4

9 -2

8 -9
8 -8

1РР

6 -3

7 -1

4

7 -9

7-7

3

1АЗ

6 -1

5

1

5 -4

5 -2

6

2

6 -9
3АЗ

5 -7

5 -3

5 -1

3

5 -9
5 -8

4 -6
4-5

2 13
1 1 4
5 6 5
52

1 3

5 - 10

4 -8

4 -4

4 -2

4 -9

2РР

3 -4

1РР

3 -1

3 -7

3 -5

3 -3
3 -2

3-8

2 -5
АР

1 -1

2 -1

2 -6

2РР

-3

-2

9-1

4- 5

LTA (WWR-KN)
Channel housing shim rod(2РР or 3РР)Beryllium block
Irradiation tube

Introduction. 99mTc gel-generator
conceptx(ZrO)2+·y(99MoO4)2-·z(H2O)
24 hours

target -natural MoO3
irradiation – 120 h

99MoO )24x(ZrO)2+ ·y ( 99m( TcO4)-·z(H2O)NaClx(ZrO)2+ ·y

98Mo (n, γ) 99Mo

(99MoO4)2- ·z(H O)
2
Cl-

+

Na+(99mTcO4)-

thermal flax - 1·1014 n·cm-2·s-1
epithermal neutron flax -0,8·1013 n·cm2·s-1
Specific Activity of 99 Mo -0,8 Ci/g
Period from EOB to EOP - 1 day
Activity of 99mTc – 18-25 GBq,
2 days precalibration- 123 -

5

JAEA-Review 2012-030

Production of 99Mo-Zr gel
Process step“Central”“Portable”
20/batch

Irradiation, 6MW,
1.1- 2.11014n/sm2·c

Dissolution of MoO3

10g MoO3
40g MoO3
3-7 days
7-10 days
2M NaOH, H2O2 2M NaOH,H2O2

Synthesis of PMNa

3M HNO3 (pH 4) 3m HNO3, (pH 4)

Synthesis of PMZr

ZrOCl2, 5-7%
ZrOCl2 , 5-7%
solution, pH 4,5 solution, pH 4,5
30 min30 minPrecipitation

Production of 99Mo-Zr gel
Process step“Central”“Portable”
20/batch

Filtration

1 hour

0.25 hour

DryingWashing /disintegration

6 hour,≤100ºC
Cool water, 310ºC, saline

4 hour,≤100ºC
Cool water, 310ºC, saline

Gel dispensing-

by volumein 6 g

40 ml saline

5-10 ml saline

Elution- 124 -

6

JAEA-Review 2012-030

Generator column matrix - Zirconium molybdate gel- contains ~25% Mo with (n,)99Mo- 99mTc eluted in normal saline- additional alumina column is needed to prevent
molybdenum break-through
0,9 % NaCl
Portable generator(99MoO4)2-x(ZrO)2+ ·y(99mTcO4)-·z(H2O)

Al2O3

Central generator(Almaty)Na+(99mTcO4)-

99mTc gel generator (INP design)

production from 2014
glazingglazing

1 .1 8 a
M o -9 9 ta r g e t
re c e iv in gro o m , D

D 1 /2 0
glazing

D 1 /2 1

D 1 /2 2

A ir- lo c k o f T c -9 9 m
g e n e ra to r
p ro d u c tio n a re a ,D /Ñ

L e a d c o n ta in e r
d is in fe c tio n , D

glazing

D 1 /2 5

T c - 9 9 m g e n e ra to r p ro d u c tio narea, C1 .1 8W a s te e x it,
NC

1 .3 1Q u a ra n tin e , s to ra g e
a n d p a c k in g a re afo r p ro d u c ts , N C

1 .1 7
E D 01

E D 02L e a d c o n ta in e r
c le a n in g , N C

1 .1 1

Dosimetry control

glazingglazing

D 1 /2 6

E D 03- 125 -

7

JAEA-Review 2012-030

Technical standard for 99mTc solution

Technical standard : Kazakh Pharmacopoeia
Monograph ФС РК 42- 930-06
1.

Parameter
Description

2.

Mechanical inclusions

3.

Content of NaCl, mg/ml

8 - 104.

Volume activity, MBq/ml

17-3700

5.

Nominal volume, ml

6.

pH

7.

Content:
99Mo, %
Other gamma-impurities, %Radiochemical
purity, %Chemicalpurity, ppm
Sterility

8.
9
10

Level
Colorless, clear solution
Not identified≤10
4.0 – 7.0
≤ 0.02≤0.002
99.9
Mo≤ 5, Zr≤ 5, Al≤2
Sterile- 126 -

8

JAEA-Review 2012-030

99mTc gel generator production,

quality control

QC of raw materials and chemicals(provided by QC Lab and National Center of drug expertise according to
Kazakh Pharmacopoeia)
In process control:(provided by QC Lab according to developed SOP)
99Mo activity concentration, pH of Zr solution, sodium polymolybdate 99Mo
solution, zirconium polymolybdate 99Mo solution, temperature of drying
air, Microbiological control of working area (air, surface of equipment,
clothes)
QC of product “Sodium pertechnetate, 99mTc injection”, 99mTc gel-generator(provided by QC Lab according to SOP based on Kazakh Pharmacopoeia)•
Identity, pH, Mechanical impurities, Radionuclide impurities,
Radiochemical purity, Chemical impurities, Content of NaCl, Activity
concentration, Sterility

Quality parameters of 99mTc solution from portable
and central gel-generator
Parameter
Color

Pharmacopoeia(42- 930-2006)

Central gel-generator(2001-2010)

ColorlessPortable gel-generator(12 batches)

Colorless

Colorless

Mechanical inclusions Not identified

Not identified

Not identified

Content of NaCl, 8 - 10
mg/ml
Volumeactivity, ≥17
MBq/mlNominal volume, ml 10

8.8 - 9.2

8.9 - 9.3

130-2200

170-3275

1 - 1010pH

6.0 - 7.0

6.0 - 7.0

Not identified
0.00002 – 0.00008

Not identified
0.00003 – 0.00005

99.1-99.9

99.2-99.9

SterileSterile4.0 – 7.0

Content:
99Mo, %
≤ 0.02Other
gamma- ≤0.002
impurities, %Radiochemical
99.9
purity, %Sterility
Sterile- 127 -

9

JAEA-Review 2012-030

Chemical impurities of 99mTc solution fromportable and central gel generators
Impurities, Pharmacopoeia Pharmacopoeiappm(42- 930-2006)(42- 930-2006)

Al
As
Ba
Be
Bi
Cd
Cr
Cu
Fe
Hg
Mn
Mo
Ni
Pb
Sb
Sn
Zn
Zr

2.0
1.0
0.1
0.005
0.1
1.0
0.1
0.05
0.25
2.5
0.01
5.0
0.1
0.1
1.0
0.1
2. 5
5.0

Centralgel-generator(2001-2010)

Portablegel-generator(9 batches)

0.001 - 0.1
0.0050.026 - 0.085
0.00005
0.0002
0.0002 0.0005
 0.0005
0.005
0.0001
0.005 - 0.092
0.005 - 1.0
0.008 - 0.075
0.05 - 0.95
0.0002
0.0005
0.284 - 0.690<0,00020.001 - 0.05
0.0050.052 - 0.167
0.00005
0.0002
0.0002 0.0005
 0.0005
0.005
0.0001
0.005 - 0.015
0.005 - 0.182
0.025 - 0.082
0.03 - 0.09
0.0002
0.0005
0.315 - 0.805<0,0002

2.0
1.0
0.5
1.0
0.1
0.05
0.25
0.03
5.0
0.1
0.2
1.0
0.1
2. 5
5.0

Proven quality (portable gel generator,
Almaty)99mTc fromgel generator

European
Pharmacopeia

Content of radionuclide impurities, % 99Mo other radionuclide impurities≤ 0,02≤ 0,002≤ 0,1≤ 0,01

Content of inactive impurities, µg/ml
 Al
 Mo
 Zr

≤ 2,0
≤ 5,0
≤ 5,0

≤ 2,0

Parameter

Radiochemical purity,%≥ 99,0≥ 95,0рН

4,0 - 7,04,0 - 8,0

Sterility
Pyrogenity

SterileSterilePyrogen freePyrogen free

99mTc solution activity on the 2-nd day after generator production (3-d after irr.):

0.5-0.6 Ci with natural molybdenum irradiated 4-5 days in 1·1014 n/(cm2s)
The same generator can easily accommodate larger column → 1 Ci activity
85% of 99mTc activity in first 5 ml → 2-3 GBq/ml- 128 -

10

JAEA-Review 2012-030

99mTc gel generator production,waste• During production of gel-generators the low active liquidand solid waste are generated• Practically: for 1Ci 99Mo - 0,5 l of liquid waste (187W –0,3mCi, T 1/2 -24 h; 92mNb– 0,3mCi, T 1/2 -10d,
24Na- 0,3mCi, T 1/2 -15 h; 99Mo - 50mCi, T 1/2 -66 h;)• Solid waste – ampoule material, used 99Mo-Zr gel,disposable materials• Gas waste are not generated

Molybdenum recovery from used gel
Mass, mg

Mo:Zr ratio, %
(RFA)

1000

51:49

Primary Mo-Zr gel

1st Test with NaOH
1. Dissolution

2. Analysis of Mo-Zr gel after recovery of Mo

Solvent

The mass of Mo-Zr gel
after elute, mg

Solvent

Dissolved Mo, %(-ray spectrometry)

Mo:Zr ratio, %(RFA)

Concentration of Mo,
mg/ml

1M NaOH640

1M NaOH

94.95

5:95

5.27

0.5M NaOH

654

0.5M NaOH

92.90

8:92

5.16

2st Test with NH4OH
1. Dissolution

2. Analysis of Mo-Zr gel after recovery of Mo

Solvent

Solvent

The mass of Mo-Zr gel
after elute, mg

Dissolved Mo, %(-ray spectrometry)

Mo:Zr ratio, %(RFA)

Concentration of Mo,
mg/ml

1M NH4OH

682

1M NH4OH

64.20

28:72

3.56

0.5M NH4OH

679

0.5M NH4OH

69.00

31:69

3.80

- 129 -

11

JAEA-Review 2012-030

Conclusion:• Activated Mo-99 with low specific activity (0,5- 1,0 Ci/g)can be effectively used for production of Tc-99m
generators due to application of Gel-technology• Gel-technology can be effectively used for production
both central and portable Tc-99m gel-generators• High quality of Tc-99m solution• Small quantity of radioactive waste, no gas waste, nohigh level radioactive waste- 130 -

12

2012/6/20
JAEA-Review 2012-030

Current Status on Production of 99Mo
in Indonesia
A. Mutalib
Center for Radioisotopes and Radiopharmaceuticals
National Nuclear Energy Agency
Kawasan PUSPIPTEK, Serpong

Outline Presentation• National Nuclear Energy Agency (BATAN) and PT Batan Teknologi(Batantek)• R&D Programs at the CRR• Development of 99Mo/99mTc Generators• Development of 99mTc Extractor combined with two chromatograpic
columns- 131 -

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2012/6/20
JAEA-Review 2012-030

BATAN and PT Batantek
 BATAN

A government R&D Institute under coordinationof Ministry of Research and Technology

 PT Batantek

A state-owned company under supervision of
Ministry of State Enterprises
Producer:• Fission Product 99Mo
• 131I• 192Ir sealed-source for NDT (industry)• Fuel element for nuclear research reactor

BATAN has the following functions :• To assess and prepare the national policy in the field of research,development, and the beneficial uses of nuclear energy.

• To develop and utilize production
technology of radioisotopes andradiopharmaceuticals for healthcare.

Center for Radioisotopes and Radiopharmaceuticals- 132 -

2

2012/6/20
JAEA-Review 2012-030

Cyclotron Technology

R&D Programs at the CRR

Sealed-Sources

Cyclotron CS30

Radioisotope Production
Technology

Unsealed-Sources• Radiotherapy
137Cs, 60Co
• Brachytherapy:
125I, 103Pd, 192Ir

Radiopharmaceuticals

Radiolabeled Compounds
Radionuclide Generators
Radiopharmaceutical Kits

RIA/IRMA Kits
Molecular Radiotracers

Nuclear Reactor 30 MW

Irradiation Facilities at the G.A Siwabessy Reactorip

BERYLLIUM BLOCK REFLECTOR

S= Beam Tube

NRS = Normal Rabbit System
Fuel Assembly
CIP = Central Irradiation Position
Beryllium Reflection Element
IP= Irradiation Position (core)IR= Irradiation Position
FRS = Fast Rabbit System
PRTF = Power Ramp Test Facility
NRF = Neutron Radiography Facility (out of core)- 133 -

3

2012/6/20
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Radiopharmaceuticals for Molecular Imaging dan “Targeted Therapy”
HOSPITALPET Radiopharmaceuticals
of 18F,124I, 96Zr, 64Cu

18F,124I,

GMP basedradiopharmaceuticalproduction facility

96Zr,123I,

111In, 64Cu

PET-CT
SPECT
radiopharmaceuticals
of 123I,111In,
99mTc, 131I

Cyclotron CS30

SPECT-CT
99mTc,131I,

Therapeutic
radiopharmaceuti177Lu,186/188Re,
cals of 131I,153Sm,
64Cu
166Ho,177Lu,
186 /188Re
Radiopharmaceuticals based on simple organic
molecules and biomolecules (peptides,
monoclonal antibodies, hormones, etc.)
153Sm,166Ho,

Nuclear Reactor

Therapy

Development of 99Mo/99mTc Generators
TargetNatural molybdenum

Uranium-235
HEU(UO2) LEU (metal foil)

98Mo(n.g)99Mo

235U(n, f) 99Mo + FP

dissolution~ 8 hrs

precipitation

Zr(NO3)4

Purification I

DryingPurification II~ 7 hrs

PZC~ 5 hrs

Crushing
Grinding

99MoO 24

Sieving
99mTcO 4~ 2 hrs
F.P based
Generators

PZC based
Generators

Alumina
Gel based
Generators
99mTcO 4

99mTcO 4- 134 -

42012/6/20
JAEA-Review 2012-030

PZC Based 99Mo/99mTc Generator

99mTc Extractor combined with two chromatographic columns(n, )99Mo

MEK

99Mo/99mTc

solution

Base Alumina

MEK

Aquabidest
NaCl 0,9%(Saline solution)
Radiopharmaceutical kits:
MDP, DTPA, MIBI,
Acid Alumina Tetrafosmin, HYNIC-Folate

Labeling

Reused for the
next extraction
Aqueous phase:
Na299MoO4

Organic phase:
MEK-99mTc

Injectable
MEK waste Aqueous
waste Na99mTcO4 Radiopharmaceuticals- 135 -

5

2012/6/20
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Thank youfor your attention- 136 -

6

2012/6/20
JAEA-Review 2012-030

Irradiation Facility
at the RSG-GASRSG GAS- 137 -

1

2012/6/20
JAEA-Review 2012-030

GA. SIWABESSY REACTOR
Location
Operator
Construction
First criticality
Reactor Type
Nominal Power
Fuel
Moderator/Coolant
Reflector

Serpong, Indonesia
Center for Multi Purpose
Reactor GA. Siwabessy
started 01/01/1983

07/29/1987

Open Pool Light Water
Reactor
30 MW
MTR type, U3Si2-Al
Light Water
Be, H2O

Irradiation Facilities at the G.A Siwabessy Reactorip

BERYLLIUM BLOCK REFLECTOR

S= Beam Tube

NRS = Normal Rabbit System
Fuel Assembly
CIP = Central Irradiation Position
Beryllium Reflection Element
IP= Irradiation Position (core)IR= Irradiation Position
FRS = Fast Rabbit System
PRTF = Power Ramp Test Facility
NRF = Neutron Radiography Facility (out of core)- 138 -

2

2012/6/20
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Table Irradiation Facility at the GA. Siwabessy

ReactorFacility
CIP

Position
D‐6, D‐7, E‐6,E‐7

Application
Irradiation, RI production

IP1
IP2

D‐6
D‐9

Irradiation, RI production ,topaz
Irradiation, RI production,topaz

IP3
IP4
PRTF

E‐4
G‐7
Out‐coreNR

Out‐core

Irradiation, RI production,topaz
Irradiation, RI production,topaz
Power ramp test for small pin of
PWR/BWR fuelNeutron radiography of fuel material.

Irradiation
Holes

IR‐1 (A‐2)
IR‐2 (K‐3)
IR‐3 (K‐10)

Irradiation
irradiation
irradiation

Table Irradiation Facility at the GA. Siwabessy Reactor (cont’)
Facility
Beam Tubes

Rabbit System
NTD dan topaz

Position

Application
S-1
Iodine loop, I-125 radioisotope
production
S-2
Neutron Radiography
S-3
Not used
S-4
TAS, triple axis spectrometry
S-5
Neutron guide equipped by FCD, SANS,HRSANS and HRPD.
S-6
PD (powder diffractometer).
B-1, C-1, D-1, E- NAA, irradiation
1, F-1.
Out-core
Irradiation of silicone crystals.
Max. diameter of sample 7 ‘and lengthof 40 cm

- 139 -

3

2012/6/20
JAEA-Review 2012-030- 140 -

42012-06-20
JAEA-Review 2012-030

Specialist Meeting on Mo-99 Production by (n,gamma) Method

Current status of research on the (n,γ)Mo/Tc-99m
generator as a mid-term approach2012. 03
Ul-jae Park
Radioisotope Research Division

99mTc

Problem and Solutions
Approach to Secure 99mTc Supply in Korea
Why problematic?
Actions to be taken
Currently Available Technologies in Korea
Options that is in progress
Status on New Reactor
Summary2- 141 -

1

2012-06-20
JAEA-Review 2012-030

99mTc?

• Technetium-99m(99mTc)– Decay Product of 99Mo
– Most important radioisotope fordiagnosis in nuclear medicine– More than 80% of total practices
performed with this isotope– Generally, prepared fromgenerators

Problems?
Country

Reactor

1st
Criticality

MarketShare (%)

StatusCanada

NRU*

1957

38

Operation (Problem)

1961

26

• Netherland
Situation:
HFR

Stop

(‘08.08.- ‘09.02.)– In 2008, short supply of 99mTc became an
Stop

Belgium

BR2
1961
16
('08.09. - 11.)international
issue
FranceOSIRIS
196699m~
Operation– StableFRJ-2
supply of
Tc no~ longer guaranteed
Germany
1962
Operation
S. Africa

SAFARI

1965

16

Operation• Primary Cause:– Unscheduled or prolonged stops of reactors
producing 99Mo by fission– Aging of the reactors but no new ones available– No alternative technologies available for 99mTc- 142 -

2

2012-06-20
JAEA-Review 2012-030

Crisis5

Solution• No immediate solution available unless new
production reactors constructed to take the duty
over from the old ones.
• Some available technologies may be applicablefor the production of 99mTc at a certain extent
until a new reactor is ready.
• Combination of such available technologies andthe construction of a new reactor is a goodstrategic approach to meet 99mTc demand.

- 143 -

3

2012-06-20
JAEA-Review 2012-030

Approach to Secure 99mTc Supply in Korea

Currently Available Technologies in Korea• Solvent extraction of
99mTc from (n,γ)99Mo– Practiced since 1970th• New Adsorbent for
generators– Recently demonstrated theperformance for 188W/188Re
Generator

Solvent Extraction Unit

• On the basis of (n,γ)99Mo from HANAROHigh Performance Adsorbent- 144 -

42012-06-20
JAEA-Review 2012-030

Option I: Solvent Extraction
Basic idea
Methylethylketone (MEK) easily reacts
with 99mTc but not dissolvable in water– Hot cell production
99mTc: 10Ci (99Mo)/week

Limited by intensive labor andtransportation– On site production
Need new production units
Need operator training– Considered as a temporary
expedient

Hot Cell Unit
On Site Unit

Option II: New Adsorbent for Generator• A high performance adsorbent demonstrated.

Performance of Adsorbent
100 x Normal Alumina

1Ci 188W/188Re generator
188W Loaded on column : 1Ci

Extraction efficiency : 75~90%
Impurity : ~10-4 %- 145 -

5

2012-06-20
JAEA-Review 2012-030

Why problematic?

Still many reactors available for (n,γ)99Mo !
HEU
(n,f)

HANARO(n,γ)
V. VialSaline99Mo

99Mo
Sp. Ac.
(Ci/g)

99mTc

Low,
(1.5)

High,
(1,000 ~10,000x)

Generator has a limited space toaccommodate molybdenum.

Option II: How it works
a

O
O

S

O

+

O

O
O

Alumina back bone

b

O

1-

O
Re
O O

2-

W

O

O

O

O

+

O

O
O

2-

S

O

O

Alumina back bone

+ NaCl+ H 2O

Alumina back bone

W

O

OH

+ NaReO 4+ HCl

Alumina back bone

Adsorption of 188W (99Mo)and desorption of 188Re (99mTc) on the adsorbent

Korea atomic energy research institute- 146 -

6

2012-06-20
JAEA-Review 2012-030

Option II: KAERI Generator Column• Uses just a pre-packed column: As easy as the production of (n,f)99Mo/99mTc generator
Within seconds

Compartment of Adsorbent
Loaded with 99Mo

Pass 99Mo solution
through column
Solution

Normalized Activity

1.2
1.0

100%

0.8

70%
30%

0.6

0.0

Assembling

15%

0.4
0.2

Column Washing

Adsorbent Layer

Inert

-0.2

0.0

0.2

0.4

0.6

Inert

0.8

1.0

1.2

Normalized Distance

Option II: Progress of Research• Production rate: 60Ci/week• Generator capacity: 0.5Ci/ea (HANARO IP hole)• Research Schedule:
Current Stage(2012)- 147 -

7

2012-06-20
JAEA-Review 2012-030

Option II: Progress of Research

Option II: Progress of Research

1 hr
3 hr
99mTc-MDP, SPECT

1 hr

3 hr

99mTc-MDP, SPECT/CT- 148 -

8

2012-06-20
JAEA-Review 2012-030

Option II: Progress of Research

Concentration Devices for the Re-usage
99mTc-concentration system(Shield : Lead and Plastic)Na+ (MO4)- in saline(5-10ml)
M = 188Re, 99mTc
Na+
Ion exchange reaction
Ag+
NaCl removal column
ClAgCl ↓Anion Exchange resin
Adsorption :(MO4)-

0.5~1ml of Saline(MO4)Na+ (MO4)-

L-Prototype

I-Prototype

Animal SPECT/CT image

99m Tc-MDP image using

commercial available
generator(SAMYOUNG
Unitech.) and RI
concentration system

99m Tc-MDP image using(n,r)99mTc generator
developed in KAERI and
RI concentration system

Option III: New Reactor and LEU Target• Can resolve all problems.
• Becomes a main RI supplier.

- 149 -

9

2012-06-20
JAEA-Review 2012-030

Option III: New Reactor and LEU Target
Project Period

2012 ~ 2016 (5 years)

Reactor Site

Site : 50,000 m2
Total Floor Space : 15,000 m2 (Rx : 5,000 m2, Auxiliary : 10,000 m2)

Facility

Classification
Reactor
RI Production & Research Facility
Fission Mo /LEU Target Production Line
Irradiation (Neutron Doping) Facility
Waste Treatment & Storage
15 MW, flexible core with (Max. Flux (<0.625eV) : 34x1014 n/cm2․s )Fuel : U-Mo plate type fuel, LEU
Control rod drive mechanism : Driven from bottom
Mo Target : LEU (UAlX or Atomized U)

Specification

Option III: New Reactor and LEU Target

2017 (New Reactor)
2011

100% supply to domestic users
Mo-99, I-131, Ir-192

RI supply

Main RI supplier

Mo-99 (100% import )I-131 (60% supply)
Ir-192 (90% supply)

RI

Production technologies
I-125, P-32, Lu-177, Ho-166 , Sm-153
Generator : Re-188, Y-90, Tc-99m

Production Capacity(Ci/yr)Mo-99

100,000

I-131> 2,000

I-125

100

Ir-192300,000- 150 -

10

2012-06-20
JAEA-Review 2012-030

Summary• Short supply of 99mTc is not a domestic problem but aninternational problem.
• Seriousness is this problem could be a prolonging problem as
the reactors are getting aged but no replacement available.
• Three time-dependent options are proposed to secure thesupply of 99mTc after reviewing currently available
technologies and the ultimate solution.
– Solvent extraction as the immediate solution– (n,γ)99Mo/99mTc generator by using KAERI adsorbent– New reactor and LEU target• Specially, research on the second option is on-going to meet
30% of domestic demand.

Thank you !

- 151 -

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2012-06-20
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Specialist Meeting on Mo-99 Production by (n,gamma) Method

Introduction of (n, gamma)Mo-99
Production in KAERI
2012. 03
Ul-jae Park
Radioisotope Research Division

Brief introduction of HANARO and its
irradiation facilities for RI production
Target and capsule (or rig) for neutron
irradiation
Production capacity for (n,gamma)Mo-99
Cask, Transportation
Supplying (n,gamma)Mo-99 to Japan2- 153 -

1

2012-06-20
JAEA-Review 2012-030

Brief introduction of reactor and irradiation facilities
Characteristics of HANARO

Design Feature• TypeOpen-tank-in-pool• Power
30MWth• Coolant
Light Water• Reflector
Heavy water• Fuel Materials U3Si, 19.75%
enriched• Absorber
Hafnium• Reactor Building Confinement• Typical flux at port nose
2x1014 nv• 7 horizontal ports & 36 vertical holes• Vertical hole for cold neutron source• Operation Cycle
29 days/13days

Brief introduction of reactor and irradiation facilities
Fields of Utilization
 Neutron Beam Applications

Scattering & Diffraction
Neutron RadiographyEx-core Neutron-irradiation including
BNCT Fuel/Material Irradiation Tests
Capsules
Fuel Test Loop

 RI Production
 Neutron Activation Analysis
 Neutron Transmutation
Doping of Semiconductor- 154 -

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2012-06-20
JAEA-Review 2012-030

Brief introduction of reactor and irradiation facilities

Vertical Holes

Status of Irradiation Facilities

CT : Capsule Irradiation Facility
IR : Fuel Test Loop, RI Production
OR : Capsule Irradiation & RI Production
IP : RI Production
HTS : Hydraulic Transfer System
for RI Production
PTS : Pneumatic Transfer System
for Neutron activation Analysis
NTD : Neutron Transmutation
Doping of Silicon

Brief introduction of reactor and irradiation facilities

Vertical Holes

Usage of irradiation holes for RI production
RegionInner-coreOuter-core

Reflector

SymbolNo. of

Holes forholes

RI Prod.

RI

Thermal neutron flux
(n/cm2 sec)

CT

1

IR

2

1

Ir-192

3.0～4.0×1014

4.0×10142

Ir-192

2.2～3.0×1014

OR

4

HTS

1

IP

17

4

I-131(2),Lu-177,Mo-99, R. RI

2.5×1013～1.5×1014

LH

1

1

I-131

8.5×1013

8.0×1013- 155 -

3

2012-06-20
JAEA-Review 2012-030

Target and capsule (or rig) for neutron irradiation
Irradiation Target for (n,γ)Mo-99 Production

Natural MoO3

Enriched MoO3

99Mo Enrichment

24.4%

98.42%

Purity99.5%

99.982%

Shape of MoO3

PowderPowder1.45

1.45

Apparent Density

MoO3
Inner capsule
Outer capsule

Target and capsule (or rig) for neutron irradiation
Capsule and Rig

CapsuleRig

Double Capsule
Size (mm)- Inner : ø31x85

ø59x1040- Outer : ø38x130
Material
Size of specimen area (cm3)
Max. MoO3/
capsule (g)

Al

Al

45

-

65

IP Rig- 156 -

42012-06-20
JAEA-Review 2012-030

Target and capsule (or rig) for neutron irradiation
Production capacity of Mo-99


7 days Irradiation
99Mo Activity, Ci/target (60g MoO )3

IR-2
TargetNaturalMo
98%
Enriche
d Mo

OR-3

IP-15

EOB

4d

6d

EOB

4d

6d

EOB

4d

6d

150

55

33

81

30

18

38

14

8

413

150

91

261

95

58

135

49

30

Results of Monte Carlo Simulation

Cask, Transportation
Container - Type A
Model

Size(mm)

Weight (kg)

ø50x86xt8(Pb)

0.87

SM-C2

ø66x102xt15(Pb)

2.13

SM-C3

ø85x124xt23.5(Pb)4.7

480 mCi

45.7(40)

40 Ci

SM-C1
KRI

KRI-ASM -01

KRI-ASM -01

ø298x330(over pack)Ø155x189(Pb shield)

RemarkKRI-SM- 157 -

5

2012-06-20
JAEA-Review 2012-030

Supplying (n,)Mo-99 to Japan Situation

 Only one reactor is available in Korea and it is required to have areserved space for the readiness of domestic supply of Tc-99m.
 The irradiation positions are always limited for routine
Production.
 There were only few containers currently available.

 Possibility
The routine supply of Mo-99 to Japan is not realistic.
 KAERI would be pleased to supply small quantities of Mo-99 forresearch purpose for the limited time period.

Thank you !

- 158 -

6

2012-06-20
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Brief introduction of reactor and irradiation facilities
Pools in the Reactor Building
Spent Fuel Storage Pool

Service Pool

Reactor Pool- 159 -

7

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6/20/2012
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HANARO Complex

2- 161 -

1

6/20/2012
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HANARO, the neutron source

High-flux Advanced Neutron Application ReactOr

Reactor Pools

Spent Fuel
Storage Pool

Service Pool

ReactorPool

4- 162 -

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6/20/2012
JAEA-Review 2012-030

Reactor Structure and Design Features

Design Features
Type
Open-tank-in-poolPower
30 MWthCoolant
Light Water
Reflector
Heavy water
Fuel Materials
U3Si, 19.75%
enriched
AbsorberHafnium
Reactor Building Confinement
Max Thermal Flux 4.5x1014 n/cm2s
Max fast Flux
2.0x1014 n/cm2s
Typical flux at port nose
2x1014 n/cm2s
7 horizontal ports & 36 vertical holes
Vertical hole for cold neutron source
Operation Cycle
28 days5

Vertical Holes

Status of Experimental Facilities
IR,CT: Capsule Irradiation Facility
OR : Capsule Irradiation & RI Production
IP: RI Production
HTS : Hydraulic Transfer System(LH)
for RI Production
PTS : Pneumatic Transfer System
for Neutron activation Analysis
NTD : Neutron Transmutation
Doping of Silicon
IR: Fuel Test Loop
CNRF : Cold Neutron Research Facility

Horizontal Tubes
ST2 : High Resolution Powder Diffractometer,
Four Circle Diffractometer
NR : Neutron Radiography Facility
CN : Small Angle Neutron Spectrometer
IR : Ex-core Neutron-irradiation Facility for BNCT & DNR
ST1 : Polarized Neutron Spectrometer, PGAA, RSI
ST3 : Vertical ReflectometerST3 : Horizontal ReflectometerST3 : High Intensity Powder DiffractometerST4 : Triple Axis Spectrometer6- 163 -

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6/20/2012
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Neutron flux of Irradiation Holes

7

Reactor Operation Days8

- 164 -

4

6/20/2012
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Capsules and Rigs for RI Production

Capsules for IP & HTS

Rigs for IR & OR

9

9

RI Production Facility

 Bank1
 4 concrete cells (60Co, 92Ir)

 Bank2


11 lead cells (32P, 99mTc, 51Cr, 192Ir, etc.)

BANK1

BANK2

BANK3

BANK4

 Bank3


6 lead cells (131I, 166Ho, 125I)

 Bank4

4 lead cells (99Mo/99mTc generator)

 Clean Room

Clean room

KGMP for quality products

1010- 165 -

5

6/20/2012
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RI Production
RIs produced in HANARO
Type

RI

Non-sealed Source

Sealed Source

Uses

NuclideForm

Tc-99mNaTcO4

Medical diagnosis

Mo-99

MoO3

Parent nuclide for Tc-99m

I-131

NaI, Capsule

Medical diagnosis and treatment

Ho-166Ho(NO3)3

Medical treatment

Dy-165Dy(NO3)5

Radiation synovectomy

Au-198

Colloid

Liver diagnosis

Cr-51

Na2CrO4

Bio/Medical application

P-32

Na2HPO4

Agricultural application

H3PO4

Genetic study

CylinderNDT(Non-Destructive Test)

Ir-192
Co-60

Seed, pin

Medical treatment

CylinderIndustrial use, NDT

11

11

Utilization of the HANARO

User Community
Industry
University
InstituteInternational Cooperation

Platform for Nuclear
Technology Research

Cold Neutron Beam
NeutronRadiography

National Demand Basic R&D
 Competitivenessin Technology

Neutron
Irradiation
Research

BeamResearch

Reactor Operation
Experimental Facility
Operation

RI
Production&
Utilization

Operation &Development

Fuel Test Loop
Material Irradiation
Neutron Activation
Analysis
NTD

Public Welfare
Industrial
Application

HANARO SAFETY
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Irradiation Facilities in the HANARO

13

Material/Fuel Irradiation Using Capsule
Reactor Materials Tests
Reactor Vessel Materials
Reactor Pressure Tube Material: Zr-2.5Nb
Structural Materials
Fuel Materials Tests
HANARO Fuel
U-Mo Fuel
Advanced PWR Fuel
DUPIC Fuel
U-Zr Alloy
Fundamental Researches
SPND
Semiconductor, Magnetic Materials
Neutron Dosimetry
Zr-1Nb-1Sn-X Alloy, Zircaloy-4
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Instrumented Capsule for Material Irradiation Test
CharpyTube

5th stage

Design Characteristics

R-CT

Hardness/경도/인장Tensile Total Length : ∼6m ( 60mm D x 870mm H)
 Available Space(Max.): 40mm D × 600mm L
 5 Stages Independent Temp. Control
 Max. Temp. Control : Up to 500℃
 He Atmosphere : 1 atm∼3×10-3 torr (He) Instrumentation : 14 T/Cs, 5 Micro-Heaters, 4 /Ms
 Available Specimen : Tensile, Charpy, R-CT, SP,
Tube, hardness, PCVN, MBE, TEM, etc.
 Related Facilities: Temperature Control System, Supporting System, Cutter, Cask, etc

Applications
 Material Tests- Reactor Pressure Vessel- Reactor Core Materials- CANDU Pressure Tube Materials
 Safety and Integrity-Related Tests
 Study on the extension of reactor lifetime Industry Application Material Tests
 Fundamental Research
15

Instrumented Capsule for Nuclear Fuel Test
Design Characteristics
 Total Length : 5,000 mm
 Diameter of Outer Tube : 56 mm
 Length of Outer Tube : 730 mm
 3 Mini-Elements Fuel Rod
 Control of Irradiating Environment Use of Mixed Gas (He/Ne)

Applications
 Fuel Pellet Irradiation Test
- Advanced PWR Fuel
 Fuel Design Data Production- Center/Surface Temperature of Fuel Pellet- Internal Pressure of Test Fuel Rod- Deformation of Fuel Pellet
 Fundamental Research16- 168 -

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Creep/Fatigue Test Capsule
Design Characteristics
 Total Length : ~6,000 mm
 Diameter of Outer Tube : 60 mmD x 997 mmL Irradiation specimen : 1, 2 & 4
 Instrumentation : 8 TCs, 2 heaters, 1 LVDT Creep & fatigue tests temperature : Max. 600 ℃
 Irradiation condition : 1 atm ~ 30 torr (He)Applications
 Examination of nuclear materials throughstudy of creep and fatigue behaviors
 Study on the extension of reactor lifetime
 Study for Fundamental Research

17

Fuel Test Loop Facility
Commissioned at Oct. 2009
Applications
Integral Fuel Irradiation Tests
Fuel Qualification Tests
High Burn-up Fuel Tests
Water Chemistry and Corrosion Tests
Non-fissile Tests of Pressure Tube Material

In-Pile Section► Design Pressure : 17.5 MPa► Design Temperature : 350 ℃

Out-Pile System18- 169 -

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Neutron Activation Analysis (NAA)
Characteristics
Multi-Element Analysis and Neutron Irradiation
Pneumatic Transfer System (3 Irradiation Holes)
High Thermal Neutron Flux(1013 n/cm2s), Cadmium Ratio( 200)
KOLAS Accredited Laboratory since 2001
Gamma Ray Counting System
PTS Layout
Hood 1
DiverterHood 2

Droppedout
PTS #1- Ld & Re

PTS #2
Receiver1

Timer
Hood 3

GM
Tube

2
PTS #2
Receiver 2

1
A

PTS#1
ControllersCounter
end

DiverterCounterend withwindow

PTS#2
Controller

PTS #2 Controller

SafetyTank

Irradiation Rabbits

Pool
HVAC

N2 Tank

PTS #1
Receiver 2
Delay
Stacker

NAA 1 : 772.38 mm fromthe Core
NAA 2 : 625.28 mm
NAA 3 : 471.53 mm

1 2 3

Reacter Pool
Auto
LoaderShieldingRoom

RX Hall

Hand
LoaderSafetyTank

N2 Tank

PTS Room

Counting
Room

PTS #1 Controller& Receiver

19

Neutron Transmutation Doping (NTD)
Production of high quality Si Semiconductor
Services using NTD1 & NTD2 holes
Irradiation of 5“, 6" and 8” ingots
High Uniformity & Accuracy
In Commercial Service from 2002

neutronFz-Silicon

30Si

Neutron
Irradiation

31P

Doped
Silicon

Wafer

IGCT

GTO

IGBT

Rectifier Diode20- 170 -

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IMEF and PIEF

21

21

Conclusion
 RR will continue its important roles ;to Enrich Our Knowledge,
to Enhance Our Life Quality and
to Develop National Economy !
 International cooperation is very effective way todevelop and to advance technology in a narrow RR
society.

22

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STRATEGIES FORMOLYBDENUM-99 SUPPLYSECURITY IN MALAYSIA

Rehir Dahalan
Malaysian Nuclear Agency
Bangi, 43000 Kajang
Selangor, MALAYSIA

Nuclear Medicine Centers- 173 -

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Neutron activated Mo-99
Initial work in 1986
Irradiation position Central Thimble of PUSPATITRIGA Mk II Research Reactor (wet channel)
Neutron flux: 1.0 X 1013 neutrons cm-1s-1
Target: natural abundance Molydenum Trioxide(Merck)Total weight: 90 gms

Target capsules: Al 6061 (TIG welding)Total weight of MoO3: 80 gms
Irradiation time: continuous 75 hours
Cooling time: 18 hours
Yield: 1700 mCi Mo-99
 98Mo abundance : 24.13%),• σ (Thermal): 0.13b- 174 -

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Molybdenum trioxide – inexpensive,
easily available, with provenradiation and thermal stabilityMo% inisotope natural
Mo
Mo-92 14.84

Activation Halfproduct
life
Mo-93

6.9 h

Mo-98

24.13

Mo-99Tc99m

66h-6h

Mo100

9.63

Mo-10114.6m

MEK Tc-99m
Fabricated stainless steel plant for Methyl Ethyl
Ketone extraction of Tc-99m with pneumatic
valves and vacuum line for liquid transfers
Dissolved in conc. NaOH and stirring achieved
with compressed air
Allowing 10 minutes for separation of aqueousand organic phases
Draining of aqueous phase and transfer of Tc99m in MEK in evaporation chamber.

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Allowing 15 minutes for cooling then addition ofsterile normal saline
Filtering through 0.22 micron sterile filter into
sterile evacuated vialReady for dispensing and dispatch daily

MEK Extraction of Tc-99m
Radioactive MoO3
Powder
Dissolution
with freshly
prepared NaOH

Filtrationvia
sterile0.22 micron
filterExtractionof Tc-99m
with MEK

Separationof Tc-99m
MEK

QC testing,External
packaging,
HP certification

Evaporation
Of MEK

Additionof normal
salineDispatch

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Observations
License for manufacture obtained from NPCB in
1989
Yield of Tc-99m low and sufficient for only 1-2
nuclear medicine center (HKL & UMMC)
Reactor requires sufficient operator forcontinuous 75 hours operation
Hospital requires Tc-99m at 8 am daily,
therefore production began before 5 am andpersonnel must arrive much earlier
Operation ceased in middle of 1989

Zirconium Molybdate – 99Mo (Zr99Mo)
Gel generator
Na299MoO4 + ZrOCl2.8H2O

ZrO99MoO4.xH2O +2NaCl

SUITABLE FOR PORTABLE/TRANSPORTABLE GENERATOR Based on conversion of (n,γ)99Mo solution into an
insoluble zirconium molybdate -99Mo gel powder A dual column system consisting ofi) A primary Zr99Mo gel column andii) A secondary acidic alumina purification
column Simple separation of 99mTc with normal saline
provides ready to use Na99mTcO4- 177 -

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Properties of gel generator – elutionof 99mTc
Zr99Mo gel Amorphous powder with open pores structure
containing hydroxo and aquo surface groups boundto the metal Free diffusion of 99mTc occurs from Zr99Mo gels withnormal saline / water Presence of interlayer water molecules is essential for free diffusion
 Gel generator performance- directly related to thegel structure which in turn depends on molar ratio of
Zr : Mo, Concentration, pH, drying temperature etc
 Modification of molybdate pH---different polymolybdates- favours
random structure---increases 99mTc elution efficiency

Preparation of zirconium molybdate
generator
Radioactive Na2MoO4
Solution

Mixing with
Zr solution(GEL)

VacuumFiltration(CAKE)

Tansferintocolumns,
sealing,
bed washing

Placement inGenerator
assembly

Drying of
gel cake(LUMPS)

Dispersion(GRANULES)QC testing,External
packaging,
HP certification

Drying(FREE
FLOWINGGRANULES)

Dispatch12- 178 -

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Requirements for production of gel
generators
Easy access to Mo99 of adequate specificactivity, say 0.4 – 1 Ci/g
Chemical process amenable to automation
for zirconium molybdate preparation
Shielded, well-equipped facility
Suitable generator components andassembly

Challenges in chemical processingCompletion of all steps in a single day
operation to minimize 99Mo decay
Filtration and drying are ‘rate-determining’stepsDrying conditions critical (temp below 1200C)
to prevent structural damage to gelpH adjustment of gel is a critical parameter tobe maintainedReproducibility of final Zr99Mo gel product14- 179 -

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Fission Mo-99
Imported fission Mo-99 (1991- current)
 ANSTO BATAN South Africa

5 – 15 Ci/week- 180 -

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Observations
Logistic problem associated with carrier(airlines)
Intermittent unavailability of fission Mo-99(technical difficulties from manufacturer)Difficulties in sourcing from other manufacturers
maily because quantity utilized is relative small- 184 -

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Other Tc-99m Generator Imports
Amersham, UK
CIS BIO, FrancePolatom, Poland

Possibility of local fission Mo-99
production from LEU
Uranium targets
Reactor flux
Handling of highly radioactive targets fromreactor to hot-cells
Hot-cells capable of containing the fission Mo-99
process
Fission Mo-99 processing technology
Management of fission by products- 185 -

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Small Scale Mo-99 Production from Irradiated Al-clad. UAlx-targets with LEU

Target: 20g Al, 15 g LEU as UAlx
371 ml4 m NaNO3, 3M NaOH

Dissolving process
Filtration +acidification

8 weeks
storage
Xe, NH3

Liq.N2-Trap

Filter cake
Iodine gas
trap

200 ml 1 M HNO3
700 ml H2O
100 ml 0.01 M NH4OH

Exhaust system

Al2O3
column process

100 ml 1 M NH4OH
50 ml 0.1 M NH4OH
50 ml H2O

Filtrate & wash
1220 ml HLW

DOWEX-1
column process

25 ml 1 M NH4HCO3
Evaporation
1 ml conc. HNO3
1000 oCHigh temperature
treatment

20 ml 1 M NaOH

Dissolving residue

Filtrate & wash
200 ml LLW

Liquid Waste treatment:

Total liquid waste volume ~1500 mL
U-free, after 8 weeks evaporation and solidification

Final product: [99Mo] Na2MoO4Alkaline Process – Product & Waste
Flowsheet
IrradiatedTargets

Hot Cell
Processing
Facility

Gaseous Waste
holding/filtration

Mo-99 Product

LLSW
General waste

ILSW
Uranium Filters

ILSW CellILLW
I-LLLW
Waste
LLLW
Liquid waste
Liquid waste
Columns
General andfrom dissolution
from
Decontaminatio
purificationn Waste- 186 -

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Alkaline Process – Estimated Distribution
for Selected Radionuclides
NuclideBa-137m
Ba-140
Cd-115
Ce-141
Ce-144
Cs-136
Cs-137
I-132
La-140
Nb-95
Pm-147
Pm-148
Rh-106
Ru-103
Ru-105
Ru-106
Sr-89
Sr-90
Te-127
Te-132
Y-91
Y-91m
Zr-95

U Ppte

ILLW

H
H
H
H
H

VL
VL
L

H
L
H
H
M
M
M
M
H
H
M
M
H
H
L

H
H
L
L
M
L
L
L
L
L
L
VL
VL
M
M
L
L
M

H = high
M = medium
L = low
VL = Very low
blank =~0

Uranium Filters
Accumulation of U precipitate within filters in cell.
Package filters into primary can.
Raising primary can into secondary can in flask.
Transport to Waste Operations facilities.
Transfer into tertiary stainless steel can for metal
O-ring + weld sealing.
Transfer into other flask and transfer to storage
building for placement in former spent fuelstorage holes.

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Estimated Dominant Activity in U
PrecipitateFirst two weeks: Ba-140, Ce-141, I-132,La140, Rh-103m, Ru-103, Sr-89, Te-132, Y-91,
Zr-95.
At 10 – 12 weeks: Ce-141, Sr-89, Y-91.
At 20 - 22 weeks (at a 10-fold lower level than categories
above): Ce-141, Ce-144, Nb-95, Pr-144, Rh103m, Ru-103, Ru-106, Sr-89, Y-91, Zr-95.

Intermediate Level Liquid Waste Accumulation in in-cell storage tanks.
 Transfer and accumulate in adjacent storage tanks.
 Future transfer to new facility for synroc processing
to dry – calcine – can – hot isostatic pressing (HIP).
 Storage as low volume immobilised ILSW in existing
ILSW storage facility.
 Activity dominated by Nb-95, Zr-95, Cs-137 at 1
year decay.
 Dominated by Cs-137 after significant decay (~> 2years).

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Lower Level Liquid Waste
Accumulation in in-cell storage tanks.
Transfer and accumulate in adjacent storage
tanks for further decay.
Future transfer to new facility for drying.
Storage as consolidated solidified waste.
Activity dominated by Ru-106/Rh-106 in longer
term.

External Liquid Storage Tanks
External liquid tanks to store and decay ILLW
and LoLLW.
Direct shielded connection from in cell storage
tanks.
Transfer facilities to new treatment facility for
ILLW.
Transfer facilities to flask for LoLLW.
Holding and decay strategy (allow for puredecay period).

- 189 -

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Gaseous Waste
Accumulation in in-cell evacuated tanks.
Released into cell ventilation system on rotationbasis after holding for decay of noble gases.
HEPA filtration.
Carbon adsorption filtration.

Observations
Upgrade of reactor to neutron flux approximate
1014 neutrons cm-1 s-1
Procure established uranium targets
Procure established fission Mo-99 processing
technology
Upgrade current waste management facilities
Market fission Mo-99 beyond Malaysia- 190 -

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Cyclotron production Tc-99m
 Model – GE PETtrace
 Open-shielded type
 The 16.5 MeV (proton ) and 8.5
MeV ( deutron) cyclotron
 1.5 Tesla ( magnetic field)
 built in 6 Targets port (equipped with 3 targets) Dual beam

Direct cyclotron production of
Tc-99m 100Mo (p,2n) 99mTc
Limited to nuclear medicine centers nearby thecyclotron- 191 -

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References Saraswathy Padmanabhan, Production of Tc-99m Generators
Using Neutron Activation Mo-99
 Gerd J Beyer, German Experiences in Fission-based Mo-99 And
Recent revitalization by GSG
 G J Parsons, L Tan, W Townsend, Waste Handling and Processing
from Production of Mo-99
 Brigitte Gue´rin et al, J. NUCLEAR MEDICINE • Vol. 51 • No. 4 •April 2010 . Cyclotron Production of 99mTc: An Approach to theMedical Isotope Crisis.

- 192 -

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PUSPATI
TRIGA
ReactorPUSPATI TRIGA Reactor

PUSPATI TRIGA
Reactor (RTP) is a
1MW pool type TRIGAMkIIFirst critical: 28 June1982
2

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PUSPATI TRIGA Reactor - Specification
Type:
Critical:
Fuel:No. Fuel :1 MW TRIGA MkII
28 June 1982
19.9% enriched U-ZrH
112 (Core #14)
(10@20%wt/, 16@12%wt, 86@8.5%wt)
Coolant:
Light Water
Moderator:
Light Water
Reflector:
GraphiteControl Rod:
Boron Carbide - 3 FFCR & 1 AFCR
Neutron Source: Americium Beryllium
Cooling System: Primary (natural convection), Secondary
Irradiation Facility: PTS, Central Thimble, Dry Tube, 4 Beamports,
Rotary Rack,Thermal Column, NuR2, SANS,
Maximum Thermal Flux: 1.777 ± 0.031x 1013 n/cm2/s
Total Flux: 2.061 ± 0.081 x 1013 n/cm2/s

PUSPATI TRIGA Reactor – Vertical Cross-section

4- 194 -

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Fuel Element

Reactor Core

Fuel Element

Central Thimble

Control rod- 195 -

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ALUMINIUM CAN – WELDED /WET TYPE- 196 -

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99Mo/99mTc-Generator and 99mTc-Radiopharmaceutical Service

in Thailand

Mr. Jatupol SANGSURIYANRadioisotope Center
Thailand Institute of Nuclear Technology, Bangkok THAILAND

New TINT center ̴ 70 km. North –East Bangkok- 197 -

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Our Institution• Thailand Institute of Nuclear Technology,TINT, is a
public organization recently separated from the Office ofAtoms for Peace,OAP, was by-law, established by Royal
Decree. Establishing of Thailand Institute of Nuclear
Technology is effective in April 21,2006.

Radioisotope center, TINT

• one of five service centers of TINT
• responsible for research, development andservice of radioisotopes andradiopharmaceuticals for medical, agricultural,industrial and research applications.

- 198 -

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Radioisotope center, TINT
Have the policy to provide high quality products withreasonable price
Main product – medical isotopes
Supply to nuclear medicine centers/hospitals- 11 centers in Bangkok city
- 9 centers in other provinces

List of products routinely service
1. I-131 (solution, Diagnostic /Therapeutic capsule)
2. I-131 labeled compounds(Hippuran,MIBG)
3. Sm-153 EDTMP
4. Phosphorus-32
5. Kits for Tc-99m
radiopharmaceuticals- 199 -

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New product and service• Centralized radiopharmacy service since 2010
-Deliver 99mTc-radiopharmaceutical finishedproducts as Unit Dose to nuclear medicine
centerProduce from
Tc-kit - local productTc-generator - imported

New products and service
Already service routinely
99mTc-Hynic TOC

188Re-HEDP
188W/188Re-generator(Polatom)- 200 -

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New product and service
Clinical study
I-131 Rituximab
68Ga -NOTA-RGD
68Ga -DOTA-TATE
68Ga –Citrate

68Ge/68Ga-generator (E&Z)

99mTc-radiopharmaceuticals• No production Tc-99m for more than 10 year,since1998• But during 1984 -1997 we had produced Tc-99m anddelivered to 8-9 nuclear medicine centers in Bangkok
city- 201 -

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99mTc-radiopharmaceuticals

Tc-99m production during 1984 -1997

Production technology - MEK extraction process
Target- 100g MoO3 (natural)
Irradiation facility- TRR-1/M1 research reactor
n flux 2.38 X 1013 n.cm-2.s-1
production yields

1-1.5 Ci/day(from ̴ 3 Ci after 48 hr. irradiation)

99mTc-radiopharmaceuticals

Hot cell for MEK extraction process- 202 -

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99mTc-radiopharmaceuticals

in cell MEK extraction process

99mTc-radiopharmaceuticals

Our research reactor TRR-1/M1
Type - TRIGA Mark III
Max.power 2 MW
Operating power 1.2 MW
46 hr./week(Tuesday-Friday)
Operating 10 month/year- 203 -

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99mTc-radiopharmaceuticals

Our research reactor TRR-1/M1
In-core irradiation facility and n flux ( n.cm-2.s-1 )• Hold dia. Ø 1.5 in.
CT
2.38 X 1013
C8
1.92 X 1013
C12 2.09 X 1013
F3
1.23 X 1013
F29 1.12 X 1013

99mTc-radiopharmaceuticals

Application of research reactor TRR-1/M1
Isotopes production 1998-2012 no Tc-production
Main production I-131 (and some Sm-153 and P-32)
max capacity for I-131 5-6 Ci/w eek ( 1/3 of total demand)•

Hot Cells
for I-131- 204 -

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99mTc-radiopharmaceuticals• Tc-99m radiopharmaceuticals had been developed in
our institute since 1982
• Cold Tc-kit products were developed and produced forroutine supply to nuclear medicine center in Thailand• 70-80% Tc-kit or more than 8,000 kit /y(base on 2010)
supplied by our institute99mTc-radiopharmaceuticals• development of Tc-99m radiopharmaceutical
New production facility to be GMP certified- 205 -

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99mTc-radiopharmaceuticals• development of Tc-99m radiopharmaceutical
New production facility to be GMP certified
Grade Bclean room
1,500 kit/batch

99mTc-radiopharmaceuticals
99mTc-radiopharmaceuticals service• Bangkok city
- Centralized ( 50% ) by 2 center, TINT and GMS- on-site compounding(50%)• Out side Bangkok/ other cities- on-site compounding (100%)- 206 -

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99mTc-radiopharmaceuticals
99Mo/99mTc-generator

Supplier-100% imported by 2 private supplier
Local demand -500-625 Ci as Mo-99 /y (base on 2010)
product 1

product 2

99mTc-radiopharmaceuticals
99Mo/99mTc-generator• Under shortage of world Mo-99 supply since 2007
• TINT is looking for potential technology that can serve stableavailability of Tc-99m radiopharmaceutical for local use• During 2002-2005 we had participated FNCA project on
development of Tc-generator based on PZC gel, but we candeveloped only lab.scale and low activity Tc-generator due tolimitation of existing facilities and quality of PZC -absorbent.

- 207 -

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99mTc-radiopharmaceuticals
99Mo/99mTc-generator• In the same time we are still interested in n,f Mo-99 based
technology for Tc-generator production• A project for production and supply of Tc-99m isotope or
99Mo/99mTc-generator may be initiated in the next fiscal year
• In the next few months project proposal will be submitted
for approval and funding

Thank you- 208 -

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Thai Research Reactor-1/
Modification 1 (TRR-1/M1)Thailand Institute of Nuclear Technology (Public Organization)“To be the excellence center for research, development, and utilizationof nuclear technology for sustainable development of the country"

History of Thai Research Reactor
In 1960, The cabinet ordered theAtomic Energy Commission (AEC)
to arrange the reactor building
construction. The construction took
2 years to complete.

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History of Thai Research Reactor
In April 1962, the prime ministerat the time came for grand opening
ceremony and named the reactor“Thai Research Reactor-1”

History of Thai Research Reactor

In October 1962, scientists,
engineers and reactor staff
loaded nuclear fuel elementsin the reactor core and thereactor reached criticality
for the first time in Thailand.

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History of Thai Research Reactor

In 1977, the reactor core was
replaced by a new TRIGA Mark-III
type. The reactor name was changed
to “Thai Research Reactor-1/
Modification 1” to reflect thismajor modification.

Today’s Thai Research Reactor- 215 -

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Main components of TRR-1/M1
 Reactor core

 Neutron source Fuel elements
 Control rods
 Moderator
 Neutron Detector

 Cooling system
 I&C system
 Shielding

Reactor core of TRR-1/1
CONTROL RODS

NE
UT 6
RO IN
N CH
BE ES
AM
PO

IN-COREIRRADIATION
HOLES

RT

ES
C H AM
IN E
6 NB
RO
UT
NE

RT
PO

UPPER GRID PLATE
218.28 cm

8.89 cm DIA
x 10 holes

122.40 cm

LOWER GRID
PLATE
54.61 cm

SAFETY PLATE- 216 -

42012/6/22
JAEA-Review 2012-030

Current Core Configuration
8600
G14
S
S-2

8605
G12
8631
G11
8618
G10
8615
G9
ND
G8

8635
F11
8590
F10
8632
F9
8573
F8
8558
F7
9219
F6
8595
G6

8623
G15
8570
F12
10024

E9

9223
E8
9864
E7
10041
E6
10023

E5

8612
F5
IIT
G5

8565
G16
8569
F13
9224
E10
10026
D7
10043
D6
10018
D5
10046
D4

9859
E4
8643
F4
8649
G4

8609
G18

8604
G17

8630
F16

8602
F15

8645
F14

9863
E13

9868
E12

9860
E11

REG
10245
D10

10027
D9

9227
TC
D8

10088
C7

9228
TC
C6

10091
C5

8650
B4

8596
B3

TR

CT
A1

8566
B2

C4
10090
C3
10020
D3

10092
C1

10025
D2

9867
E3

SAFE
10097
D1

9225
E2

IIT
F3

9221
E1

8579
F2

8563
G3

8627
G2

8638
F1
S-1

8621
G21

8562
F17

G22

8578
F18

9862
E14
10022
D11

10094
D12

IIT
C8

10093
C9

8568
B5

10089
C10
10049
C11

IIT
C12

10095

FFR

10096

FFR

9866
E23

9865
E24

IIT
F29

8606
F30
8585
G36

8639
G35

8594
G24
8617
F20 8561
F21

9861
E16

10021
D14
10048
D15
SH2
10247
D16
10019
E22

8611
F28
8610
G34

S-3

8637
G26

10042

SH1
10246
D13

E17

10044

E18

10045
E19
10047
E20

9220
E21

Core No. 16
 109 fuel elements loaded
 Fuel elements arranged
in hexagonal
configuration
 2 types of fuel elements


8634
G23

8651
F19

9226
E15

8572
B6

8564
B1

10050
C2

ND
G20

8583
F22

8574
G27

8575
F23

8624
G28

8597
F24

8620
G29

8559
F25

ND
G30

9222
F26
8599
F27 8619
G32
IIT
G33

 8.5 wt%
 20 wt%

8.5 wt-% FUEL ELEMENT

AIR FOLLOWER CONTROL ROD

20 wt-% FUEL ELEMENTISOTOPE IRRADIATION TUBEINSTRUMENTED FUEL ELEMENTNEUTRON DETECTOR

FUEL FOLLOWER CONTROL ROD

NEUTRON SOURCE

Locations of fuel elements
8600
G14
S
S-2

8605
G12
8631
G11
8618
G10
8615
G9
ND
G8

8635
F11
8590
F10
8632
F9
8573
F8
8558
F7
9219
F6
8595
G6

8623
G15
8570
F12
10024

E9

9223
E8
9864
E7
10041
E6
10023

E5

8612
F5
IIT
G5

8565
G16
8569
F13
9224
E10
10026
D7
10043
D6
10018
D5
10046
D4

9859
E4
8643
F4
8649
G4

8604
G17
8645
F14
9860
E11
9227
TC
D8
10091
C5
TR
C4
10090
C3
10020
D3

9867
E3
IIT
F3
8563
G3

8609
G18
8602
F15
9868
E12
10027
D9
9228
TC
C6

8596
B3
8566
B2
10050
C2
10025
D2

9225
E2
8579
F2
8627
G2

8630
F16
9863
E13
REG
10245
D10

10088
C7

8650
B4
CT
A1
8564
B1
10092
C1
SAFE
10097
D1

9221
E1
8638
F1
S-1

ND
G20
8562
F17
9862
E14
10022
D11

IIT
C8
8568
B5
8572
B6

IIT
C12
10096

FFR

9865
E24
8606
F30
8585
G36

8621
G21
8578
F18
9226
E15
10094
D12
10093
C9
10089
C10
10049
C11
10095

FFR

9866
E23
IIT
F29
8639
G35

G22

8651
F19
9861
E16
SH1
10246
D13
10021
D14
10048
D15
SH2
10247
D16
10019
E22

8611
F28
8610
G34

8634
G23

8594
G24
8617
F20 8561
F21

10042

E17

10044

E18

10045
E19
10047
E20

9220
E21

8583
F22
8575
F23
8597
F24
8559
F25

9222
F26
8599
F27 8619
G32
IIT
G33

S-3

8637
G26
8574
G27
8624
G28
8620
G29
ND
G30

8.5 wt-% FUEL ELEMENT

AIR FOLLOWER CONTROL ROD

20 wt-% FUEL ELEMENTISOTOPE IRRADIATION TUBEINSTRUMENTED FUEL ELEMENTNEUTRON DETECTOR

FUEL FOLLOWER CONTROL ROD

NEUTRON SOURCE- 217 -

5

2012/6/22
JAEA-Review 2012-030

Neutron source

60o
TYP

1/16 R

0.62
DIA
45o

0.40

3/32
DIA. THRU Am-Be neutron

1 1/2

3/8
1/8

source ~ 3 Ci
 To initiate fission
reaction

3/4
DIA

0.605 +.000 DIA-.005

10 7/8

7 15/16

1/2
1/2
FLAT
1/2
7/16
1/2 – 20 UNF – 2A

TAP DRILL 29/64 DIA – 9 DEEP

0.605 +.000 DIA
005

Reflector
 Top and Bottom graphite bars
 Thermal column- 218 -

6

2012/6/22
JAEA-Review 2012-030

Irradiation Facilities (2) Out-Core Facilities:
 Rotary Specimen Rack for NAA
 A1, A4, CA2, CA3, TC for NAA
 Wet Tube and Void tank for

Gemstone Irradiation

Shielding Water in pool
 High density concrete varying in thickness(from 1. 5 ft to 4.5 ft )- 219 -

7

2012/6/22
JAEA-Review 2012-030

Cooling systemNaturalconvection

Instrument and Control system- 220 -

8

2012/6/22
JAEA-Review 2012-030

Operation Schedule Nominal Operating Power 1.2 MW
 46 Hours / Week

(Tuesday – Friday) Operation: 10 Months
 Shutdown for yearly maintenance: 2Months

Irradiation Facilities In-Core Facilities:
 Radioisotope Production (I-131, P-32,

Sm-153, and etc.)

 Locations: CT,C8, C12, F3, F29, G5, G33

 Pneumatic Sample Transfer for NAA
 Location: G22- 221 -

9

2012/6/22
JAEA-Review 2012-030

In-core Irradiation Facilities
8600
G14
S
S-2

8605
G12
8631
G11
8618
G10
8615
G9
ND
G8

8635
F11
8590
F10
8632
F9
8573
F8
8558
F7
9219
F6
8595
G6

8623
G15
8570
F12
10024

E9

9223
E8
9864
E7
10041
E6
10023

E5

8612
F5
IIT
G5

8565
G16
8569
F13
9224
E10
10026
D7
10043
D6
10018
D5
10046
D4

9859
E4
8643
F4
8649
G4

8604
G17
8645
F14
9860
E11
9227
TC
D8
10091
C5
TR
C4
10090
C3
10020
D3

9867
E3
IIT
F3
8563
G3

8609
G18

8630
F16

8602
F15

9863
E13

9868
E12

REG
10245
D10

10027
D9

10088
C7

9228
TC
C6

8650
B4

8596
B3

CT
A1

8566
B2

8564
B1

10050
C2

10092
C1

10025
D2

SAFE
10097
D1

9225
E2

9221
E1

8579
F2
8627
G2

8638
F1
S-1

ND
G20
8562
F17
9862
E14
10022
D11

IIT
C8
8568
B5
8572
B6

IIT
C12
10096

FFR

9865
E24
8606
F30
8585
G36

8621
G21
8578
F18
9226
E15
10094
D12
10093
C9
10089
C10
10049
C11
10095

FFR

9866
E23
IIT
F29
8639
G35

G22

8651
F19
9861
E16
SH1
10246
D13
10021
D14
10048
D15
SH2
10247
D16
10019
E22

8611
F28
8610
G34

8634
G23

8594
G24
8617
F20 8561
F21

10042

E17

10044

E18

10045
E19
10047
E20

9220
E21

8583
F22
8575
F23
8597
F24
8559
F25

9222
F26
8599
F27 8619
G32
IIT
G33

S-3

8637
G26
8574
G27
8624
G28
8620
G29
ND
G30

8.5 wt-% FUEL ELEMENT

AIR FOLLOWER CONTROL ROD

20 wt-% FUEL ELEMENTISOTOPE IRRADIATION TUBEINSTRUMENTED FUEL ELEMENTNEUTRON DETECTOR

FUEL FOLLOWER CONTROL ROD

NEUTRON SOURCE- 222 -

10

2012/6/22
JAEA-Review 2012-030

Out-core Irradiation Facilities
THERMALCOLUMN

NE
UT 6
RO IN
N CH
BE ES
AM
PO
RT

VOID TANKNE

8 INCHESNEUTRON BEAM PORTOUT CORE
IRRADIATION
HOLES

T
OR
ES P
CH AM
IN E
6 NB
RO
UT

8 INCHESNEUTRON BEAM PORTTACADMIUM
COVEREDIRRADIATION
HOLES

A4
A1

ROTARY SPECIMEN
RACK

P5

P1
P4

P3

P2

WET
TUBE

Irradiation Facilities (3) Neutron Beam Ports Neutron Radiography Neutron Diffractometry PGNAA- 223 -

11

2012/6/22
JAEA-Review 2012-030

Neutron Radiography

PGNAA- 224 -

12

2012/6/22
JAEA-Review 2012-030

Neutron Diffractometry

Rietveld Refinement of Silicon Powder, data from OAEP ND,using Cu(220)-focusing Monochromator300

counts/300 sec.

200

100

0
I
I I

I

I

I

I

I
I

II

I

I
II

I
I

I

I

I

I
I

II

I

I
II

I

II

I

I

I
I I

I
I II

II

I I

I
I II

I
I II

I
I I

I I

I
I II

I
I II

-100

-200
20

30

40

50

60

70

80

90

100

2 THETA (degrees)

Training and Education
 Reactor operator training Reactor experiment course for university

students (in cooperation with
Chulalongkorn University) Technical tours for knowledge
dissemination to the public- 225 -

13

2012/6/22
JAEA-Review 2012-030

Reactor Core Management
Strategy
 Meet all safety requirements
 Minimize power peaking factor
 Optimize fuel utilization
 Maximize neutron fluxes in irradiation

facilities- 226 -

14

2012/6/22
JAEA-Review 2012-030

Utilization of TRR-1/1
HRD

NeutronActivation
Analysis

Gems
Coloration

Research& Development

Isotopeproduction
Neutronradiography

Technical tours

Products from TRR-1/1
Fission track in Polycarbonate film
Pore size (micrometer)
1.4
1.2
1.0
0.8
0.6
0.4

etching 60 min.
etching 90 min.

0.2
0.0
3

3.5

4

4.5

5

5.5

6

6.5

7

Etching concentration (Normal)

Gems

Isotopes

NDT

Analysis

Personnel

Researches

Knowledge

Reactor Safety awareness- 227 -

15

2012/6/22
JAEA-Review 2012-030

ขอบคุณค่ะ / ครับWelcome to Thai Research Reactor- 228 -

16

JAEA-Review 2012-030

Current Status of Mo-99 Production in
the HFRSpecialist Meeting on
Mo-99 Production
JAEA, Tokyo
10 March 2012
Frits Moet M.Sc.

NRG Introduction
The Dutch centre of expertisefor nuclear technology
Products and services

Research & Development, Qualification
Testing and Consultancy.
Production of (Medical) Radioisotopes andIrradiation Services.

Locations
PettenArnhemNumber of employees
360- 229 -

1

JAEA-Review 2012-030

Unique combination of facilities

Hot cell laboratory

Radioisotope processing lab.

Actinides laboratory

Low Flux Reactor

Radiological
Laboratory

600 FTE

360 FTE

230 FTE

180 FTEHigh Flux Reactor- 230 -

2

JAEA-Review 2012-030

Isotopes produced in the HFR
Mo-99/Tc-99m
Ir-192
Sr-89
Sm-153
I-131
Y-90
Au-198
Re-186
Er-169P-32
I-125
Others include:- Yb-169, Ho-210, Tm-170, Pt-195m, etc

Mo-99 European Supply Chain
Mo-99“bulk”
liquidU-235
targets

Mo-99/Tc-99m
generators
Generator
Production“TargetIrradiation”

Mo-99
facility“Target dissolution & Mo-99
processing”“Loading Mo-99
on the Mo-99/Tc99m column”

HFR

Covidien, Petten (NL)

Covidien, Petten, (NL)

BR2

IRE, Fleurus (B)CIS / IBA Saclay, (F)ReactorOsirisGE Healthcare, (UK)Maria

Europe - top 3

Hospitals

Rez

6- 231 -

3

JAEA-Review 2012-030

Molybdenum Supply Chain

7

Supply network from HFR- 232 -

4

JAEA-Review 2012-030

HFR supply network to Japan

Historical Mo-99 Global Market Share10- 233 -

5

JAEA-Review 2012-030

HFR General Characteristics
•
•
•
•
•
•
•

•

HFR Data
Power
45MW
Cycle Length
28 or 29 days
Number Cycles
10.5 cycles/year
Availability
290 days/year
Maximum Thermal Flux
2.6 x 1018n/m2/s
Maximum Fast Flux
1.8 x 1018n/m2/s
Total Irradiation Channels (60 cm height)- Core17- Poolside10- Others
3- Total30
Of which, 2 channels dedicated to Core Integrity Survey

PSF: Pool Side Facilities

Schematic of theHigh Flux Reactor(HFR) Petten, The
Netherlands
CR

CR

CR

CR

CR

CR

HFR Reactor Core+

Beryllium reflector
Fuel

HB: Beam Tubes

Experimental or isotope
production position
CR Control rod- 234 -

6

JAEA-Review 2012-030

PSF: Pool Side Facilities

Schematic of theHigh Flux Reactor(HFR) Petten, The
Netherlands
CR

CR

CR

CR

CR

CR

HFR Reactor Core+

Beryllium reflector
Fuel

HB: Beam Tubes

Experimental or isotope
production position
CR Control rod

PSF: Pool Side Facilities

Schematic of theHigh Flux Reactor(HFR) Petten, The
Netherlands

Mo
Mo
Mo
Mo
Mo

CR

CR

CR

HFR Reactor Core

Mo
CR

CR

CR
Mo

+

Beryllium reflector
Fuel

HB: Beam Tubes

Experimental or isotope
production position
CR Control rod- 235 -

7

JAEA-Review 2012-030

HFR Mo-99 Production Capacity
•
•
•
•

•

HFR Data - Usage of Irradiation Channels
Maximum Thermal Flux
2.0 x 1018n/m2/s
HEU Targets<5g HEU each
Quantity Targets
6 – 9 Per Channel
Normal Mo-99 HEU Irradiation Channels- Core3- Poolside
4- Others0- Total7
Maximum Mo-99 HEU Irradiation Channels- Core- Poolside- Others- Total

3
8
0
11

HFR Mo-99 Production Capacity
•
•
•
•
•
•
•

HFR Data – Availability of Mo-98 Irradiation Channels
Typical Existing Re-loadable Thermal Facility
Peak Thermal Flux
1.6 x 1018n/m2/s
Number Irradiation Cans
18 positions
Size Irradiation Cans
25mm dia. x 80mm h
Irradiation Can Capacity
30cm3 per Can
Number of Can positions
Open to discussion
Number of Irradiation Channels
Open to discussion•

No existing experience of Mo-98 irradiations••

No existing experience of Mo-98 shipment
Extensive experience in shipping HEU Targets and other highactivity shipments worldwide- 236 -

8

JAEA-Review 2012-030

Thank You
NRG committed toWorldwide Healthcare
and Energy Supply

General Nuclear Medicine Slides- 237 -

9

JAEA-Review 2012-030

Nuclear Medicine

The importance of Medical Isotopes
Medical isotopes are used in more than 20.000
hospitals and clinics worldwide:•40 million diagnostic studies per year
•4 million therapy procedures.

•Diagnostic (short lived isotopes,

Mo-99/Tc-99m>80% of the total use)

•Pallation (pain relief from cancer,
rheumatoid arthritis, etc)

•RITT (RadioIsotope Target Therapy - short

lived isotopes linked to ‘smart’ biomolecules)•Brachytherapy (longer lived isotopes assources and implantable seeds)

19

Diagnostics

Technetium-99m – Workhorse of Nuclear Medicine

•
•
•
•
•
•
•

Daughter of Molybdenum-99 (half life 66 hours)
General use since early seventies
Low patient dose
Short half-life (six hours)>30 diagnostic applications
Useful for range of different organs
Relatively low cost20

- 238 -

10

JAEA-Review 2012-030

Technetium-99m:- Important Applications• Cardiology (heart failure)• Oncology/Bone (cancer diagnoses
and staging)• Urology (kidney function)• Neurology (brain function anddementia)• Infection imaging

21

Therapy

Iridium-192 and Iodine-125

12 M€
9 M€

(Min EZ)

20 M€

Treatment of prostatecancer with Iodine-125

Treatment of lung tumors
with Iridium-192

22- 239 -

11

JAEA-Review 2012-030

NRG Response to Supply Crisis Slides

European Reactor Position Paper•

In 2008, world Mo-99/Tc-99m supply shortages werealready occuring, the OECD had already started to call thefirst international security of supply meeting.

•In 2009 with the NRU (Canada) out of operation and the
HFR scheduled for a 6 month essential repair, it was clearthat the supply problems would continue.

•

The various Agencies of the European Commission had
also started various actions to investigate the problem.

•

The main Reactors in Europe involved in producing
Medical Isotopes agreed that it was essential that they had
a common position.

•

The result was a common“Position Paper by CEA (F), NRG (NL),SCK-CEN (B) and TUM (D)”.

- 240 -

12

JAEA-Review 2012-030

A Summary (1)•The historical strength and success of the European Mo99/Tc-99m supply model has been based upon an
integrated network of Reactors and Processors.

•

During the supply crisis European Reactors andProcessors very significantly increased production to
maximum sustainable levels.

•

Reactor operators like NRG actively helped other reactors(e.g. Maria, Poland) to be added into the European supply
network.

A Summary (2)•The EC DG Energy took many of the key points from thecommon reactor “position paper” in it’s advice to theEuropean Parliament which was accepted.

•

The actions to be taken have been debated in the EC
Atomic Questions Group.

•

Europe recognises that future security of supply isessential and must be based on an integrated network of
Reactors and Processors, with both European andInternational partners.

- 241 -

13

JAEA-Review 2012-030

THE FUTURE•

The European response will be to strengthen the existing
integrated supply network.

•

The European Commission in cooperation with the
Association of Isotope Producers and Equipment
Suppliers is working to create an “Observatory”
mechanism to oversee key supply issues.

•

The NRG response is to build the new Reactor – PALLAS.

HFR BPL Repair Slides- 242 -

14

JAEA-Review 2012-030

NRG/HFR response to the supply crisis•

When the HFR was operational:- Mo-99 production capacity was increased by 50%.
- Facility utilisation was increased.
- Total Mo-99 production was increased by around100% to produce about 70% of world demand.
- Non medical work was sacrificed•

The HFR Repair was very carefully planned- Then successfully executed- as shown in the following slides.

Summary• No safety incidents; low dose rates• Return to services (45MW)
9 September 2010; 15:50 hrs- 243 -

15

JAEA-Review 2012-030

Repair project- Remove components and systems- Radiation shielding- Drain pool
- Remove concrete- Drilling- Cutting by hand
- Repair / (replace) reducers- Restore concrete- Refill pool
- In Service Inspection- Remove shielding- Install components and systems- Re-commissioning- Core reload & system checks- Start-up and return to services

Remove concrete – core drilling- 244 -

16

JAEA-Review 2012-030

Diamond drill for core drilling

Reducer – cleaning process- 245 -

17

JAEA-Review 2012-030

Reducer – with deformations removed

HFR returned to service- 246 -

18

JAEA-Review 2012-030

We are happy to say that the
HFR resumed vital role forWorldwide Healthcare
and Energy Supply- 247 -

19

This is a blank page.

表１．SI 基本単位
表１．SI 基本単位
SI 基本単位
SI 基本単位
基本量
基本量
名称
記号
表１．SI 基本単位
名称
記号
長
さ メSI
ー基本単位
トル m
長
さ
メ
ー
ト
ル
m
基本量
質
量 キログラムkg
名称
記号
質
量 キログラム kg
時
間
秒
s
長
さメ ー トル m
時
間ア ン 秒ペ ア kgA s
質電量流キログラム
電流 ア ンビ
ペンア KA
時熱力学温度間 ケル秒s
熱力学温度 ケ ル ビ ン mol
K
電物 質 流量アモン ペ アル A
物
質
量
モ
ル
mol
光
度ケカルンビデンラ Kcd
熱力学温度
物光 質
光

量度
モカ ン デル ラmolcd
度 カ ン デ ラ cd

国際単位系（SI）
国際単位系（SI）
国際単位系（SI）
表２．基本単位を用いて表されるSI組立単位の例

表２．基本単位を用いて表されるSI組立単位の例
SI 基本単位
組立量
SI 基本単位
名称
記号
組立量
記号
面 表２．基本単位を用いて表されるSI組立単位の例
積 平方メートル 名称
m2
2
SI
基本単位
面
積
平方メートル
体 組立量
積 立法メートル
m3 m3
名称
記号
積 立法メートルm
速 体さ ， 速 度 メートル毎秒m/s
2
面加 速 さ速 ， 速
積度平方メートル
2 m/sm
度 メートル毎秒
メートル毎秒毎秒m/s
3
体波 加-1 m/s2mm
速 積数立法メートル
度 メートル毎秒毎秒
毎メートル
-1
速 波
さ ， 速 度 メートル毎秒m/s
数 毎メートル密 度 ， 質 量 密 度 キログラム毎立方メートルkg/m3 m加 密 度速3m/s2 kg/m
2，質
量度度
密メートル毎秒毎秒
度 キログラム毎立方メートルキログラム毎平方メートル
面
積
密
-1
kg/m
波
数 毎メートルm
3 3 kg/m2
面
積
密 積キログラム毎立方メートル
度 キログラム毎平方メートル
立方メートル毎キログラム
比
体
m
/kg
密度，質量密度kg/m
積 立方メートル毎キログラムA/m22m3/kg
密 度度キログラム毎平方メートル
アンペア毎平方メートル
面電 比積流 密体kg/m2
磁 電界 流の 強密 さ アンペア毎メートル
A/m
度
アンペア毎平方メートル
比体(a)積 立方メートル毎キログラム
m3/kg A/m
3
量磁
濃度
界 ，
の 濃強度 モル毎立方メートル
さ アンペア毎メートル mol/m
2 A/m
電
流
密(a) 度 アンペア毎平方メートル
A/m
3
質
量
濃
度 キログラム毎立法メートル
kg/mmol/m3
度 モル毎立方メートル A/m
磁 量
界 濃の度 強 ，さ濃アンペア毎メートル
3
輝 質 (a)量度 カンデラ毎平方メートル
cd/m2
度 キログラム毎立法メートル量 濃度， 濃濃度(b)モル毎立方メートル
mol/m3kg/m
（数字の）
１
13
屈
折
率
度 カンデラ毎平方メートル
cd/m2
質 輝量
濃
度(b)キログラム毎立法メートル
kg/m
（数字の）
１ １
1
(b)
比 透 磁
（数字の）
折 率度
率 カンデラ毎平方メートル輝 屈
cd/m2 1(b)(b)concentration）は臨床化学の分野では物質濃度
（数字の）
１
1
比 折透 磁（a）量濃度（amount
（数字の）
１
1
屈
率 率
(b)
concentration）ともよばれる。
（数字の） １1
比 （substance
透 磁 率

101

表３．固有の名称と記号で表されるSI組立単位
SI 組立単位
表３．固有の名称と記号で表されるSI組立単位
他のSI単位による SI基本単位による名称
記号
表３．固有の名称と記号で表されるSI組立単位
組立単位
SI表し方
表し方
組立単位
（ｂ）
組立量
平
面
角 ラジアン(ｂ)rad SI 他のSI単位による
m/mSI基本単位による
1
名称 (ｂ)
記号組立量（ｂ）
他のSI単位による(c)
2/ 2
SI基本単位による立体
角 ステラジアン
表し方
表し方
sr
1
m
m
名称
記号
表し方（ｂ）
表し方
（ｄ）
周
波
数角ヘルツ
Hzrad
平
面
s-1 m/mラジアン(ｂ)1（ｂ）
平面
角 ラジアン(ｂ)rad (c)1（ｂ） m/m
-2 2
(ｂ)
2/
力
Nsr
立
体
角ニュートンm kg
ステラジアン
1
（ｂ）
(ｂ)
2ms m
立
体
角 ステラジアン
sr(c)
1
m2/m
（ｄ）
周
圧
力
応
力数パスカル,波PaHz
s-1 s-2
ヘルツ
N/m2
m-1 kg
（ｄ）
周
波
数 ヘルツ
Hz
s-1 2
-2
力 ネ ル ギ ー , 仕 事 , 熱 量 ジュール
ニュートン
N
エ
J
N
m
m
kg
s-2
kg
m
力
ニュートン
N
m kg
s-2-1s-3
2
-2
2
圧 事力
応
力
パスカル
率力， 工
率
，
放
射
束
ワット
,
Pa
W
J/s
2N/m
-1 kg
-2 kg s
m
s
m
圧仕
応
力
パスカル
,
Pa
N/m
m kg s2-2
電
荷 ギ, ー ,電仕 事
気, 熱量量クーロンC
s A m-2
ジュール
kgs
エエネネルルギー, 仕事, 熱
量 ジュールJ J
N m N m m2 kg
s -3 -1
2s A -3
電 位 差 （ 電工
圧 ），, 起 電射力束ボルト
V
W/A
m2 kg
ワット
kg s
m-3
仕仕事事率率，，工 率率， 放放
射 束 ワットW W
J/s J/s
m2 kg-2 s -1 4 2
静電
容量量ファラド
F
C/V
クーロン
sAs A
電電 荷荷 , , 電 電 気 気 量 クーロン
C C
sm
A 2 kg-3 -2-3 -1
2
気抵
V/AW/A
2 kg
-1s A
電位位
, 起
力オーム
ボルト
ms-3skg
電電
差差
（（
電電
圧圧
））
, 起
電電
力抗ボルト
VΩ V
W/A
kg
AA
mm
-2
-1 3 -12 4 2
A/VC/V-2 kg-1-2 4s A
静 ン 電ダ
電 ク 容タ容 ン 量スファラド
量ジーメンス
ファラド
静コ
FS F
C/V
kgm
s-2kg
A2-1 s A
mm2
Wb Ω
Vs V/A
2 A-3 -2
2 kg -3
電
抗ウエーバ
オーム
電磁
気気
抵抵
抗束オーム
Ω
V/A
mm
kgm
s skg
A-2s A2-2 -1-2 s -1A
3 2
Wb/m
コ磁
ST S
A/V A/V mkg
コ ンン ダ束ダ ククタ密タン ンス度ジーメンス
ステスラ
ジーメンス
kgm-2s-2
A-1
kg
s3 A2
2
-2
イ
ン
ダ
ク
タ
ン
ス
ヘンリー
H
Wb/A
2 kg -2-1 -2 -1
磁磁束 ウエーバWb
Vs Vs
kgms2skg
AAmm
束 ウエーバWbs A
セ ル シ ウ ス 温 度 セルシウス度(ｅ)
℃
K
2
磁磁
束束
密密
度 テスラ
T T
s-2 A-1 -2 -1
Wb/m
Wb/m2 kg光
束度ルテスラーメンlm
cd kg-2s -2A
cd sr(c)イイ ンン ダダ ククタ タン ンス ヘンリー
H H
Wb/AWb/A
A s-2 A-2
m2 kg
ms2 kg
照度スルヘンリークス
lx
m-2 cdlm/m2(ｅ) (ｅ)セセ ルル シシ ウウス ス温 （度℃K -1 Kｆ ）度（ｄ）
セルシウス度Bq ℃s
放 射 性 核 種 の 放 射 能 温 セルシウス度
ベクレル
(c)
光光
束 ルーメンlm lm
cd srcd sr(c) cd cd
吸収線量, 比エネルギー分与, 束 ルーメン
2
-2 2 -2
グレイ
J/kg
m
s
照照
度ル
lxGy lx
lm/m
度ク
ルス
クス
カーマ
lm/m2 m-1 cdm-2 cd
（ｆ）
（ｄ）
s
（ ｆベクレル）（ｄ） Bq-1
放線量当量,
射性核種
の
放
射
能
Bq
放 射 性 核周辺線量当量,
種 の 放 射 能方向 ベクレル（ｇ）
2 s
Sv
J/kg
m s-2
吸収線量, 比エネルギー分与, シーベルト
性線量当量,
個人線量当量 グレイ
GyJ/kg
吸収線量, 比エネルギー分与,
m2 s-2 2 -2
グレイ
Gy
J/kg
カーマ
m s
酸
活
性 カタールkat
カーマ 素
s-1 mol
線量当量, 周辺線量当量, 方向（ｇ）
J/kg
線量当量, 周辺線量当量, 方向m2 s-2 2 -2
シーベルト
（ｇ）Sv
性線量当量, 個人線量当量

照 射 線 量 （ Ｘ 線 及 び γ 線 ） クーロン毎キログラム
C/kg
kg-1 sA輝
度 ワット毎平方メートル毎ステラジアン W/(m2 sr) m2 m-2 kg
s-3=kg s-3線
量率 グレイ毎秒
Gy/s
m2 s-3性濃度 カタール毎立方メートル kat/m3
m-3 s-1 mol
放射
強度 ワット毎ステラジアン
W/srm4 m-2 kg s-3=m2 kg s-3
放射輝
度 ワット毎平方メートル毎ステラジアン W/(m2 sr) m2 m-2 kg s-3=kg s-3
酵
素
活
性
濃
度 カタール毎立方メートル kat/m3
m-3 s-1 mol

放
射
吸
収
酵
素
活

ヨ ク ト

y

h

時

1h =60 min=3600 s°
1°=(π/180)
rad s
時度
日 h 1h =60
d min=3600
1 d=24 h=86
400 s
’ 1 1’=(1/60)°=(π/10800)
rad
日分
d
d=24
h=86
400 s rad度°
1°=(π/180)
1”=(1/60)’=(π/648000)
rad度秒° ” 1°=(π/180)
rad分’
1’=(1/60)°=(π/10800)
rad
ヘクタール
1ha=1hm2=104m2 rad分’ha 1’=(1/60)°=(π/10800)
秒”
1”=(1/60)’=(π/648000)
3
3
3-3 3 radリットル
L，l
1L=11=1dm =10 cm
=10
m
秒
”
1”=(1/60)’=(π/648000)
rad
2
ヘクタール
ha 1ha=1hm=104m22
トン
1t=103 kgヘクタール
hat 1ha=1hm=104m2L，l 1L=11=1dm3=103cm3=10-3m3

表７．SIに属さないが、SIと併用される単位で、SI単位で表される数値が実験的に得られるもの
名称
記号
SI 単位で表される数値
表７．SIに属さないが、SIと併用される単位で、SI単位で
表７．SIに属さないが、SIと併用される単位で、SI単位で
電 子 ボ 表される数値が実験的に得られるもの
ル ト 表される数値が実験的に得られるもの
eV 1eV=1.602 176 53(14)×10 -19J
-27
ダ
ル
ン 記号
Da 1Da=1.660
記号SI 単位で表される数値単位で表される数値
SI86(28)×10
kg
538
名称 ト 名称
-19
-19
1u=1
eV Da
J
1eV=1.602
176
53(14)×10
電統一原子質量単位
子 電
ボ 子
ル ボ
ト ル
eVu ト1eV=1.602
176 53(14)×10 J-27-27 11m
uaン1Da=1.660
1ua=1.495
97886(28)×10
706 538
91(6)×10
ダ ト単
ル ン位トDa
Da 1Da=1.660
kg
86(28)×10
ダ天 ル文
kg
538
統一原子質量単位u Da 1u=1 Da
統一原子質量単位
u
1u=1

11
11
文天 単 文 位 単ua 位1ua=1.495
ua 1ua=1.495
m
978 706 91(6)×10m
978 706 91(6)×10

天

表８．SIに属さないが、SIと併用されるその他の単位
名称
記号
SI 単位で表される数値
バ 表８．SIに属さないが、SIと併用されるその他の単位
ー
ル bar １bar=0.1MPa=100kPa=105Pa
表８．SIに属さないが、SIと併用されるその他の単位
水銀柱ミリメートルmmHg 1mmHg=133.322Pa
名称
名称 記号
記号 SI 単位で表される数値
SI 単位で表される数値
5ス ト ロ ールム barÅ １bar=0.1MPa=100kPa=10１Å=0.1nm=100pm=10-10m
バオ ン グ ーPa5バ(a)SI接頭語は固有の名称と記号を持つ組立単位と組み合わせても使用できる。しかし接頭語を付した単位はもはや実際には、使用する時には記号rad及びsrが用いられるが、習慣として組立単位としての記号である数字の１は明(a)SI接頭語は固有の名称と記号を持つ組立単位と組み合わせても使用できる。しかし接頭語を付した単位はもはや
コヒーレントではない。
示されない。
コヒーレントではない。
(b)ラジアンとステラジアンは数字の１に対する単位の特別な名称で、量についての情報をつたえるために使われる。
(c)測光学ではステラジアンという名称と記号srを単位の表し方の中に、そのまま維持している。
(b)ラジアンとステラジアンは数字の１に対する単位の特別な名称で、量についての情報をつたえるために使われる。
実際には、使用する時には記号rad及びsrが用いられるが、習慣として組立単位としての記号である数字の１は明(d)ヘルツは周期現象についてのみ、ベクレルは放射性核種の統計的過程についてのみ使用される。
実際には、使用する時には記号rad及びsrが用いられるが、習慣として組立単位としての記号である数字の１は明示されない。
(e)セルシウス度はケルビンの特別な名称で、セルシウス温度を表すために使用される。セルシウス度とケルビンの(c)測光学ではステラジアンという名称と記号srを単位の表し方の中に、そのまま維持している。
示されない。
単位の大きさは同一である。したがって、温度差や温度間隔を表す数値はどちらの単位で表しても同じである。
(d)ヘルツは周期現象についてのみ、ベクレルは放射性核種の統計的過程についてのみ使用される。
(c)測光学ではステラジアンという名称と記号srを単位の表し方の中に、そのまま維持している。
(f)放射性核種の放射能（activity
referred to a radionuclide）は、しばしば誤った用語で”radioactivity”と記される。
(e)セルシウス度はケルビンの特別な名称で、セルシウス温度を表すために使用される。セルシウス度とケルビンの(d)ヘルツは周期現象についてのみ、ベクレルは放射性核種の統計的過程についてのみ使用される。
(g)単位シーベルト（PV,2002,70,205）についてはCIPM勧告2（CI-2002）を参照。
単位の大きさは同一である。したがって、温度差や温度間隔を表す数値はどちらの単位で表しても同じである。
(e)セルシウス度はケルビンの特別な名称で、セルシウス温度を表すために使用される。セルシウス度とケルビンの(f)放射性核種の放射能（activity
referred to a radionuclide）は、しばしば誤った用語で”radioactivity”と記される。
単位の大きさは同一である。したがって、温度差や温度間隔を表す数値はどちらの単位で表しても同じである。
表４．単位の中に固有の名称と記号を含むSI組立単位の例(g)単位シーベルト（PV,2002,70,205）についてはCIPM勧告2（CI-2002）を参照。
(f)放射性核種の放射能（activity
referred to a radionuclide）は、しばしば誤った用語で”radioactivity”と記される。

m sA

10-24

da

リットル

Sv
J/kg
m s
シーベルト
(a)SI接頭語は固有の名称と記号を持つ組立単位と組み合わせても使用できる。しかし接頭語を付した単位はもはや性線量当量,
個人線量当量kat
s-1 mol-1酸素
活性 カタールkats mol(b)ラジアンとステラジアンは数字の１に対する単位の特別な名称で、量についての情報をつたえるために使われる。

C/m

カ

p
f
a
z
y

3
リットル
L，l 1L=11=1dm3=10
cm3=10-3m3
トン
t
1t=103 kg
トンt
1t=103 kg

酸 コヒーレントではない。
素活性 カタール

電 束 密 度 ， 電 気 変 位 クーロン毎平方メートル

デ

d
c
m
µ
n

表６．SIに属さないが、SIと併用される単位
名称 表６．SIに属さないが、SIと併用される単位
記号
SI 単位による値
分
min=60s
名称 min 1記号
SI 単位による値
表６．SIに属さないが、SIと併用される単位
時
h
1h
=60SI
min=3600s
名称単位による値
分 記号min1
min=60s
日
d
1
d=24
h=86
400
s
分
min 1 min=60s

組立量

-1
2
射 線 量 （ Ｘ 線 及 び γ 線 ） クーロン毎キログラム
C/kg
sA-2 mol-1-1 4 2モ照誘 ル エ ネ電 ル ギ ー ジュール毎モル率 ファラド毎メートル J/mol
F/m mkgkg
ms-3 kg
s A
2 -3
吸
収
線 モル熱容量量率ジュール毎モル毎ケルビン
グレイ毎秒
Gy/s K)
2 s -2 -1-2 -2-1m
モルエントロピー,
J/(mols K mol
透磁率 ヘンリー毎メートル
H/m m kg
4 m-2kg s -3 A 2
放射
強度 ワット毎ステラジアン
W/srm-1 m 2 kg s -2=m kg
s-3
照モ
射 線ル
量 （エＸ 線
及びγ線）
クーロン毎キログラム
C/kg
ネ ル ギ ー
ジュール毎モル
J/mol kg 2 sA
m kg s mol-1-3
ワット毎平方メートル毎ステラジアン W/(m2 sr)輝mm-22 kg s-3-2=kg
s -1
2 -3
吸放
収 射 線
量
率度グレイ毎秒
Gy/s
-1
m
s
3
ジュール毎モル毎ケルビン
モルエントロピー,
モル熱容量J/(molK)4 -3 m
kg-3s K2 mol
素射 活性
カタール毎立方メートル
s-1kg
mol
kat/m
放酵
強 濃 度度ワット毎ステラジアン
W/sr
mm
m-2
s =m kg s-3記号24（a）量濃度（amount concentration）は臨床化学の分野では物質濃度（b）これらは無次元量あるいは次元１をもつ量であるが、そのことを表す単位記号である数字の１は通常は表記しない。
（substance concentration）ともよばれる。
（a）量濃度（amountconcentration）は臨床化学の分野では物質濃度（b）これらは無次元量あるいは次元１をもつ量であるが、そのこと（substance concentration）ともよばれる。
を表す単位記号である数字の１は通常は表記しない。
（b）これらは無次元量あるいは次元１をもつ量であるが、そのことを表す単位記号である数字の１は通常は表記しない。

SI 組立単位(g)単位シーベルト（PV,2002,70,205）についてはCIPM勧告2（CI-2002）を参照。
組立量
SI 基本単位による名称
記号
表４．単位の中に固有の名称と記号を含むSI組立単位の例
表し方
SI 組立単位
表４．単位の中に固有の名称と記号を含むSI組立単位の例
粘度 パスカル秒Pa s
m-1 kg s-1
組立量
基本単位によるSI-2
名称記号
力 の モ ー メ ン ト ニュートンメートル
NSI
m組立単位
m2 kg s表し方
-2
基本単位による-1 s SI
面 組立量張ニュートン毎メートル
N/m
名称
記号mkg
粘表
度力パスカル秒Pas
kg s-1-1 -1
ラジアン毎秒2 m -2s =s-1表し方
m
力角 の モ ー速 メ ン ト度ニュートンメートル
Nrad/s
m 2
kg
s
m
-1 -2
-1
-1
粘
パスカル秒
Pa s
角
加
速
度度ラジアン毎秒毎秒
m mm s kg
=ss-2
rad/s
表
面
張
力 ニュートン毎メートル
N/m 2
kg s-2-3 2
熱度 ,ー放 メ射 照
力 流の密 モン 度トワット毎平方メートル
ニュートンメートルNm
kg s-1m -1kg-1s-2W/m
角
速度 ラジアン毎秒
rad/s
s =s
mm
ト張ロ ピ度ーラジアン毎秒毎秒
J/KN/m
表 容 量加,面エ ン速力ジュール毎ケルビン
ニュートン毎メートル rad/skg
s-2-2K-1
m2 -1
2
kg
角熱
mm
s-2s=s-2-2 -1-1 -1
熱容
，比
エ
トロ
J/(kg
2 K)
K
ms2-3sm
角流度ジュール毎キログラム毎ケルビン
ラジアン毎秒m s =s-1
熱比
密量度
放ン射
照 ピ度ーワット毎平方メートル, 速kg
W/mrad/s
2
比
エ
ネ
ル
ギ
ー度ジュール毎キログラム
J/kg
m2 s-2 -1-1s-2=s-2
ラジアン毎秒毎秒
熱角容 量 , 加エ ン ト速ロ ピ ー ジュール毎ケルビン
J/K rad/s2 m2 kgms-2m
K
ワット毎メートル毎ケルビン
熱 流 密伝度 , 放
導 射 照率度W/(mK) mkg s-1-3s-3K-1
ワット毎平方メートル J/(kgW/m
比熱
熱容量，比エントロ
ピ ー ジュール毎キログラム毎ケルビン
K)K -2
m2 s-1-2kg
3
-2 -1
積量 エ
ギピ
J/mJ/K
2 -2kg2 s
熱 容
, エネル
ンル
トギロ
ージュール毎立方メートル
ジュール毎ケルビン J/kg
比体エネーージュール毎キログラムs m -3kg -1s K
mm
2 -2
-1 -1
のエ
さワット毎メートル毎ケルビン
V/mkg
ジュール毎キログラム毎ケルビン
比 熱 容界量，比ト ロ率
ピーボルト毎メートル
J/(kg
熱電
伝
導ン強
W/(m
K) K)
sA
K
m
mmkg
s-3s K3-3
2 -2
密 ギギ ー度ジュール毎立方メートル
3J/kg
-1 m
C/m
比 積 エエ荷 ネ
ルークーロン毎立方メートル
ジュール毎キログラム J/m
体電
ネ ル
mm
kgsAs-2s
表面
電
荷率クーロン毎平方メートル
C/m2
m-2 m
sA
-3 -1s-3 K-1
ワット毎メートル毎ケルビン
導K)
電熱 界 伝 の強さ ボルト毎メートル
V/m W/(mmkg
A
2-2 s kg
電 束 密 度 ， 電 気 変 位 クーロン毎平方メートル C/mm sA-1-2
3ー ジュール毎立方メートル
電体 積 荷エ ネ 密 ル ギ度 クーロン毎立方メートル
m -1 kg
J/m3 m-3-3sA
C/m
4 s2
誘
電
率 ファラド毎メートル
F/m
m
2-2 kg s A
-3 -1
表電
面
C/m V/m m sA界の 電 強 荷 クーロン毎平方メートル
さ ボルト毎メートル
m
kg
-2 s
-2 A
磁
率 ヘンリー毎メートル
H/m
m kg s-3 A
電透
C/m2C/m3 m-22sA
電束 密 度荷， 電 気密 変 位 クーロン毎平方メートル
度 クーロン毎立方メートル
m -2sA2 -1-3 kg
ル エ 電ネ ル ギ 率ーファラド毎メートル
ジュール毎モル
J/mol 2 mm
mol
誘モ
F/m
kg-1-2ss4sA
A
表面
電
荷 クーロン毎平方メートル
C/m
2 m -2-1-1
ジュール毎モル毎ケルビン H/m
モル熱容量J/(mol K)
2 mmkgkg
透モルエントロピー,
磁率 ヘンリー毎メートル
s-2-2sA-2K mol

表５．SI 接頭語
表５．SI 接頭語
接頭語
記号 乗数
接頭語
記号
乗数
接頭語
記号-1 乗数
接頭語
ヨ 24 タ表５．SIＹ
デ -1 シ
d
10
10
ヨ
タ 接頭語
Ｙ -2
デ
シ
10
10
21
ゼ 21 タ 記号
Ｚ 乗数
セ ン
チ 記号
c
10
10
乗数
接頭語
接頭語
-2
ゼ
タ
Ｚ
セ
ン
チ
10
10
18
-3
24
-1
エ
ク
サ
Ｅ
ミ
リ
m
10
10
ヨ
タ
Ｙ
デ
シ
d
10
10
ク サ
Ｅ
1018 エ10-3 ミ µ リ
2115
-2-6
ペ
タ
Ｐ
マイクロ
10
10
ゼ 15 タＺ
セ ン-6 チ
c
10
10
ペ
タ
Ｐ
マイクロ
10
10
1812
-3-9
10
10
エテク
サラ ＥＴ
ミナ -9 リノ mn
10
10
ラ 10
Ｔ
ナ p ノ
1012 テ
15 9-6-12 ピ10
ギ
ガ
Ｇ
コ
10
ペ 9 タ
Ｐ
マイクロ
µ
10
10
ギ
ガ 10
Ｇ-15 10-12 ピ f コ
10
12 6
テメ
ラガ ＴＭ
ノ
n
1010
10-9 ナフェムト
6
-15
3
-18
メ
ガ
Ｍ
フェムト
10
10
9
-12
10 ギキ
10 ピア
ガロ Ｇｋ
コト p a
10
10
-18
-21
ｋ
62
-15
103クガキト Ｍｈロ 10
ゼ10プ トア f z ト
10
10
メヘ
フェムト
10
2
-21
1
-24
3
-18
ｈ
10 ロヘ
カ ク
y ト
10
10
キデ
ｋdaト 10
アヨ10クトトゼ aプ
10
1
-24
-21
ゼ 10
プ ト ヨ zク ト
102 ヘ10ク トデ ｈ カ 10da
乗数

ー

ル

bar

１bar=0.1MPa=100kPa=10 Pa

海里mmHg
Ｍ 1mmHg=133.322Pa
１M=1852m
水銀柱ミリメートル
水銀柱ミリメートル mmHg 1mmHg=133.322Pa
-28 2
-10
ー ロ ー ムン Åb １Å=0.1nm=100pm=10
１b=100fm2=(10-12cm)2=10
オバン グ ス ト
m m
オン
ト
ー
Å １Å=0.1nm=100pm=10-10m
ッ グ ス里トロＭ
１kn=(1852/3600)m/s
knム１M=1852m
海ノ海ー
里 Ｍ １M=1852m
-28 2
バネ
ー
ンパ bNp １b=100fm2=(10-12cm)2=10m
バー ル Ｂン SI単位との数値的な関係は、
b１b=100fm2=(10-12cm)2=10-28m2
ノベ
ット kn １kn=(1852/3600)m/s
対数量の定義に依存。
ノー ベ ッジ
dBト kn １kn=(1852/3600)m/s
ネデ
パル Np
SI単位との数値的な関係は、ネーベル Ｂ パ Np
SI単位との数値的な関係は、対数量の定義に依存。
デジベ ベル dB ル Ｂ
対数量の定義に依存。

デ

エ
ダ
エポ
ダス

表９．固有の名称をもつCGS組立単位
ジ
ベ
ル dB
名称
記号
SI 単位で表される数値-7

ル
グ erg 1 erg=10 J
表９．固有の名称をもつCGS組立単位
-5
イ
ン 記号
dyn 1 dyn=10
表９．固有の名称をもつCGS組立単位
N
名称
SI 単位で表される数値-2-7
記号P=1 dyn
cm単位で表される数値=0.1Pa s
ルア 名称
グズ ergP 1 1erg=10
J sSI
2 -1
2 -1
トエイー クルンス dyn
Stグ1 1dyn=10
erg
St =1cm
s =10-7-4m
1-5N
erg=10J s-2-54
チア ル イズブ Psbン1 1P=1
sb =1cd
cd ms-2
ポス
ダ
dyn
dyn
scm
cm-2=10=0.1Pa1 dyn=10N-2-4 24 -1 -2
2 -1
sr
スフ ト ポーォ ク アスト Stphズ1 1Stph=1cd
s cm
=10
ms lx
scm =0.1Pa s
P=1cm
1 P=1
dyn10
-2
-2 -2 4 -2
2 ms
-1 -2
Galス1 1sbGal
=1cm
=10cd
スガ チス ル
=10
St=1cdト ーブルクsb
1cm
St s=1cmsm=10-4m2 s-1
2-2
4-8 -2
ｪ
ル
Mx
=10
Wb
1
Mx
=
1G
cm
フマ クスス
ォ ウ
ト
ph
10cm
lx =104cd m-2
チル
ブ1 ph=1cd
sb 1srsb =1cd
-2 -2 -4 -2-2
4
Gト1 1Gal
=10
G
cm
ガガ
=1cm
s-2=10msT
フウ
ォルス Galph=1Mx1 ph=1cdsr
cm 10 lx
3 2-1ｃ）
Oe 1 1Mx
-2
-2
-2
(10cm
/4π)A
mWb
Oe= 1G
マエ ク
ウ ドｪ（ ルMx
ル スステ ッ=10-8

ル Gal 1 Gal =1cm s =10 msガガ-8=1Mx cm-2 =10-4T 2（c）３元系のCGS単位系とSIでは直接比較できないため、等号「マウ ク ス スウ Gｪ ル1 GMx
1 3Mx = 1G-1cm =10 Wb 」
は対応関係を示すものである。
エルス
テ ッ ド （ ｃ ） Oe 1 Oe (10 /4π)A m
-2
-4

ガ

ウ

ス

G

1 G =1Mx cm =10 T（c）３元系のCGS単位系とSIでは直接比較できないため、等号「3-1エ ル ス テ ッ ド （ ｃ ） Oe 1 Oe (10 /4π)A mは対応関係を示すものである。
表10．SIに属さないその他の単位の例」（c）３元系のCGS単位系とSIでは直接比較できないため、等号「名称記号
SI 単位で表される数値は対応関係を示すものである。
キュ
リ表10．SIに属さないその他の単位の例
ー Ci 1 Ci=3.7×1010Bq-4
レ ン名称
ト ゲ ン 記号R
1 R = 2.58×10
C/kg
SI 単位で表される数値-2
10
rad 1 1Ci=3.7×10
表10．SIに属さないその他の単位の例
rad=1cGy=10
キラ ュ
リ
ード Ci
Bq Gy
-2
-4 SI
記号
rem 1 1R
rem=1
cSv=10
レレ ン ト ゲ名称
ンム R= 2.58×10
C/kgSv単位で表される数値

」

10
-2
γー1 1γ=1
キ ン ュ ドマリ rad
Ci nT=10-9T
1 Ci=3.7×10
Bq
ラガ
rad=1cGy=10
Gy-4-2ェ ンル トムミゲ
レR
1cSv=10
R = fm=10-15m
2.58×10C/kg
レフ
rem ン1 1フェルミ=1
rem=1
Sv-2 mg = 2×10-4kg
1メートル系カラット= 200
ラン
rad nT=10-9T
ガメートル系カラット
マ γ ド1γ=1
1 rad=1cGy=10
Gy
-2
1rem
Torr =
325/760)
フト ェ
ミル Torr
fm=10-15m
レ ル
ム1フェルミ=1
1 (101
rem=1
cSv=10Pa
Sv

標 準
1γ
atm =1γ=1
101 325Pa
メートル系カラット= 200 mg = 2×10-4kg
ガ 大 気ン 圧 atm
マ1メートル系カラット
nT=10-9T
1cal=4.1858J（｢15℃｣カロリー），4.1868J
トカ(101 325/760)fm=10-15m
Paフェ
ル Torrcalミ1 Torr = 1フェルミ=1ロ
リ ルー
標 準 メートル系カラット
大 気 圧 atm 1（｢IT｣カロリー）4.184J（｢熱化学｣カロリー）
atm = 1メートル系カラット
101 325 Pa= 200 mg = 2×10-4kg
ミク
ロンµ 1cal=4.1858J（｢15℃｣カロリー），4.1868J
1 µ =1µm=10-6m
ト リ
カロ
ー cal ル Torr 1 Torr = (101 325/760) Pa

標
ミ

ク

カ
ミ

（｢IT｣カロリー）4.184J（｢熱化学｣カロリー）

準 大 気 圧 atm 1 atm = 101 325 Pa
ロンµ
1 µ =1µm=10-6m
1cal=4.1858J（｢15℃｣カロリー），4.1868J
ロ
リ
ー cal
（｢IT｣カロリー）4.184J（｢熱化学｣カロリー）

ク

ロ

ン

µ

1 µ =1µm=10-6m（第8版，2006年改訂）（第8版，2006年改訂）

この印刷物は再生紙を使用しています