REPORT OF
THE FNCA2007 WORKSHOP ON
THE UTILIZATION OF RESEARCH REACTORS

Following the agreement at the Eighth Coordinators Meeting 2007 in Tokyo, Japan, the FNCA 2007 Workshop on The Utilization Of Research Reactor took place as follows;

The Opening Ceremony began with the Welcoming Address by Dr. Aang Hanafiah Wangssatmaja, FNCA Coordinator of Indonesia, Deputy Chairman of BATAN, and followed by the Opening Address by Dr. Fumio Sakurai, Chairman for the field of URR, FNCA Deputy Director General, Nuclear Science Research Institute, Tokai Research and Development Center, Japan Atomic Energy Agency (JAEA) and Dr. Hudi Hastowo, Chairman of BATAN

Two special lectures were given.

Invited Lecture:“Experience on utilization of RSG-GAS reactor” (Ms. Endang Susilawati, BATAN, Indonesia)

Invited Lecture:“Solar system abundances of the elements” (Prof. Mitsuru Ebihara, TMU, Japan )

Then Parallel Sessions of Neutron Activation Analysis Group (NAA) and Research Reactor Technology Group (RRT) were carried out.

(RRT)

 This year is the final year of current project. Through the activities for three years, the establishment of a common neutronics calculation technique on core management among member countries will enhance safe and stable operations as well as effective utilizations of research reactors of participating countries. The following four activities were discussed; (1) examination of calculated results on the burn-up of domestic core and application for core management or utilization by SRAC or MVP, (2) discussion of the final report of current project, (3) discussion of the evaluation of current project and (4) determination of the concrete activity schedule of next project.

(NAA)

Participants made technical tour to Research Facilities; Siwabessy Research reactor (RSG-GAS reactor), NAA laboratory, the fuel element fabrication lab, and radioactive waste management lab., Serpong.

Summary Reports of NAA and RRT were presented by the participants.

REPORT OF
THE FNCA2007 WORKSHOP ON
NEUTRON ACTIVATION ANALYSIS (NAA)

The session starts by introduction of participants, their experience and involvement in FNCA activities.

Prof. Mitsuru Ebihara, TMU, Japan, presents the first session on gPerspective of the FNCA 2007 Workshop on URR/NAAh in which he emphasizes on the activities to be carried out during the workshop. This includes presentation and discussion on:

1. Country report I: gThe Progress report of environmental sample analysis and its reflection to the environmental protection authority for the past one year.h
2. Country report II: gThe summary of phase 2 (2005-2007) : sample analysis and its reflection to the environmental administrationh
3. Presentation and discussion on three-years-evaluation
4. Discussion on phase III in which all participating countries have expressed their needs to continue in the third phase including Australia.
5. Subprojects

1. Presentation on results of Round Robin Test
2. Others
   

Country Reports I

gThe progress report of environmental sample analysis and its reflection to the environmental protection authority for the past one year.h from each participant was presented in the alphabetical order of countries:

Bangladesh : Dr. Syed Mohammod Hossain presents the research work as follows:
According to the work plan for 2007, arsenic contamination was assessed in 50 tubewells water, 60 corresponding usersf and 30 soil samples from Faridpur, Narayanganj and Munshiganj Districts of Bangladesh using INAA technique. The results revealed that most of the investigated tubewells and their users are highly contaminated. Some of soils are also containing high arsenic compared to the world reference values.

China: Prof. Bangfa Ni presents the research work carried out as follows:
Twice a month of air particulates sampling was carried in 2007 at Liangxiang site. Multielements data of air particulates from 2004 to 2006 have been interpreted primarily by using EPA PMF and Statgraphic Plus softwares. Five source factors have been found. It is an important support for local EPA decision of environmental management. Twelve to thirty three elements have been determined in 12 bio-reference materials by using k₀ INAA and relative methods. The results are very valuable for certification.

Indonesia: Mr. Sutisna presents the research work carried out as follows:
The analytical results of elemental distribution of marine environmental samples for the Jakarta bay and north-coastal of Banten have been presented. The pollutant monitoring of marine environment has been done by using k₀-INAA, under collaborative work with Environmental Management Agency of DKI Jakarta. The elemental distribution obtained could give valuable information of condition and a level contamination of location/region interested.

Japan: Dr.Yasuji Oura presents the research work carried as follows:
Collection of PM_2.5 particulates was started at Hachioji, suburban area, and Koto, urban area, in 2006 in cooperation with Tokyo Metropolitan Research Institute for Environmental Protection. The particulate concentration level of PM_2.5 was almost same at Hachioji and Koto. About 30 elemental concentrations were determined by k₀-INAA or a comparison NAA method. The median values of almost elemental concentrations at Koto were about 2 times in average higher than Hachioji. Elemental correlations were significantly different between Hachioji and Koto.

Korea: Mr. Jong-Hwa Moon presents the research work as follows :
Sixty air particulate (PM_2.5) samples were collected from the roof-top of the third engineering building of Chungnam National University. Twenty eight elements were analyzed by using INAA. The variation of the mass concentration for the PM_2.5 was 5.6 to 64.5 g/m³ and the average was 30.9 g/m³. Following the analytical results, the elements such as Al, K, Fe and Na are placed in the highest level and Dy, Sc, Hf and Cs are in the lowest level.

Malaysia ; Mr Md Suhami Bin Elias presents the research work as follows :
The k₀-INAA method was applied for analysis of trace and toxic elements in marine environmental samples. About 29 trace and toxic elements were determined in marine sediment samples collected from 30 locations at east coast of Peninsula Malaysia. The major contribution sources of trace and toxic elements are from anthropogenic activities such as agriculture, industrial, vehicle and ore mining activities.

Philippines : Ms Preciosa Corazon Pabroa presents the research work on as follows:
The paper focused mainly on the elevated Pb levels in Valenzuela compared with the other sampling sites such as in ADMU, POVEDA and NAMRIA. Pb source factors show up in both the PM_10-2.5 and the PM2.5 fractions. CPF analysis shows Pb source coming from almost all directions. Results in Valenzuela City indicate the need to do a more comprehensive evaluation of the area to determine the sources of Pb and formulate measures to bring down its ambient levels. Valenzuela average for 2006 is in exceedance of the Philippine NAAQ long term standard and the US EPA long term standard.

Thailand: Dr. Sirinart Laoharojanaphand presents the research work as follows:
Core sediments from six sampling sites in Thalaenoi Conservation Area were collected in February of 2007. Total of 188 samples were analyzed by comparative INAA. Mn, Al and Fe showed depth variation and site dependence. Additionally, some samples contain high level of Mn comparing to the Thailand Standard of Maximum Allowable Value in Soil. Arsenic levels in all samples are higher than this standard. The data will be made available to the Department of Environmental Promotion, Ministry of Natural Resources and Environment, the end-user of the project.

Vietnam : Dr. Nguyen Ngoc Tuan presents the research as follows :
For the past one year in 2007, three coastal areas of Nha Trang, Phan Thiet and Ganh Rai bays have been chosen. The methods for analysis are ko-INAA, RNAA and AAS. More than 20 elements in marine environmental samples (sediment, sea water and biota) collected in above areas have been determined. Some 700 targets in 30 various samples were reported in this workshop.

Country Reports II

The summary of phase 2 (2005-2007): gSample analysis and its reflection to the environmental administrationh from each participant was presented in the alphabetical order of countries:

Bangladesh : Since Bangladesh joined the FNCA/URR/NAA group in 2006, there is no summary report of phase 2 to be presented at this workshop. The research on gAnalysis of Environmental Pollution due to Discharge of Industrial Effluents Using Neutron Activation Analysis Techniqueh is presented here.
The physicochemical parameters, salinity and heavy metal contamination were assessed in the effluents discharged from textile industries in the Dhaka Export Processing Zone using INAA technique. The guidelines have given how to develop a cost effective pilot plant for the treatment of waste water prior to their disposal to the surrounding environment. In addition, Cr toxicity is assessed in the soils of existing Hazaribag tannery industries and the proposed Tetuljhora tannery area. The Hazaribag area is highly contaminated with Cr, whereas Tetuljhora soils contain normal levels compared to the worldwide background.

China: Prof Bangfa Ni presents the summary report of phase 2 as follows:
More than 400 filters have been collected by using GENT sampler and nuclear pore filter since 2004 at Liangxiang area, Beijing. More than 400 filters were analyzed. About 40 elements were determined by k₀-INAA. Data interpretation is being carried primarily by using statistic softwares. Five pollutant sources were identified primary, such as soil, limestone, combustion, traffic and dust. The information of the program provided supported the local governmentfs decisions. Some small coal mine and limekiln surroundings were closed by local government. Some cement factories were removed. Coal burning was replaced by natural gas or liquefaction gas. The results indicated that the air quality at Liangxiang area was improved. The skill and knowledge of young scientists are improved greatly for NAA and air pollution study. End usersf acceptance and understanding effectiveness of air pollution study using INAA and statistical software were increased.

Indonesia: Mr. Sutisna presents the summary report of phase 2 as follows :
The Environmental Management Agency (BPLHD) DKI Jakarta has a commitment to use the nuclear analytical technique in the marine environmental monitoring. The collaboration with the agency established since end of 2005 is ongoing. The distribution of elements on sediment samples taken from 30 sampling sites at the Jakarta Bay, and 10 sampling sites at north coastal of Banten have been evaluated. The elements of Sc, Cr, Rb, Sr, Ag, Sb, Cs, La, Ce, Nd, Eu, Yb, Th, U, Fe, Zn, Pb, Cu, Cd and Br have been reported. The result of this project was evaluated to be very useful.

Japan: Prof. Mitsuru Ebihara presents the summary report of phase 2 as follows :
This report summarized the activity which the NAA group of Japan have been performing during the period of 2005-2007 and evaluated it in comparison with the goals established at the beginning of this period. Sampling of SPM has been implemented at three sites of Hachioji (suburban), Sakata (rural) and Koto (urban), and SPM samples were analyzed by INAA. Sampling was also performed at Cape Hedo (Okinawa), which is remote from human activity as well as industrial one. As a result, it is evaluated that our goals set for the three years period are mostly fulfilled.

Korea: Mr. Jong-Hwa Moon presents the summary report of phase 2 as follows :
In order to monitor an air quality in Daejeon City by using an annular denuder sampler and INAA, the mass concentration and the concentration of twenty eight elements of PM_2.5 were determined. The data obtained will be provided to an environmental authority that is under the auspices of the Ministry of the Environment, after being published as research papers in domestic and/or international journals.

Malaysia : Mr. Md Suhami Bin Elias presents the summary report of phase 2 as follows :
The research and development activities has been undertaken through various collaboration mechanisms that included participation in research activities organized at national and international level. Nuclear Malaysia has collaboration with the universities and government agencies such as UTM, UiTM, UPM, UKM, UMT, AELB and Malaysian Fisheries Institute. Whilst at the international level Nuclear Malaysia have close collaboration with IAEA and FNCA program. The objective of the collaboration between Nuclear Malaysia and national as well as international institutions is to encourage research activities related to nuclear analytical techniques. The marine sediment samples were collected from east coast Peninsula Malaysia. The samples were analyzed by INAA technique. The data obtained from this study will be important to the environmental agency for the proper management of the country.

Philippines : Ms Preciosa Corazon Pabroa presents the summary report of phase 2 as Follows :
Valenzuela sampling site has PM₁₀ and PM_2.5 average levels increasing since 2004 to 2006, exceeding the Philippine NAAQ long term standard for PM₁₀ in 2006. PM₁₀ Pb levels in Valenzuela in 2005 shows elevated levels in both the PM _10-2.5 and the PM_2.5 fractions which is reflected in the PMF runs showing Pb source factors in both the coarse and the fine fractions. Conditional Probability Function (CPF) analysis indicates this pollutant to come from almost all directions. Linkages were established with the Environmental Management Bureau, the Palawan local government and the Makati Environmental Protection Council. A total of 40 air filter samples from the POVEDA and UST sampling sites were analyzed by both the EDXRF system (from PNRI) and INAA (from TMU). With Pb in particular, EDXRF shows an advantage over the use of the INAA in that the latter is not capable of analyzing Pb.

Thailand: Dr. Sirinart Laoharojanaphand presents the summary report of phase 2 as follows :
The utilization of research reactor under scope of FNCA was applied to two subprojects namely air quality study in Saraburi Province and pollution study of Thalaenoi Conservation Area. The air particulate data were utilized by the Pollution Control Department, Ministry of Natural Resources and Environment. Results from Thalenoi project will be combined with additional information to be utilized by the Aquatic Research and Development Unit, Environmental Research and Training Center, Department of Environmental Promotion, Ministry of Natural Resources and Environment. The results will be presented at a technical meeting hosted by the Department of Environmental Promotion and Pattalung Provincial Authority to increase public awareness on sustainable utilization of Thalaenoi.

Vietnam : Dr. Nguyen Ngoc Tuan presents the summary report as follows :
Some 18-25 elements in marine environmental samples in Nha Trang, Phan Thiet and Ganh Rai bays are determined. The targets are Al, As, Br, Ca, Cd, Ce, Co, Cr, Cs, Cu, Eu, Fe, Hg, K, La, Mn, Na, Pb, Rb, Sb, Sc, Se, Sm, Sr, Th, U, V and Zn. Marine environmental samples chosen are sea water (18 elements), biota (22 elements) and sediment (25 elements). Analytical techniques for the elemental analysis are INAA, RNAA and AAS. More than 60 samples are collected in three costal areas. The obtained analytical results will be the database to study and monitor marine environmental situations in the future.

Presentation and discussion on Three-Year-Evaluation (2005-2007)

Each country presents its own evaluation.

Discussion on proposal for Phase III

The proposal for continuation of URR/NAA for the next phase (Phase III) has been agreed by all countries participating including Australia which has expressed their interest in joining the activities. Three common targets for phase III has been identified. These are geochemical samples (6 countries), food samples (7 countries) and environmental samples (7 countries). As soon as the proposal has received the approval from the next FNCA Coordinatorfs Meeting in early 2008, the e-mail communication will be generated for each country to set up goal for each year as well as the goal of the whole project. The conclusion will be presented at the future FNCA 2008/URR/NAA meeting. Since in the past experience almost all countries participated are using ko-INAA extensively in their own projects, the party agreed that the validation of the technique is also another topic to be carried out in phase III. Thailand has expressed the request for the assistance for the hands-on launching of k₀-INAA in Thailand. From the results of phase III, several technical papers are expected to be published in local and/or international journals.

Target materials proposed for the period of 2008-2010

SUMMARY REPORT OF
THE FNCA2007 WORKSHOP ON
THE RESEARCH REACTORS TECHNOLOGY (RRT)

1. Objective

The Objective of Research Reactor Technology (RRT) project is “Sharing Neutronics Calculation Techniques for Core Management and Utilization of Research Reactors”. The main objectives of the 3rd Workshop (WS) are to (1) examine calculated results on the burn-up of domestic core and application for core management or utilization by the distributed common code (Standard Reactor Analysis Code (SRAC) of Japan) in each participating country, and to (2) discuss the final report of this project, and to (3) discuss the evaluation of this project and to (4) determine the concrete activity schedule of next project.

2. Opening and confirmation of agenda

At the beginning of the workshop, Mr. H. Sagawa expressed his thanks for active cooperation for the RRT-project. He has reconfirmed the objectives and the full schedule for RRT- project shown in Table 1. This year is the final year of current project. Through the activities for three years, the establishment of a common neutronics calculation technique on core management among member countries will enhance safe and stable operations as well as effective utilizations of research reactors of participating countries. He has reviewed the activities toward the 3rd WS shown in Table 2 which was agreed in the 2nd WS in Philippines, August 2006. He has also proposed the schedule of the 3rd WS shown in Table 3. He mentioned that there were four activities to be discussed in the 3rd WS. Those are (1) examination of calculated results on the burn-up of domestic core and application for core management or utilization by SRAC or MVP, (2) discussion of the final report of current project, (3) discussion of the evaluation of current project and (4) determination of the concrete activity schedule of next project.
All participants have agreed with the proposed schedule.

Table 1. Research Reactor Technology Project Schedule

Table 2. Project Activity in the 3rd Work shop, agreed in the 2nd WS

Table 3. Activity Schedule in the 3rd WS

3. Participants

We had 9 participants from China, Indonesia, Korea, Japan, Malaysia, Thailand and Vietnam. In addition, 4 researchers from Indonesia participated in the WS as observers. The list of participants is attached in Attachment A.

4. Country reports on Burn-up Calculation of Domestic Cores and Application for Core Management or Utilization by common code

Most of members have reported that they carried out burn up calculation for their domestic cores by SRAC or/and MVP Codes. The results of changing effective multiplication factor following operation time of reactor were presented by each member. The results of burn up calculation from SRAC or/and MVP codes showed very good agreement with experimental data or calculated results by other code system. However, few member countries need more time to finish burn up calculation for their domestic cores.
MVP code was applied to investigate for neutronics calculation and utilization in some member countries like NTD, I-131 production or design new research reactor. The summary of country report from each country is given in the following:

(1) China (Prof. Luzheng Yuan)
China has used the SRAC code to deal with the burn-up and neutronics calculation for CARR. The results obtained show a slight difference if comparing with that obtained from domestic code system. Although till now China has not conduct the core fuel management and utilization of research reactor in detail, it is sure that SRAC code is convenient and useful in dealing with this field. Because in China, computer code run under LINUX system the SRAC code based upon is not so familiar like windows system, now its usage is only in the initial stage. So, the inconsistency in the results have not been explained satisfactorily which will be further dealt with later.

(2) Indonesia (Mr. Surian Pinem)
Calculation of the RSG-GAS core deals with effective multiplication factor (k_eff) and fuel burn-up distribution using SRAC-COREBN code at BOC and EOC. Value of k_eff from calculation by SRAC-COREBN compared with k_eff value from measurement result. Calculation of k_eff and fuel burn-up for the first core and typical working core (TWC) compare with IAFUEL code result. The calculation of k_eff value of first core closed to the measurement result. Fuel burn-up distribution calculation result for TWC with SRAC-COREBN showed good agreement compared with IAFUEL code.

(3) Japan (Mr. Tomoaki Kato)
SRAC has been being used for core burn-up calculation of JRR-3. PIJ, PIJBURN, ANISN and CITATION in SRAC are used for making group constant as preparation for core burn-up calculation. COREBN and HIST in SRAC are used for core burn-up calculation. Core burn-up calculation by SRAC can indicate the increasing of excess reactivity caused by burn-up of cadmium wire as burnable poison and the decreasing of excess reactivity caused by burn-up of uranium. From the results of core burn-up calculation, it is found that burn-up of cadmium wire as burnable poison is not exactly evaluated by SRAC. So as to solve the problems, the number of burn-up steps is changed from15 to 60, and nuclear data library is changed from JENDL 3.2 to JENDL 3.3. After that, it is found that the result from modified method is better compared with the result from current method, although the result has uncertainty still.
Calculation of NTD (Neutron Transmutation Doping) silicon is carried out by MVP code. It is very important for making NTD silicon to distribute nuclide ³¹P into silicon ingot. Then, the calculations of flux distribution, neutron capture reaction rate and neutron spectrum inside of silicon ingot are carried, because ³¹P is produced by the reaction of ³⁰Si(n,γ)³¹Si and the decay of ³¹Si. From these results of calculations, it is confirmed that the performance of NTD silicon at the JRR-3 is enough.

(4) Korea (Mr. Hak-Sung, Kim)
The preliminary nuclear characteristics of the reference AHR core have been studied. Major analyses have been undertaken for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, and the control rod worth. For the analysis, the MCNP, MVP, and HELIOS codes were used. The results by MCNP and MVP showed good agreements and can be summarized as follows.
For a fresh, unperturbed core condition, the fast neutron flux (E_n≥1.0MeV) reaches 1.47×10¹⁴ n/cm²s and the maximum thermal neutron flux (E_n≤0.625eV) reaches 4.45×10¹⁴ n/cm²s in the core region. In the reflector region, the maximum thermal neutron flux is estimated to be 4.06×10¹⁴ n/cm²s.
For the analysis of the equilibrium cycle core, the irradiation facilities in the reflector region were considered. The core model has 4 tangential beam tubes and a total of 18 vertical irradiation holes with different diameters in the reflector region. The cycle length was estimated as 38 days and the excess reactivity at a BOC was 103.4mk, and 24.6mk at a minimum, was reserved at an EOC. The assembly average discharge burn-up was 54.6% of the initial U-235 loading. The maximum peaking factor (Fq) was calculated as 2.56 and the corresponding linear power was 106.2kW/m. The shutdown margins by the 1st and 2nd shutdown systems were estimated to be 27.4mk and 46.3mk, respectively. Both the isothermal temperature coefficient and the power coefficient were negative, so the AHR core is characterized as being inherently safe.

(5) Malaysia (Ms. Zarina Binti Masood)
The 11^th RTP core configuration has been calculated using MVP code. Treatment of the nuclear reactions is based on continuous-energy method where the cross-sections were processed from the evaluated nuclear data files JENDL 3.3. The physical quantities were then estimated by the tallying process using collision and track length estimators. The effective multiplication factors, fission densities and flux distribution for the RTP core have been obtained. However, in future there are plans to do the burn-up calculation using SRAC and MVP.

(6) Thailand (Mr. Narin Klaysubun)
The Thai Research Reactor -1/Modification 1 (TRR-1/M1) is an open pool-type TRIGA - Mark III. The reactor uses the basic uranium-zirconium hydride (UZrH) fuel The first core configuration of TRR-1/M1 fuel is based on 8.5 wt-% and 20 wt-% of Uranium per total weight both of them are 20% enriched. The reactor has a steady-state thermal power rating of 2 MW.
The modeling of TRR-1/M1 by SRAC is performed in two major steps, i.e., the cross section generation step and the core calculation step. The group cross section generation is performed by the PIJ module of SRAC which utilizes 2D collision probability method. The TRR-1/M1 are classified into two major models, i.e., fuel type lattice and non-fuel type lattice. The COREBN module of SRAC performs the reactor calculation using the group cross section libraries generated in the cross section generation step. The COREBN model utilizes 80 axial layers with 60 axial layers representing the fuel section. The core excess reactivity calculated by SRAC is generally over-estimated when comparing with the operation data. However, the core excess reactivity of Core loading 1 calculated by SRAC and that calculated by MVP. This suggests that the core modeling of SRAC calculation is consistent with MVP. On the other hand, the differences between BOC and EOC core excess reactivities calculated by SRAC are rather different from the operation data. These differences are related to the burn-up modeling of SRAC and hence it needs further improvements.
The CORE1 of TRR-1/M1 was modeled by MVP at critical with 100 fuel elements including 4 fuel-follower control rods, 1 air-follower control rod, 3 neutron detectors and a central thimble irradiation channel in hexagonal lattice with a cylindrical frame. The burn-up calculations were performed with the MVP and MVP-BURN codes using ENDF/B-VI point wise cross-section library for all nuclides. The burn-up calculations were performed in three-dimensional geometry modeled for the whole core of each core cycle using the standard chain model (u4cm6fp50bp16T). TINT is applying the MVP code into TRR-1/M1 core management system for core loading and its utilizations. For each core cycle, MVP is used to simulate and study an optimum core. The core excess reactivity, control rod worth, the neutron flux distribution, and pin power profile are calculated and compared with SRAC’s results. The MVP code will be also used to study for the new project of in-core gemstone facility for its design and utilization.

(7) Vietnam (Mr. Nguyen Kien Cuong)
SRAC Code was used for neutronics calculation and burn up of Dalat Nuclear Research Reactor. Macro cross section for fuel assembly and non-fuel material were prepared by PIJ, ANISN, CITATION Codes. For burn up calculation, the axial of fuel assembly was divided to 12 depletion nodes, each node 5cm. HIST and COREBN codes were applied for burn up calculation and after each step of burn up, Beryllium poisoning was carried out to correct the results of excess reactivity. It is show that the very good agreement calculated results and experimental data about excess reactivity after this correction. The results of burn up distribution were compared with scanning method (ratio Cs¹³⁴/Cs¹³⁷) and have a very good agreement (under 10%) except some fuel assemblies in periphery have big discrepancy.
MVP/GMVP Code also was applied for DNRR in cell and cell burn-up calculation, whole core calculation including core burn up (MVP-Burn). Beside with purpose for core management, MVP Code was applied to calculate for radioisotopes production of I-131 in neutron trap of DNRR

5. Final Report

The theme of the project “Sharing Neutronics Calculation Techniques for Core Management and Utilization of Research Reactor” attains to an end at the workshop in 2007. At the last workshop in 2006, it was agreed that three-years-final report from 2005 to 2007 should be made as a summarized document of three-years activity.
At the session, each participant from each country presented the status of the final report and discussion about the report was carried out. And then, it was agreed that every participant complete to make final report and submit it by end of November 2007. The summary of final report from each country is as follows;

(1) China (Prof. Luzheng Yuan)
In China, several code systems with slight difference are adopted for neutronics and burn-up analysis by different users. WIMS and CITATION codes are used for core cell calculations in CARR design. Meanwhile the codes ORIGEN2 and MCNP are used to verify the whole calculation results at some particular depletion points.
Having set the modeling for fuel, core cell, reflector and irradiation channels and with the few group constants following its depletion of fuel contents, the cell calculation can be done to obtain neutron flux, multiplication factor and nuclides densities.
In SRAC, a fuel meat with cladding and both side plates, a moderator are homogenized by PIJ to obtain a fine group effective cross section for a fuel plate and both side plates region, then homogenize and collapse together with the other H₂O gap region and the other aluminum region, to obtain the few group effective cross section of a fuel element region. The similar procedure is done with the control rod region, irradiation tube region, reflector region etc. Couple with CITATION code, the core calculation can be completed by using the SRAC code.
Some results such as k_eff and nuclides densities show consistency with a slight difference between two systems. One reason might be resulted from the SRAC modeling much more simplified than that in domestic code system because of initial usage. So the further work should be performed to find out the points which result in these inconsistencies.

(2) Indonesia (Mr. Surian Pinem)
RSG GAS core burn-up calculations are carried out using IAFUEL code. IAFUEL code is 2-D neutron diffusion method code. For utilization, such as radioisotope analyses, the neutronic parameters are determined by Batan-EQUIL-2D code.
SRAC has been being used for core calculation of RSG GAS. PIJ, ANISN and CITATION in SRAC are used for making group constant as preparation for core calculation. CITATION in SRAC is used for core calculation. Core calculation results have good agreement with measured value on excess reactivity, control rod worth and thermal neutron flux. SRAC has been being used for core burn-up calculation of RSG GAS. COREBN and ASMBURN in SRAC are used for core burn-up calculation. In the calculation, the optimum operation cycle length has been determined. SRAC will be used for in-core fuel management of RSG GAS reactor and for optimize the radioisotope target.
Calculation of first core RSG GAS using Monte Carlo MVP codes has been carried out. The result of calculation has good agreement with experimental value. The RSG GAS core burn-up calculation will be carried out with MVP-BURN. The MVP code will be used for analyzing the radioisotope target, such as HEU and LEU targets for Mo production.

(3) Japan (Mr.Tomoaki KATO)
The draft of final report on part of Japan was introduced.
Japan adopted JRR-3 as a subject of neutronics calculation.
Concerning core calculation by SRAC, PIJ, ANISN and CITATION are used for making effective cross section data as preparation for core calculation. CIATION in SRAC is used for core calculation. Core calculation results have good agreement with measured value such as excess reactivity, control rod worth and thermal neutron flux.
Concerning core burn-up calculation by SRAC, PIJ, PIJBURN, ANISN and CITATION are used for making effective cross section data as preparation for core burn-up calculation. COREBN and HIST are used for core burn-up calculation. Core burn-up calculation results can indicate the increasing of excess reactivity caused by burn-up of cadmium wire and decreasing of excess reactivity caused by burn-up of uranium.
SRAC has been being used for core management since more twenty years before. Several items such as excess reactivity, fuel burn-up ratio, are calculated by SRAC as the work of core management, because these items are required for safety reactor operation.
Concerning core calculation by MVP, calculation model of JRR-3 was made preciously. The calculation result on keff at the state of just critical has good agreement with measured value.
Concerning core burn-up calculation by MVP-BURN, calculation model of JRR-3 was made. All burn-up regions are not divided along vertical direction of geometry for input data. The calculation result on keff changing is different from operation data. The investigation on the difference is the future work.
Calculation of NTD (Neutron Transmutation Doping) silicon is carried out by MVP code. It is very important for making NTD silicon to distribute nuclide ³¹P into silicon ingot. Then, the calculations of flux distribution, neutron capture reaction rate and neutron spectrum inside of silicon ingot are carried, because ³¹P is produced by the reaction of ³⁰Si(n,γ)³¹Si and the decay of ³¹Si. From these results of calculations, it is confirmed that the performance of NTD silicon at the JRR-3 is enough.

(4) Korea (Mr. Hak-Sung, Kim)
Recently KAERI launched a design project of the AHR (Advanced HANARO Reactor) in consideration of the future needs for a new research reactor based on the experiences of the HANARO. The basic neutronic code system for the AHR design is the HANAFMS (HANARO Nuclear Analyses and Fuel Management System) which was verified through the HANARO construction and operation. Now, design stage of AHR, the MCNP/HELIOS codes are mainly used for core design. We are planning to adopt the SRAC system for analyzing the neutronic characteristics of the AHR for comparison with the results of the HANAFMS.
We installed the SRAC code system successfully. It was confirmed through comparison of keff values for the sample calculation on the JRR-3 core. And we performed criticality calculation of AHR core using SRAC. The results of the SRAC code were slightly different with the results of the MCNP. Therefore the modeling of CITATION should be reconsidered. For further study of SRAC, more detail core model should be adopted for CITAION and then the burn-up calculation should be studied in order to complete application of SRAC code on AHR.
We also calculated basic neutonic characteristics and burn-up of AHR core using MVP code. MCNP and MVP show a good agreement for their flux levels within a statistical calculation error range and burn-up calculation results of MCNP/HELIOS and MVP also agreed well. We have a plan to adopt the MVP code for design of complicated irradiation and experiment equipments if the code is equipped with full burnup capability.

(5) Malaysia (Ms. Zarina Binti Masood)
Initially the TRIGAM, TRIGLAV and WIMSD/4 codes were used for the core management of PUSPATI TRIGA Reactor. TRIGAM is 1-D two-group and TRIGLAV is a 2-D four-group diffusion code. However, TRIGAM was found to be inaccurate in predicting the burn-up of fuel elements close to the water gap, irradiation channel and control rods.
SRAC with JENDL3.2 has been being used for core calculation of PUSPATI TRIGA Reactor. Core calculation results are in good agreement with measured values on effective multiplication factor, excess reactivity and control rod worth. Thermal flux and power distribution were also obtained. Burn-up calculation using COREBN in SRAC is currently on-going but the results are not finalized yet. There are plans to use SRAC for in-core fuel management of PUSPATI TRIGA reactor.
The MVP code has been used to do neutronic calculations of the 11^th core configuration of PUSPATI TRIGA Reactor. Values for effective multiplication factor fission densities and flux distribution were obtained. In the future, there are plans to carry out burn-up calculation with MVP-BURN and use MVP for core management.

(6) Thailand (Mr. Narin Klaysubun)
The Thai Research Reactor -1/Modification 1 (TRR-1/M1) is an open pool-type TRIGA - Mark III. The reactor uses the basic uranium-zirconium hydride (UZrH) fuel The first core configuration of TRR-1/M1 fuel is based on 8.5 wt-% and 20 wt-% of Uranium per total weight both of them are 20% enriched. The reactor has a steady-state thermal power rating of 2 MW.
TRR-1/M1 adopts the computer code named TRIGAP for the reactor core calculation. The code is designed for TRIGA reactor based on the Diffusion Theory. The computer program is designed to estimate the k-eff with the ring-smeared two group diffusion constants using the effective 2-g cross-sections for all types of unit cells (fuel and non-fuel) which are stored in the library as a function of burn-up. The fuel element is registered by its identification number and tracking burn up by user from burn up time to time. Assume that all unit cells are of equal volume, the cross-section was calculated with WIMS-S transport code in 18-group transport approximation. TRIGAP library together with the history of elements from the data base for TRIGAP program.
TRR-1/M1 Core and Burnup Calculations by SRAC Code System. The modeling of TRR-1/M1 by SRAC is performed in two major steps, i.e., the cross section generation step and the core calculation step. The group cross section generation is performed by the PIJ module of SRAC which utilizes 2D collision probability method. The COREBN module of SRAC performs the reactor calculation using the group cross section libraries generated in the cross section generation step. The core excess reactivity calculated by SRAC is generally over-estimated when comparing with the operation data. However, the core excess reactivity of Core loading 1 calculated by SRAC and that calculated by MVP. This suggests that the core modeling of SRAC calculation is consistent with MVP. On the other hand, the differences between BOC and EOC core excess reactivities calculated by SRAC are rather different from the operation data. These differences are related to the burnup modeling of SRAC and hence it needs further improvements.
Neutronics Calculation of Domestic Core by Use of MVP Code. The burn-up calculations were performed with the MVP and MVP-BURN codes using ENDF/B-VI point wise cross-section library for all nuclides. The burn-up calculations were performed in three-dimensional geometry modeled for the whole core of each core cycle using the standard chain model (u4cm6fp50bp16T). MVP is used to simulate and study an optimum core. The core excess reactivity, control rod worth, the neutron flux distribution, and pin power profile are calculated and compared with SRAC’s results. The MVP code will be also used to study for the new project of in-core gemstone facility for its design and utilization.
Through this project, SRAC and MVP codes have been introduced to replace the TRIGAP (a domestic code) as fuel management tools. As for calculation results, MVP gives good calculation results comparing with experimental results. On the other hand, SRAC calculation results are still not in good agreement with experimental results, thus the SRAC modeling (especially the burnup modeling part) is needed to be further improved. Even though, MVP gives good calculation results, it is computationally involved. Therefore, MVP is not practical to be used to optimize the reactor core configuration. The planned methodology for TRR-1/M1 fuel management is to use SRAC as a preliminary calculation code while MVP is to be used as a reference code. In this scheme, the optimum core loading pattern will be determined by SRAC and the optimum core pattern derived from SRAC will be modeled in detail by MVP calculation.

(7) Vietnam (Mr. Nguyen Kien Cuong)
After 3 years using SRAC and MVP Codes, almost reactor physics parameters of DNRR were calculated. The obtained results from the codes are very good agreement with experimental data or calculated results by MCNP. Both Codes now are using for core and fuel management of DNRR. In SRAC Code, macro cross section of fuel and other materials were prepared and collapsed from 107 neutron energy groups to 7 groups. Cross section of Fuel assembly and all components inside the core were calculated by PIJ. Top and bottom of fuel assembly were done by PIJ and ANISN. Six control rods (B₄C material - 4 shim rods and 2 safety rods) were considered black absorption material by determining ratio of neutron flux and neutron current at boundary following 7 neutron energy groups. JDL3.3 and ENDF/B6.8 libraries were mainly used in all calculation. PIJBURN was used to calculate burn up of fuel assembly from 0% U-235 burn up to 90% with 24 steps. For burn up whole core calculation, each fuel assembly was divided 12 nodes (5cm/node). Results of burn up calculation were confirmed by compare with experimental data. MVP Code was used to calculate critical, neutron flux and power distribution of fresh core. Burn up calculation was carried out by MVP and MVP-Burn Code. The obtained results were also very good compare with experimental data. I-131 production was calculated by MVP code in two cases different arrangement of capsules at neutron trap of DNRR. It was showed that 3 Capsules were put following axial better than 3 Capsules were put same position.

6. Evaluation of current project

Each country presented the evaluation of current project. Through the activities for three years, all countries had at least one expert who can use SRAC and MVP fluently. Some member countries are trying to apply SRAC for core management. Some member countries use SRAC or MVP for enhanced utilization of research reactors, for examples, RI production, irradiation of gemstones and so on. Moreover, some member countries are using SRAC for design of new research reactor. For the proposal of next project, many counties desire the safety analysis and thermal hydraulic analysis for the reactivity insertion accident (RIA) and the loss of flow accident (LOFA). The summary of the evaluation from each country is shown in Attachment B.
Mr. H. Sagawa asked the members to discuss the evaluation again with each PL and to send the final version of the evaluation to the FNCA secretary, Dr. K. Yamashita and Mr. H. Sagawa as soon as possible. All members agreed with that.

7. Next project

Mr. H. Sagawa expressed the proposal of new project after current project. The objective of a new project is “sharing safety analysis techniques for safety operation of research reactors”, because many countries wish reactivity insertion analysis and thermal hydraulic analysis. The title of a new project became “safety analysis of RIA and LOFA for research reactors”. He proposed that COOLOD and EUREKA code was useful in the project. Moreover, he proposed the concrete activity schedule for three years shown from Table 4 to Table 6.
All participants have agreed with the proposed calculation codes and schedule.

Table 4. Schedule of new project for 1st year

Table 5. Schedule of new project for 2nd year

Table 6. Schedule of new project for 3rd year

8. Discussions and Summary

In the 3rd Work Shop, we reported and discussed the country report about burn-up calculation results and application by common code, final report, evaluation of current project and next project.
The results of burn up calculation from SRAC or/and MVP codes showed very good agreement with experimental data or calculated results by other code system. MVP code was applied to investigate for neutronics calculation and utilization in some member countries like NTD, I-131 production or design new research reactor. Through the report and discussion on the final report and the evaluation of current project, it was found that all countries could enhance the neutronics calculation techniques for core management and utilization of research reactors. Moreover, “safety analysis of RIA and LOFA for research reactors” was determined as a next RRT project.
Finally, this is the last year of current project, but all members do hope that they continue the good relationship and information exchange in order to enhance more and more the neutronics calculation techniques for Research Reactors.

Attachment
A. Participant List
B. Evaluation and Review of FNCA Project Activity (Draft)

LIST OF PARTICIPANTS
THE FNCA2007 WORKSHOP ON
THE UTILIZATION OF RESEARCH REACTORS

Evaluation and Review of FNCA Project Activity (Draft)

1. Review of achievement of proect

1-1) Training on SRAC code - How many reactor experts can use SRAC code? (Optional ; How many reactor experts can use MVP code?)

As the preparation of the 1^st workshop, Japan distributed manuals and source programs of SRAC. At the 1^st workshop, Japan introduced how to install and how to execute the SRAC. In addition, Japan provided general information about SRAC and introduced procedure of JRR-3 core calculation as an example.
As the preparation of 2^nd workshop, member countries tried core calculation of their reactors by SRAC. Japan consulted them. At the 2^nd workshop, member countries reported core calculation result of their reactor by SRAC. In addition, Japan introduced the procedure of JRR-3 core burn-up calculation as an example.
As the preparation of 3^rd workshop, member countries tried core burn-up calculation and calculation for core management and enhanced utilization for their reactors. At the 3^rd workshop, Japan and member countries report results of them.
All countries have at least one person who can use SRAC. We hope the expert popularize usage of SRAC to other experts of their own countries.
Concerning MVP code, at least, one person of all member countries can use MVP code as an optional common code.

1-2) Examples of application of SRAC code for your research reactor.(Optional ; Examples of application of MVP code for your research reactor)

All member countries became able to use SRAC effectively to enhance neutronics calculation techniques.
Some member countries are trying to apply SRAC for core management. Near future, they will certainly succeed in it.
Some member countries use SRAC or MVP for enhanced utilization of research reactors. For examples, Indonesia calculates about RI production by MVP, Thailand calculates about design of in-core gemstone facility by MVP and Vietnam calculates about I-131 production from irradiated TeO₂ by MVP code.
Some member countries are using SRAC for design of new research reactor. Calculation result from SRAC is compared with the result of their domestic code, so that the reliablitity of their domestic codes is confirmed.

2. Comments and proposal for possible next phase of three years

2.-1) Specific proposal of activities to improve research reactor safety in your country, which is tentatively agreed for the project plan from 2008-10.

Many counties desire the safety analysis and thermal hydraulic analysis for the reactivity insertion accident (RIA) and the loss of flow accident (LOFA).
Japan proposes “Safety Analysis of RIA and LOFA for Research Reactors” as a new project.
For example, change of reactor power and fuel temperature with reactivity addition event caused by control rod withdrawal accident and fall-down accident of an irradiation rig is able to be evaluated at the project. Calculation codes to be used for the project are “COOLOD” and “EURECA”. These codes were developed by JAEA. The project will certainly contribute to enhance skill for safety analysis by member countries. Member countries agreed “Safety Analysis of RIA and LOFA for Research Reactors” as a new project.

2.-2) Comments on how to improve project implementation.