ZZ-FSXLIB-JD99, MCNP nuclear data library based on JENDL Dosimetry File 99

International Nuclear Information System (INIS)

Shibata, Keiichi

2007-01-01

Description: JENDL Dosimetry File 99 processed into ACE for Monte Carlo calculations. JENDL/D-99 based MCNP library. Format: ACE. Number of groups: Continuous energy cross section library. Nuclides: 47 Nuclides and 67 reactions: Li-6 (n, triton) alpha; Li-6 alpha-production; Li-7 triton- production; B-10 (n, alpha) Li-7; B-10 alpha-production; F-19 (n, 2n) F-18; Na-23 (n, 2n) Na-22; Na-23 (n, gamma) Na-24; Mg-24 (n, p) Na-24; Al-27 (n, p) Mg-27; Al-27 (n, alpha) Na-24; P-31 (n, p) Si-31; S-32 (n, p) P-32; Sc-45 (n, gamma) Sc-46; Ti-nat (n, x) Sc-46; Ti-nat (n, x) Sc-47; Ti-nat (n, x) Sc-48; Ti-46 (n, 2n) Ti-45; Ti-46 (n, p) Sc-46; Ti-47 (n, np) Sc-46; Ti-47 (n, p) Sc-47; Ti-48 (n, np) Sc-47; Ti-48 (n, p) Sc-48; Ti-49 (n, np) Sc-48; Cr-50 (n, gamma) Cr-51; Cr-52 (n, 2n) Cr-51; Mn-55 (n, 2n) Mn-54; Mn-55 (n, gamma) Mn-56; Fe-54 (n, p) Mn-54; Fe-56 (n, p) Mn-56; Fe-57 (n, np) Mn-56; Fe-58 (n, gamma) Fe-59; Co-59 (n, 2n) Co-58; Co-59 (n, gamma) Co-60; Co-59 (n, alpha) Mn-56; Ni-58 (n, 2n) Ni-57; Ni-58 (n, p) Co-58; Ni-60 (n, p) Co-60; Cu-63 (n, 2n) Cu-62; Cu-63 (n, gamma) Cu-64; Cu-63 (n, alpha) Co-60; Cu-65 (n, 2n) Cu-64; Zn-64 (n, p) Cu-64; Y-89 (n, 2n) Y-88; Zr-90 (n, 2n) Zr-89; Nb-93 (n, n') Nb-93m; Nb-93 (n, 2n) Nb-92m; Rh-103 (n, n') Rh-103m; Ag-109 (n, gamma) Ag-110m; In-115 (n, n') In-115m; In-115 (n, gamma) In-116m; I-127 (n,2n) I-126; Eu-151 (n, gamma) Eu-152; Tm-169 (n,2n) Tm-168; Ta-181 (n, gamma) Ta-182; W-186 (n, gamma) W-187; Au-197 (n, 2n) Au-196; Au-197 (n, gamma) Au-198; Hg-199 (n, n') Hg-199m; Th-232 - fission; Th-232 (n, gamma) Th-233; U-235 - fission; U-238 - fission; U-238 (n, gamma) U-239; Np-237 - fission; Pu-239 - fission; Am-241 - fission. The data were produced on the 31 of March, 2006

Integral test of JENDL fusion file

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

Integral test of JENDL Fusion File (J-FF) is performed through analyses of available benchmark experiments. As a result, good agreement between the calculated results with J-FF and the measured data is observed as a whole. Thus, J-FF is qualified to be used for nuclear design of fusion reactors. Owing to the high quality evaluation of J-FF, cross section data in J-FF for many nuclides are recommended to be assigned as data in FENDL/E-2.0 in the IAEA Consultants` Meeting held at Karlsruhe, Germany, 24-28 June, 1996. (author)

Status of JENDL High Energy File. Evaluation method, tools, specification, release procedure, etc

Energy Technology Data Exchange (ETDEWEB)

Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-11-01

The ENDF-6 format file should be kept as a standard distribution file and it is not difficult to convert into some other form for code`s libraries. From this point of view, status of JENDL High Energy File is introduced in this report as well as evaluation strategy, recommended specification, stored nuclides and quantities, a format structure, evaluation methods and tools, and release plan. (author)

Second revised edition of JENDL-3 (JENDL-3,2)

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki

1995-01-01

In Nuclear Data Center, Japan Atomic Energy Research Institute, the development of Japanese standard evaluated nuclear data library, JENDL, since the beginning of 1970s has been advanced. In June, 1994, the second revised edition of the third edition, JENDL-3,2, was opened to public. This JENDL-3,2 is expected to be utilized widely in Japan and abroad by freezing it for the time being as the final version of JENDL. This report is the contents of the publication of JENDL-3,2 of the title 'Present state of Japanese nuclear data activities' made at Physical Society of Japan in 1994. Nuclear data are divided into nuclear reaction data and nuclear structure data. The nuclear reaction data on neutron induction, evaluated nuclear data file and the international cooperation for nuclear data are described. The history of Japanese nuclear data activities, nuclear data libraries in the world and the history of JENDL are introduced. As for the evaluation of JENDL-3, its outline, light, medium and heavy nuclei, the nuclides of nuclear fission products, main actinide nuclides, transuranic elements and gamma ray generation data are reported. As for the bench mark test of JENDL-3,1, its outline, simple system bench mark, fast reactor system, thermal neutron reactor system, shield bench mark and feedback information from bench mark test are described. The evaluation and the bench mark test for LENDL-3,2 are reported. (K.I.)

JENDL FP decay data file 2000 and the beta-decay theory

International Nuclear Information System (INIS)

Yoshida, Tadashi; Katakura, Jun Ichi; Tachibana, Takahiro

2002-01-01

JENDL FP Decay Data File 2000 has been developed as one of the special purpose files of the Japanese Evaluated Nuclear Data Library (JENDL), which constitutes a versatile nuclear data basis for science and technology. In the format of ENDF-6 this file includes the decay data for 1087 unstable fission product (FP) nuclides and 142 stable nuclides as their daughters. The primary purpose of this file is to use in the summation calculation of FP decay heat, which plays a critical role in nuclear safety analysis; the loss-of-coolant accident analysis of reactors, for example. The data for a given nuclide are its decay modes, the Q value, the branching ratios, the average energies released in the form of beta- and gamma-rays per decay, and their spectral data. The primary source of the decay data adopted here is the ENSDF (Evaluated Nuclear Structure Data File). The data in ENSDF, however, cover only the measured values. The data of the short-lived nuclides, which are essential for the decay heat calculations at short cooling times, are often fully lacking or incomplete even if they exist. This is mainly because of their short half-life nature. For such nuclides a theoretical model calculation is applied in order to fill the gaps between the true and the experimentally known decay schemes. In practice we have to predict the average decay energies and the spectral data for a lot of short-lived FPs by use of beta-decay theories. Thus the beta-decay theory plays a very important role in generating the FP decay data file

Verification of dosimetry cross sections above 10 MeV based on measurement of activation reaction rates in fission neutron field

International Nuclear Information System (INIS)

Odano, Naoteru; Miura, Toshimasa; Yamaji, Akio.

1996-01-01

To validate the dosimetry cross sections in fast neutron energy range, activation reaction rates were measured for 5 types of dosimetry cross sections which have sensitivity in the energy rage above 10 MeV utilizing JRR-4 reactor of JAERI. The measured reaction rates were compared with the calculations reaction rates by a continuous energy monte carlo code MVP. The calculated reaction rates were based on two dosimetry files, JENDL Dosimetry File and IRDF-90.2. (author)

Descriptive data of JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Shibata, Keiichi; Narita, Tsutomu [eds.; Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-02-01

The second revision of JENDL-3 (JENDL-3.2) was released in 1994. The library contains evaluated neutron nuclear data for 340 nuclides which are needed for the design of fission and fusion reactors and for shielding calculation. This report presents a brief description of the evaluation method which is given in the File-1 part of JENDL-3.2. (author)

Present Status of the JENDL Project (May, 2014)

International Nuclear Information System (INIS)

Iwamoto, Osamu; Yokoyama, Kenji

2014-01-01

The Japanese Nuclear Data Committee (JNDC) is a research committee for JAEA research activities. It has two subcommittees: Subcommittee on Nuclear Data and Subcommittee on Reactor Constants. The JENDL-4.0 Updated Files (JENDL-4.0u) have been produced to take care of errors which were found after the release of JENDL-4.0. The JENDL-4.0u includes the nuclides whose nuclear data partly revised from important and/or trivial error(s). Thirteen files for JENDL-4.0 were released in 2013. Benchmarking for fission reactor applications was done using test files of Gd-157 and Am-241, and an integral testing for Th-232 was done using KUCA critical experiments. The data which are not updated in JENDL-4.0 are continuously considered to be revised, especially for FP region nuclides. Evaluated results for 141,143 Pr, 121,123-126 Sb and 96,98-106 Ru were published. The R-matrix analysis for O-16 is in progress. For decommissioning of nuclear power plants, evaluation of activation cross sections was planned for 569 reactions with 309 nuclides and is in progress. Evaluated files for the 246 isotopes from H to Hf have been created. As the activity of High Energy Nuclear Data Evaluation WG, development of new JENDL Photonuclear Data File is in progress. Evaluation for 181 nuclides has been finished. A Symposium on Nuclear Data was held on 14-15 November 2013 at Research Institute of Nuclear Engineering, University of Fukui, Tsuruga-shi, with topics about Progress in neutron cross-section measurement and analysis, Application of Nuclear Data, Recent topics, and Progress in studies of high-energy nuclear reaction. A total of 37 papers including 23 poster presentations were presented

Evaluation of fission product neutron cross sections for JENDL

International Nuclear Information System (INIS)

1984-01-01

The recent activities on the evaluation of fission product (FP) neutron cross sections for JENDL (Japanese Evaluated Nuclear Data Library) are presented briefly. The integral test of JENDL-1 FP cross section file was performed using the CFRMF sample activation data and the STEK sample reactivity data, and the ratio of experiment to calculation was nearly constant for all the samples in the STEK measurement. Therefore, a tentative analysis was performed by applying the correction to the calculated scattering reactivity component. Better agreement with the experiment was obtained after applying this correction in most cases. The evaluation work on the JENDL-2 FP neutron cross sections is now in progress. The improvement of the data evaluation is presented in an itemized form. The JENDL-2 FP file will contain the evaluated data for 100 nuclides from Kr to Tb. The improvement and the future scope of the integral test for JENDL-2 FP data are summarized. (Asami, T.)

Decay data file based on the ENSDF file

Energy Technology Data Exchange (ETDEWEB)

Katakura, J. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

A decay data file with the JENDL (Japanese Evaluated Nuclear Data Library) format based on the ENSDF (Evaluated Nuclear Structure Data File) file was produced as a tentative one of special purpose files of JENDL. The problem using the ENSDF file as primary source data of the JENDL decay data file is presented. (author)

Important comments on KERMA factors and DPA cross-section data in ACE files of JENDL-4.0, JEFF-3.2 and ENDF/B-VII.1

Science.gov (United States)

Konno, Chikara; Tada, Kenichi; Kwon, Saerom; Ohta, Masayuki; Sato, Satoshi

2017-09-01

We have studied reasons of differences of KERMA factors and DPA cross-section data among nuclear data libraries. Here the KERMA factors and DPA cross-section data included in the official ACE files of JENDL-4.0, ENDF/B-VII.1 and JEFF-3.2 are examined in more detail. As a result, it is newly found out that the KERMA factors and DPA cross-section data of a lot of nuclei are different among JENDL-4.0, ENDF/B-VII.1 and JEFF-3.2 and reasons of the differences are the followings: 1) large secondary particle production yield, 2) no secondary gamma data, 3) secondary gamma data in files12-15 mt = 3, 4) mt = 103-107 data without mt = 600 s-800 s data in file6. The issue 1) is considered to be due to nuclear data, while the issues 2)-4) seem to be due to NJOY. The ACE files of JENDL-4.0, ENDF/B-VII.1 and JEFF-3.2 with these problems should be revised after correcting wrong nuclear data and NJOY problems.

Group cross-section processing method and common nuclear group cross-section library based on JENDL-3 nuclear data file

International Nuclear Information System (INIS)

Hasegawa, Akira

1991-01-01

A common group cross-section library has been developed in JAERI. This system is called 'JSSTDL-295n-104γ (neutron:295 gamma:104) group constants library system', which is composed of a common 295n-104γ group cross-section library based on JENDL-3 nuclear data file and its utility codes. This system is applicable to fast and fusion reactors. In this paper, firstly outline of group cross-section processing adopted in Prof. GROUCH-G/B system is described in detail which is a common step for all group cross-section library generation. Next available group cross-section libraries developed in Japan based on JENDL-3 are briefly reviewed. Lastly newly developed JSSTDL library system is presented with some special attention to the JENDL-3 data. (author)

Japanese evaluated nuclear data library version 3, JENDL-3

International Nuclear Information System (INIS)

Asami, Tetsuo; Igarashi, Shun-ichi; Ihara, Hitoshi

1989-01-01

The third version of Japanese Evaluated Nuclear Data Library, JENDL-3, has recently been compiled and issued. The major features of JENDL-3 are that it covers a much larger number of nuclides than the previous versions, that evaluation is made over the all energy regions from 10 -5 eV to 20 MeV using proper nuclear theories for each of the energy regions, and that it contains data on gamma ray generation. The present report first gives an outline of JENDL-3, and then describes the evaluation of nuclear data covering light nuclides, nuclides in structural materials, fission product nuclides, major actinide nuclides, transplutonium nuclides, and gamma ray generation. The applicability of JENDL-3 is examined through a variety of bench mark tests covering the fast reactor, thermal neutron reactor, shielding, neutronics of nuclear fusion reactor, dosimetry, and gamma ray generation data. The report also describes a library for calculation of decay heat, which has close relations with JENDL-3. It has been demonstrated that the experimental data given in this library are very reliable with high reproducibility. Additional activities planned for the future are also outlined briefly. (N.K.)

RRDF-98. Russian reactor dosimetry file. Summary documentation

International Nuclear Information System (INIS)

Pashchenko, A.B.

1999-01-01

This document summarizes the contents and documentation of the new version of tile Russian Reactor Dosimetry File (RRDF-98) released in December 1998 by the Russian Center on Nuclear Data (CJD) at the Institute of Physics and Power Engineering, Russian Federation. This file contains the original evaluations of cross section data and covariance matrixes for 22 reactions which are used for neutron flux dosimetry by foil activation. The majority of the evaluations included in previous versions of the Russian Reactor Dosimetry Files (BOSPOR-80, RRGF-94 and RRDF-96) have been superseded by new evaluations. The evaluated cross sections of RRDF-98 averaged over 252-Cf and 235-U fission spectra are compared with relevant integral data. The data file is available from the IAEA Nuclear Data Section on diskette, cost free. (author)

Estimation of covariances of Cr and Ni neutron nuclear data in JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Oh, Soo Youl [Korea Atomic Energy Research Institute, Taejon (Korea)

2000-02-01

Covariances of nuclear data have been estimated for 2 nuclides contained in JENDL-3.2. The nuclides considered are Cr and Ni, which are regarded as important for the nuclear design study of fast reactors. The physical quantities for which covariances are deduced are cross sections and the first order Legendre-polynomial coefficient for the angular distribution of elastically scattered neutrons. The covariances were estimated by using the same methodology that had been used in the JENDL-3.2 evaluation in order to keep a consistency between mean values and their covariances. The least-squares fitting code GMA was used in estimating covariances for reactions of which JENDL-3.2 cross sections had been evaluated by taking account of measurements. Covariances of nuclear model calculations were deduced by using the KALMAN system. The covariance data obtained were compiled in the ENDF-6 format, and will be put into the JENDL-3.2 Covariance File which is one of JENDL special purpose files. (author)

RRDF-98. Russian reactor dosimetry file. Summary documentation

Energy Technology Data Exchange (ETDEWEB)

Pashchenko, A B

1999-03-01

This document summarizes the contents and documentation of the new version of tile Russian Reactor Dosimetry File (RRDF-98) released in December 1998 by the Russian Center on Nuclear Data (CJD) at the Institute of Physics and Power Engineering, Russian Federation. This file contains the original evaluations of cross section data and covariance matrixes for 22 reactions which are used for neutron flux dosimetry by foil activation. The majority of the evaluations included in previous versions of the Russian Reactor Dosimetry Files (BOSPOR-80, RRGF-94 and RRDF-96) have been superseded by new evaluations. The evaluated cross sections of RRDF-98 averaged over 252-Cf and 235-U fission spectra are compared with relevant integral data. The data file is available from the IAEA Nuclear Data Section on diskette, cost free. (author) 9 refs, 22 figs, 2 tabs

A study of existing experimental data and validation process for evaluated high energy nuclear data. Report of task force on integral test for JENDL High Energy File in Japanese Nuclear Data Committee

International Nuclear Information System (INIS)

Oyama, Yukio; Baba, Mamoru; Watanabe, Yukinobu

1998-11-01

JENDL High Energy File (JENDL-HE) is being produced by Japanese Nuclear Data Committee (JNDC) to provide common fundamental nuclear data in the intermediate energy region for many applications concerning a basic research, an accelerator-driven nuclear waste transmutation, a fusion material study, and medical applications like the radiation therapy. The first version of JENDL-HE, which contains the evaluated nuclear data up to 50 MeV, is planned to release in 1998. However, a method of integral test with which we can validate the high-energy nuclear data file has not been established. The validation of evaluated nuclear data through the integral tests is necessary to promote utilization of JENDL-HE. JNDC set up a task force in 1997 to discuss the problems concerning the integral tests of JENDL-HE. The task force members have surveyed and studied the current status of the problems for a year to obtain a guideline for development of the high-energy nuclear database. This report summarizes the results of the survey and study done by the task force for JNDC. (author)

Present Status of the JENDL Project (May, 2013). WPEC Meeting, Paris, France, 23-24 May, 2013

International Nuclear Information System (INIS)

Fukahori, Tokio; Iwamoto, Osamu; Ishikawa, Makoto

2013-01-01

This paper briefly presents the present status of the Japanese JENDL Project: organization of Japanese Nuclear Data Committee (JNDC), a research committee for JAEA research activities; Nuclear Data Evaluation (JENDL-4.0 Updated Files and JENDL-4.0 Plus, benchmarking for fission reactor applications (criticality for thermal systems, adjusted library for fast systems), new evaluations, covariances, evaluation tools, fission products decay data file); other activities such as the 2012 Symposium on Nuclear Data at Kyoto University Research Reactor Institute, the initiation of conversation on Covariance Data with users, the celebration of the Golden Jubilee (since 1963) of JNDC

Activity report of Japanese Nuclear Data Committee in period of April 1997 to March 1999

International Nuclear Information System (INIS)

1999-01-01

This report presents the activities of the Japanese Nuclear Data Committee (JNDC) in the period of April 1997 to March 1999. In this period, preparation of the next version of the Japanese Evaluated Nuclear Data Library, JENDL-3.3, has progressed. Generation of the special purpose files also has been advanced. In this report, the activities of each group affiliated by the committee are presented and the following are described as special topics : (1) progress in evaluating JENDL-3.3, (2) production of JENDL Covariance File and its processing system, (3) completion of the JENDL Dosimetry File, and (4) the JENDL FP Decay Data File. (author)

Final report of advisory subcommittee on development of JENDL. Guideline for developing next JENDL

International Nuclear Information System (INIS)

Chiba, Go; Iwamoto, Osamu

2015-01-01

Advisory Subcommittee on Development of JENDL was established under JENDL Committee on fiscal year 2013 to discuss the subject concerning on development of next JENDL. Members of the subcommittee were selected from developers and users of JENDL and researchers on nuclear data measurements. They discussed guidelines on next JENDL with exchanging ideas on problems of JENDL and its solution from each view point. This report describes the results of the discussions on the problems and guideline with the present status on the JENDL development. (author)

Status of the dosimetry file for ENDF/B-V

International Nuclear Information System (INIS)

Magurno, B.A.

1979-01-01

Since the new Dosimetry File will not be ready for distrubution until February-March, 1979, this presentation is submitted as a progress report where the schedule for release and dissemination of the library was reported. The dates of release and distribution were predicated on the use of pre-determined standards for use with the new Dosimetry File. Subsequently, one of the most important standards, i.e. 235 U (n,f) has been re-evaluated, causing the delay that is responsible for the above mentioned release date

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

The International Reactor Dosimetry File (IRDF-85)

International Nuclear Information System (INIS)

Cullen, D.E.; McLaughlin, P.K.

1985-04-01

This document describes the contents of the second version of the International Reactor Dosimetry File (IRDF-85), distributed by the Nuclear Data Section of the International Atomic Energy Agency. This library superseded IRDF-82. (author)

Sensitivity analysis of Monju using ERANOS with JENDL-4.0

International Nuclear Information System (INIS)

Tamagno, P.; Van Rooijen, W. F. G.; Takeda, T.; Konomura, M.

2012-01-01

This paper deals with sensitivity analysis using JENDL-4.0 nuclear data applied to the Monju reactor. In 2010 the Japan Atomic Energy Agency - JAEA - released a new set of nuclear data: JENDL-4.0. This new evaluation is expected to contain improved data on actinides and covariance matrices. Covariance matrices are a key point in quantification of uncertainties due to basic nuclear data. For sensitivity analysis, the well-established ERANOS [1] code was chosen because of its integrated modules that allow users to perform a sensitivity analysis of complex reactor geometries. A JENDL-4.0 cross-section library is not available for ERANOS. Therefore a cross-section library had to be made from the original nuclear data set, available as ENDF formatted files. This is achieved by using the following codes: NJOY, CALENDF, MERGE and GECCO in order to create a library for the ECCO cell code (part of ERANOS). In order to make sure of the accuracy of the new ECCO library, two benchmark experiments have been analyzed: the MZA and MZB cores of the MOZART program measured at the ZEBRA facility in the UK. These were chosen due to their similarity to the Monju core. Using the JENDL-4.0 ECCO library we have analyzed the criticality of Monju during the restart in 2010. We have obtained good agreement with the measured criticality. Perturbation calculations have been performed between JENDL-3.3 and JENDL-4.0 based models. The isotopes 239 Pu, 238 U, 241 Am and 241 Pu account for a major part of observed differences. (authors)

Evaluation of the excitation function for the dosimetry reaction La-139(n,g)La-140

International Nuclear Information System (INIS)

Zolotarev, K.

1997-01-01

The activation detectors on the basis of La-139(n,g)La-140 reaction are used very often in the reactor dosimetry for determination of thermal and epithermal components of neutron spectra. At present, the cross section data for this reaction are absent in the IRDF-90 ver.2 file and in the national dosimetry files such as ENDF/B-VI Dosimetry File and JENDL-3.2 Dosimetry File. The evaluation of the La-139(n,g)La-140 reaction excitation function for the Russian Reactor Dosimetry File (RRDF-96) was carried out in the energy region 1.000E-05 eV - 20 MeV. The capture cross section in the energy range from 0.00001 eV to 160 keV is given through the evaluated MLBW resolved and unresolved resonance parameters. Small background cross section was added in the resolved resonance region for taking into account the non-statistical reaction mechanism contribution. Data for En=160 keV - 1.2 MeV were obtained from evaluated average parameters in the unresolved resonance region with taking into account the competition with the neutron inelastic channels. The La-139 radiative capture cross section between 1.2 MeV and 7.6 MeV was evaluated by means of statistical analysis of the experimental data. Above 7.6 MeV the data fit was joint to the theoretical capture cross sections calculated between 7.6 and 20 MeV. Theoretical results were normalized to the evaluated value obtained from analysis experimental data at En= 14.4 - 14.8 MeV. The results of the present evaluation were compared with the ENDF/B-VI cross section data and the recommended values of resonance integral and capture cross section at En=0.0253 eV. (author)

Present status of libraries processed from JENDL-3.2

International Nuclear Information System (INIS)

Yamano, Naoki

1996-01-01

Data libraries processed from JENDL-3.2 were produced by JAERI for use of typical applications in the area of nuclear reactor and shielding. These libraries had the distinction of being free of restrictions on distribution. The other data libraries for decay, activation, dosimetry, nuclide transmutation, burn-up, PKA, KERMA and DPA calculations are now being accumulated. The current status on the development of these libraries has been described. A discussion is presented about how to disseminate and share the products of JNDC activities. (author)

Sensitivity analysis of Monju using ERANOS with JENDL-4.0

Energy Technology Data Exchange (ETDEWEB)

Tamagno, P. [Institut National des Sciences et Techniques Nucleaires, INSTN - Point Courrier no 35, Centre CEA de Saclay, F-91191 Gif-sur-Yvette Cedex (France); Van Rooijen, W. F. G.; Takeda, T. [Research Inst. of Nuclear Engineering, Univ. of Fukui, Kanawa-cho 1-2-4, T914-0055 Fukui-ken, Tsuruga-shi (Japan); Konomura, M. [Japan Atomic Energy Agency, FBR Plant Engineering Center, Shiraki 1, 919-1279 Fukui-ken, Tsuruga-shi (Japan)

2012-07-01

This paper deals with sensitivity analysis using JENDL-4.0 nuclear data applied to the Monju reactor. In 2010 the Japan Atomic Energy Agency - JAEA - released a new set of nuclear data: JENDL-4.0. This new evaluation is expected to contain improved data on actinides and covariance matrices. Covariance matrices are a key point in quantification of uncertainties due to basic nuclear data. For sensitivity analysis, the well-established ERANOS [1] code was chosen because of its integrated modules that allow users to perform a sensitivity analysis of complex reactor geometries. A JENDL-4.0 cross-section library is not available for ERANOS. Therefore a cross-section library had to be made from the original nuclear data set, available as ENDF formatted files. This is achieved by using the following codes: NJOY, CALENDF, MERGE and GECCO in order to create a library for the ECCO cell code (part of ERANOS). In order to make sure of the accuracy of the new ECCO library, two benchmark experiments have been analyzed: the MZA and MZB cores of the MOZART program measured at the ZEBRA facility in the UK. These were chosen due to their similarity to the Monju core. Using the JENDL-4.0 ECCO library we have analyzed the criticality of Monju during the restart in 2010. We have obtained good agreement with the measured criticality. Perturbation calculations have been performed between JENDL-3.3 and JENDL-4.0 based models. The isotopes {sup 239}Pu, {sup 238}U, {sup 241}Am and {sup 241}Pu account for a major part of observed differences. (authors)

Final report of advisory subcommittee on development of JENDL. Guideline for developing next JENDL

International Nuclear Information System (INIS)

Chiba, Go

2016-01-01

An advisory subcommittee on development of JENDL was established in 2013 to discuss future development of JENDL among experts on nuclear data experiments, evaluations and applications. This paper is a summary of a final report which was submitted to the JENDL committee on March in 2014. (author)

Present status of JENDL-4

International Nuclear Information System (INIS)

Shibata, Keiichi

2009-01-01

The fourth version of Japanese Evaluated Nuclear Data Library is being developed at the JAEA Nuclear Data Center in cooperation with the Japanese Nuclear Data Committee. As for actinides, we already released JENDL Actinoid File 2008, which contains the evaluated data for 79 nuclei. The high-energy cross sections of FP nuclei have been evaluated by using the CCONE and POD statistical model codes. The nuclear data for structural materials and light nuclei are being revised. The fission product yields were updated on the basis of ENDF/B-VII.0. Ternary fission was included in the yield data. (author)

ENDF/B-5 Dosimetry Files, mod. 2 1979/81

International Nuclear Information System (INIS)

DayDay, N.; Lemmel, H.D.

1981-09-01

This document summarizes the contents and documentation of the ENDF/B-5 Dosimetry Files (Point or Group Data) released in October 1979 and modified in August 1981. The files contain data for 36 neutron reactions of 26 isotopes. The entire libraries or selective retrievals from them can be obtained free of charge from the IAEA Nuclear Data Section. (author)

Status of thorium cycle nuclear data evaluations: Comparison of cross-section line shapes of JENDL-3 and ENDF-B-VI files for 230Th, 232Th, 231Pa, 233Pa, 232U, 233U and 234U

International Nuclear Information System (INIS)

Ganesan, S.; McLaughlin, P.K.

1992-02-01

Since 1990, one of the most interesting developments in the field of nuclear data for nuclear technology applications is that several new evaluated data files have been finalized and made available to the International Atomic Energy Agency (IAEA) for distribution to its Member States. Improved evaluated nuclear data libraries such as ENDF/B-VI from the United States and JENDL-3 from Japan were developed over a period of 10-15 years. This report is not an evaluation of the evaluations. The report as presented here gives a first look at the cross section line shapes of the isotopes that are important to the thorium fuel cycle derived from the two recently evaluated data files: JENDL-3 and ENDF/B-VI. The basic evaluated data files JENDL-3 and ENDF/B-VI were point-processed successfully using the codes LINEAR and RECENT. The point data were multigrouped in three different group structures using the GROUPIE code. Graphs of intercomparisons of cross section line shapes of JENDL-3 and ENDF/B-VI are presented in this paper for the following isotopes of major interest to studies of the thorium fuel cycle: 230 Th, 232 Th, 231 Pa, 233 Pa, 232 U, 233 U and 234 U. Comparisons between JENDL-3 and ENDF/B-VI which were performed at the point and group levels show large discrepancies in various cross sections. We conclude this report with a general remark that it is necessary to perform sensitivity studies to assess the impacts of the discrepancies between the two different sets of data on calculated reactor design and safety parameters of specific reactor systems and, based on the results of such sensitivity studies, to undertake new tasks of evaluations. (author). 2 refs, 245 figs, 8 tabs

Present status of JENDL-3.3

International Nuclear Information System (INIS)

Nakagawa, Tsuneo; Shibata, Keiichi

2001-01-01

The latest version of JENDL General Purpose Library, JENDL-3.3, is being compiled. The evaluations for JENDL-3.3 have been performed mainly by the Medium-Heavy Nuclide Data Evaluation WG, the Heavy Nuclide Data Evaluation WG and the Delayed Neutron WG in the Japanese Nuclear Data Committee. The main objective of the evaluations is to solve the problems that were pointed out in JENDL-3.2. (author)

Estimation of covariances of 16O, 23Na, Fe, 235U, 238U and 239Pu neutron nuclear data in JENDL-3.2

International Nuclear Information System (INIS)

Shibata, Keiichi; Nakajima, Yutaka; Kawano, Toshihiko; Oh, Soo-Youl; Matsunobu, Hiroyuki; Murata, Toru.

1997-10-01

Covariances of nuclear data have been estimated for 6 nuclides contained in JENDL-3.2. The nuclides considered are 16 O, 23 Na, Fe, 235 U, 238 U, and 239 Pu, which are regarded as important for the nuclear design study of fast reactors. The physical quantities for which covariances are deduced are cross sections, resolved and unresolved resonance parameters, and the first order Legendre-polynomial coefficient for the angular distribution of elastically scattered neutrons. As for 235 U, covariances were obtained also for the average number of neutrons emitted in fission. The covariances were estimated by using the same methodology that had been used in the JENDL-3.2 evaluation in order to keep a consistency between mean values and their covariances. The least-squares fitting code GMA was used in estimating covariances for reactions of which JENDL-3.2 cross sections had been evaluated by taking account of measurements. In nuclear model calculations, the covariances were calculated by the KALMAN system. The covariance data obtained were compiled in the ENDF-6 format, and will be put into the JENDL-3.2 Covariance File which is one of JENDL special purpose files. (author). 193 refs

Descriptive data of JENDL-3.3

Energy Technology Data Exchange (ETDEWEB)

Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-01-01

The third revision of JENDL-3 (JENDL-3.3) was released in 2002. The library contains evaluated neutron nuclear data for 337 nuclides. This report presents a brief description of the evaluation method which is given in the MF1 part of JENDL-3.3. (author)

Current status of JENDL-3.3

Energy Technology Data Exchange (ETDEWEB)

Hasegawa, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-03-01

Current status of JENDL-3.3 is presented. Reevaluation work toward JENDL-3.3 has started last April for three years project to supply a consolidated new versions of JENDL by JAERI NDC (Nuclear Data center) with the cooperation of JNDC (Japanese Nuclear Data Committee). The working schedule has been fixed by the careful review of the summary report, `The problems of JENDL-3.2`, submitted to JNDC last March after one year discussions by a small advisory group: `Identifying the problems of JENDL-3.2`. To cope with the problems, two new subgroups are set up in the Subcommittee of Nuclear Data of JNDC. One is Heavy Mass Elements Evaluation Working Group for the re-evaluation of major actinides (Th-232, U-233,235,236,238, Pu-236,239,241,242). The other is Intermediate Mass Elements Evaluation Working Group for solving the inconsistencies between calculations and integral experiments relating to the fields of fusion neutronics and shielding applications as well as new evaluations such as Er elements. Supplying covariance data for important nuclides are one of the main feature of JENDL-3.3. Re-evaluated data will be released as JENDL-3.3 in the individual bases after the reviewing process by the experts. (author)

Usage of nuclear data library and proposal to JENDL

International Nuclear Information System (INIS)

Sasaki, M.; Iijima, S.; Kawai, Y.; Matsunobu, H.; Zukeran, A.

1987-01-01

A usage status of nuclear data libraries and proposals to the JENDL library are briefly reviewed. 1) Data bases used in plant design and fuel cycle design fields are summarized. JNDC evaluated data and JENDL are not utilized so much than expected in these fields. 2) Performances between JENDL-2 and ENDF/B-IV are compared in the FBR nuclear design field. JENDL-2 shows good characteristics. 3) Usages of JENDL-2 are shown in the LWR nuclear design field. JENDL-2 isn't always superior to the licensing library. 4) Usages of JENDL-2 are shown in the FBR shielding R D field. JENDL-2 underestimates measured spectra in case of the deep penetration experiment. 5) Based on the above usage status of JENDL-2, proposals to future JENDL and Nuclear Data Center are tabulated. (author)

Status of neutron dosimetry cross sections

International Nuclear Information System (INIS)

Griffin, P.J.; Kelly, J.G.

1992-01-01

Several new cross section libraries, such as ENDF/B-VI(release 2), IRDF-90,JEF-2.2, and JENDL-3 Dosimetry, have recently been made available to the dosimetry community. the Sandia National Laboratories (SNL) Radiation Metrology Laboratory (RML) has worked with these libraries since pre-release versions were available. this paper summarizes the results of the intercomparison and testing of dosimetry cross sections. As a result of this analysis, a compendium of the best dosimetry cross sections was assembled from the available libraries for use within the SNL RML. this library, referred to as the SNLRML Library, contains 66 general dosimetry sensors and 3 special dosimeters unique to the RML sensor inventory. The SNLRML cross sections have been put into a format compatible with commonly used spectrum determination codes

Evaluation of 54Fe(n,2n)53m+gFe reaction cross sections for high energy dosimetry applications

International Nuclear Information System (INIS)

Zolotarev, K.I.; Pashchenko, A.B.

2001-01-01

The new evaluation of excitation function for the high energy threshold 54 Fe(n,2n) 53m+g Fe dosimetry reaction in the energy range from the threshold to 20 MeV is briefly described. The cross section uncertainties and the covariance matrix were estimated simultaneously from the analysis. The adopted curve is compared to the available processed experimental data and the existing FEI-93, ENDF/B-VI and JENDL-3.2 evaluations. The ENDF-6 formatted data file is available from the Web site of the Russian Nuclear Data Center (RNDC) online (http://www.rndc.ippe.obninsk.ru). (author)

Validation of JENDL-3.3 by criticality benchmark testing

International Nuclear Information System (INIS)

Takano, Hideki; Nakagawa, Tsuneo

2001-01-01

In the thermal uranium core, the keff-values of STACY, TRACY and JRR-4 overestimated with JENDL-3.2 were improved significantly by decreasing of about 0.6% with JENDL-3.3. This is due to modification of the fission spectrum and thermal fission cross section of 235 U from JENDL-3.2 to JENDL-3.3 data. For the uranium fast cores, the discrepancies of keff values between JENDL-3.2 and 3.3 were very small. In the thermal Pu cores of TCA, the keff-values calculated with JENDL-3.3 were in good agreement with the experimental values. For Pu fuel cores of ZPPR-9 and FCA-XVII, the keff values calculated with JENDL-3.3 became larger 0.2% than those for JENDL-3.2. In small fast cores with U-233 fuel, the keff-values overestimated with JENDL-3.2 were improved considerably with JENDL-3.3, due to reevaluation of U-233 fission cross sections in the high energy region. (author)

Shielding benchmark tests of JENDL-3

International Nuclear Information System (INIS)

Kawai, Masayoshi; Hasegawa, Akira; Ueki, Kohtaro; Yamano, Naoki; Sasaki, Kenji; Matsumoto, Yoshihiro; Takemura, Morio; Ohtani, Nobuo; Sakurai, Kiyoshi.

1994-03-01

The integral test of neutron cross sections for major shielding materials in JENDL-3 has been performed by analyzing various shielding benchmark experiments. For the fission-like neutron source problem, the following experiments are analyzed: (1) ORNL Broomstick experiments for oxygen, iron and sodium, (2) ASPIS deep penetration experiments for iron, (3) ORNL neutron transmission experiments for iron, stainless steel, sodium and graphite, (4) KfK leakage spectrum measurements from iron spheres, (5) RPI angular neutron spectrum measurements in a graphite block. For D-T neutron source problem, the following two experiments are analyzed: (6) LLNL leakage spectrum measurements from spheres of iron and graphite, and (7) JAERI-FNS angular neutron spectrum measurements on beryllium and graphite slabs. Analyses have been performed using the radiation transport codes: ANISN(1D Sn), DIAC(1D Sn), DOT3.5(2D Sn) and MCNP(3D point Monte Carlo). The group cross sections for Sn transport calculations are generated with the code systems PROF-GROUCH-G/B and RADHEAT-V4. The point-wise cross sections for MCNP are produced with NJOY. For comparison, the analyses with JENDL-2 and ENDF/B-IV have been also carried out. The calculations using JENDL-3 show overall agreement with the experimental data as well as those with ENDF/B-IV. Particularly, JENDL-3 gives better results than JENDL-2 and ENDF/B-IV for sodium. It has been concluded that JENDL-3 is very applicable for fission and fusion reactor shielding analyses. (author)

Critical and subcritical mass calculations of fissionable nuclides based on JENDL-3.2+

International Nuclear Information System (INIS)

Okuno, H.

2002-01-01

We calculated critical and subcritical masses of 10 fissionable actinides ( 233 U, 235 U, 238 Pu, 239 Pu, 241 Pu, 242m Am, 243 Cm, 244 Cm, 249 Cf and 251 Cf) in metal and in metal-water mixtures (except 238 Pu and 244 Cm). The calculation was made with a combination of a continuous energy Monte Carlo neutron transport code, MCNP-4B2, and the latest released version of the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI, JEF-2.2, and JENDL-3.3 in its preliminary version were also applied to find differences in results originated from different nuclear data files. For the so-called big three fissiles ( 233 U, 235 U and 239 Pu), analyzing the criticality experiments cited in ICSBEP Handbook validated the code-library combination, and calculation errors were consequently evaluated. Estimated critical and lower limit critical masses of the big three in a sphere with/without a water or SS-304 reflector were supplied, and they were compared with the subcritical mass limits of ANS-8.1. (author)

Estimation of covariances of {sup 16}O, {sup 23}Na, Fe, {sup 235}U, {sup 238}U and {sup 239}Pu neutron nuclear data in JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakajima, Yutaka; Kawano, Toshihiko; Oh, Soo-Youl; Matsunobu, Hiroyuki; Murata, Toru

1997-10-01

Covariances of nuclear data have been estimated for 6 nuclides contained in JENDL-3.2. The nuclides considered are {sup 16}O, {sup 23}Na, Fe, {sup 235}U, {sup 238}U, and {sup 239}Pu, which are regarded as important for the nuclear design study of fast reactors. The physical quantities for which covariances are deduced are cross sections, resolved and unresolved resonance parameters, and the first order Legendre-polynomial coefficient for the angular distribution of elastically scattered neutrons. As for {sup 235}U, covariances were obtained also for the average number of neutrons emitted in fission. The covariances were estimated by using the same methodology that had been used in the JENDL-3.2 evaluation in order to keep a consistency between mean values and their covariances. The least-squares fitting code GMA was used in estimating covariances for reactions of which JENDL-3.2 cross sections had been evaluated by taking account of measurements. In nuclear model calculations, the covariances were calculated by the KALMAN system. The covariance data obtained were compiled in the ENDF-6 format, and will be put into the JENDL-3.2 Covariance File which is one of JENDL special purpose files. (author). 193 refs.

Status of LWR fuel design and future usage of JENDL

International Nuclear Information System (INIS)

Ito, Takuya

2008-01-01

For all conventional LWR fuel design codes of LWR fuel manufactures in Japan, the cross section library are based on the ENDF/B. Recently we can see several movements for the utilization of JENDL library for the LWR fuel design. The latest version of NEUPHYS cross section library is based on the JENDL-3.2. To accelerate this movement of JENDL utilization in LWR fuel design, it is necessary to prepare a high quality JENDL document, systematic validation of JENDL and to appeal them abroad effectively. (author)

Review of benchmark tests on JENDL-2B library

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki

1982-01-01

JENDL-2B is a mixed library consisting of JENDL-2 for the most important nuclides, i.e., 235 U, 238 U, 239 Pu, 240 Pu, 241 Pu, Cr, Fe and Ni and of JENDL-1 for other nuclides. Complete reevaluation work was made for these eight nuclides. The simultaneous evaluation method was adopted in the evaluation for the five heavy nuclides. The resonance structure was carefully studied for the structural materials in the unresolved resonance region up to several MeV. Benchmark tests have been made on JENDL-2B. Various core center characteristics were tested with one-dimensional model for total of 27 assemblies. Satisfactory results were obtained as a whole. The results of spectrum indices, however, suggested some inconsistent spectrum prediction. Moreover, the reactivity worths were overestimated for most of materials, and apparent C/E discrepancies were observed between the Pu and U cores. Applicability of JENDL-2B was further tested to more sophisticated problems for MOZART and ZPPR-3 assemblies. The reaction rate distributions were better predicted with JENDL-2B than with JENDL-1. The positive sodium void reactivity worth was much overestimated with JENDL-2B due to too large moderation components. The control rod worths were well predicted in MOZART, but were considerably underpredicted in ZPPR-3. (author)

Integral benchmark test of JENDL-4.0 for U-233 systems with ICSBEP handbook

International Nuclear Information System (INIS)

Kuwagaki, Kazuki; Nagaya, Yasunobu

2017-03-01

The integral benchmark test of JENDL-4.0 for U-233 systems using the continuous-energy Monte Carlo code MVP was conducted. The previous benchmark test was performed only for U-233 thermal solution and fast metallic systems in the ICSBEP handbook. In this study, MVP input files were prepared for uninvestigated benchmark problems in the handbook including compound thermal systems (mainly lattice systems) and integral benchmark test was performed. The prediction accuracy of JENDL-4.0 was evaluated for effective multiplication factors (k eff 's) of the U-233 systems. As a result, a trend of underestimation was observed for all the categories of U-233 systems. In the benchmark test of ENDF/B-VII.1 for U-233 systems with the ICSBEP handbook, it is reported that a decreasing trend of calculated k eff values in association with a parameter ATFF (Above-Thermal Fission Fraction) is observed. The ATFF values were also calculated in this benchmark test of JENDL-4.0 and the same trend as ENDF/B-VII.1 was observed. A CD-ROM is attached as an appendix. (J.P.N.)

Evaluation of covariance data for chromium, iron and nickel contained in JENDL-3.2

International Nuclear Information System (INIS)

Oh, Soo-Youl; Shibata, Keiichi.

1998-01-01

An evaluation has been made for the covariances of neutron cross sections of 52 Cr, 56 Fe, 58 Ni and 60 Ni contained in JENDL-3.2. Reactions considered were the threshold reactions such as (n, 2n), (n, nα), (n, np), (n, p), (n, d), (n, t) and (n, α), the radiative capture reaction above the resonance region, and the inelastic scattering to discrete and continuum levels. Evaluation guidelines and procedures were established during the work. A generalized least-squares fitting code GMA was used in estimating covariances for reactions of which JENDL-3.2 cross sections had been evaluated by taking account of many measured data. For cross sections that had been evaluated by nuclear reaction model calculations, the KALMAN code, which yields covariances of cross sections and of associated model parameters on the basis of the Bayesian statistics, was used in conjunction with reaction model codes EGNASH and CASTHY. The evaluated uncertainties of a few percent to 30% in the cross sections look reasonable, and the correlation matrices show understandable trends. Even though there is no strict way to confirm the validity of the evaluated covariances, tools and procedures adopted in the present work are appropriate for producing covariance files based on JENDL-3.2. The covariances obtained will be compiled into JENDL in the near future. Meanwhile, new sets of optical model and level density parameters were proposed as one of byproducts obtained from the KALMAN calculations. (author)

ERRORJ. Covariance processing code system for JENDL. Version 2

International Nuclear Information System (INIS)

Chiba, Gou

2003-09-01

ERRORJ is the covariance processing code system for Japanese Evaluated Nuclear Data Library (JENDL) that can produce group-averaged covariance data to apply it to the uncertainty analysis of nuclear characteristics. ERRORJ can treat the covariance data for cross sections including resonance parameters as well as angular distributions and energy distributions of secondary neutrons which could not be dealt with by former covariance processing codes. In addition, ERRORJ can treat various forms of multi-group cross section and produce multi-group covariance file with various formats. This document describes an outline of ERRORJ and how to use it. (author)

New nuclear data group constant sets for fusion reactor nuclear analyses based on JENDL-4.0 and FENDL-3.0

International Nuclear Information System (INIS)

Konno, Chikara; Ohta, Masayuki; Kwon, Saerom; Ochiai, Kentaro; Sato, Satoshi

2015-01-01

We have produced new nuclear data group constant sets from JENDL-4.0 and FENDL-3.0 for fusion reactor nuclear analyses; FUSION-J40-175, FUSION-F30-175 (40 materials, neutron 175 groups, gamma 42 groups), FUSION-J40-42 and FUSION-F30-42 (40 materials, neutron 42 groups, gamma 21 groups). MATXS files of JENDL-4.0 and FENDL-3.0 were newly produced with the NJOY2012 code. FUSION-J40-175, FUSION-J40-42, FUSION-F30-175 and FUSION-F30-42 were produced with the TRANSX code. KERMA factors, DPA and gas production cross-section data were also prepared from the MATXS files with TRANSX. Test calculations were carried out in order to validate these nuclear group constant sets. They suggested that these group constant sets had no problem. (author)

Critical and subcritical mass calculations of curium-243 to -247 based on JENDL-3.2 for revision of ANSI/ANS-8.15

International Nuclear Information System (INIS)

Okuno, Hiroshi

2002-01-01

Critical and subcritical masses were calculated for a sphere of five curium isotopes from 243 Cm to 247 Cm in metal and in metal-water mixtures considering three reflector conditions: bare, with a water reflector or a stainless steel reflector. The calculation were made mainly with a combination of a continuous energy Monte Carlo neutron transport calculation code, MCNP, and the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI and JEF-2.2, were also applied to find differences in calculation results of the neutron multiplication factor originated from different nuclear data files. A large dependence on the evaluated nuclear data files was found in the calculation results: more than 10%Δk/k relative differences in the neutron multiplication factor for a homogeneous mixture of 243 Cm metal and water when JENDL-3.2 was replaced with ENDF/B-VI and JEF-2.2, respectively; and a 44% reduction in the critical mass by changing from JENDL-3.2 to ENDF/B-VI for 246 Cm metal. The present study supplied basic information to the ANSI/ANS-8.15 Working Group for revision of the standard for nuclear criticality control of special actinide elements. The new or revised values of the subcritical mass limits for curium isotopes accepted by the ANSI/ANS-8.15 Working Group were finally summarized. (author)

JENDL-4.0 benchmarking for fission reactor applications

International Nuclear Information System (INIS)

Chiba, Go; Okumura, Keisuke; Sugino, Kazuteru; Nagaya, Yasunobu; Yokoyama, Kenji; Kugo, Teruhiko; Ishikawa, Makoto; Okajima, Shigeaki

2011-01-01

Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0. (author)

Jendl-3.1 iron validation on the PCA-REPLICA (H2O/Fe) shielding benchmark experiment

International Nuclear Information System (INIS)

Pescarini, M.; Borgia, M. G.

1997-03-01

The PCA-REPLICA (H 2 O/Fe) neutron shielding benchmarks experiment is analysed using the SN 2-D DOT 3.5-E code and the 3-D-equivalent flux synthesis method. This engineering benchmark reproduces the ex-core radial geometry of a PWR, including a mild steel reactor pressure vessel (RPV) simulator, and is designed to test the accuracy of the calculation of the in-vessel neutron exposure parameters. This accuracy is strongly dependent on the quality of the iron neutron cross sections used to describe the nuclear reactions within the RPV simulator. In particular, in this report, the cross sections based on the JENDL-3.1 iron data files are tested, through a comparison of the calculated integral and spectral results with the corresponding experimental data. In addition, the present results are compared, on the same benchmark experiment, with those of a preceding ENEA-Bologna validation of the ENDF/B VI iron cross sections. The integral result comparison indicates that, for all the threshold detectors considered (Rh-103 (n, n') Rh-103m, In-115 (n, n') In-115m and S-32 (n, p) P-32), the JENDL-3.1 natural iron data produce satisfactory results similar to those obtained with the ENDF/B VI iron data. On the contrary, when the JENDL/3.1 Fe-56 data file is used, strongly underestimated results are obtained for the lower energy threshold detectors, Rh-103 and In-115. This fact, in particular, becomes more evident with increasing the neutron penetration depth in the RPV simulator

Benchmark tests of JENDL-1

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki; Hasegawa, Akira; Takano, Hideki; Kamei, Takanobu; Hojuyama, Takeshi; Sasaki, Makoto; Seki, Yuji; Zukeran, Atsushi; Otake, Iwao.

1982-02-01

Various benchmark tests were made on JENDL-1. At the first stage, various core center characteristics were tested for many critical assemblies with one-dimensional model. At the second stage, applicability of JENDL-1 was further tested to more sophisticated problems for MOZART and ZPPR-3 assemblies with two-dimensional model. It was proved that JENDL-1 predicted various quantities of fast reactors satisfactorily as a whole. However, the following problems were pointed out: 1) There exists discrepancy of 0.9% in the k sub(eff)-values between the Pu- and U-cores. 2) The fission rate ratio of 239 Pu to 235 U is underestimated by 3%. 3) The Doppler reactivity coefficients are overestimated by about 10%. 4) The control rod worths are underestimated by 4%. 5) The fission rates of 235 U and 239 Pu are underestimated considerably in the outer core and radial blanket regions. 6) The negative sodium void reactivities are overestimated, when the sodium is removed from the outer core. As a whole, most of problems of JENDL-1 seem to be related with the neutron leakage and the neutron spectrum. It was found through the further study that most of these problems came from too small diffusion coefficients and too large elastic removal cross sections above 100 keV, which might be probably caused by overestimation of the total and elastic scattering cross sections for structural materials in the unresolved resonance region up to several MeV. (author)

Critical and subcritical mass calculations of curium-243 to -247 based on JENDL-3.2 for revision of ANSI/ANS-8.15

Energy Technology Data Exchange (ETDEWEB)

Okuno, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kawasaki, Hiromitsu [CRC Solutions Corporation, Hitachinaka, Ibaraki (Japan)

2002-10-01

Critical and subcritical masses were calculated for a sphere of five curium isotopes from {sup 243}Cm to {sup 247}Cm in metal and in metal-water mixtures considering three reflector conditions: bare, with a water reflector or a stainless steel reflector. The calculation were made mainly with a combination of a continuous energy Monte Carlo neutron transport calculation code, MCNP, and the Japanese Evaluated Nuclear Data Library, JENDL-3.2. Other evaluated nuclear data files, ENDF/B-VI and JEF-2.2, were also applied to find differences in calculation results of the neutron multiplication factor originated from different nuclear data files. A large dependence on the evaluated nuclear data files was found in the calculation results: more than 10%{delta}k/k relative differences in the neutron multiplication factor for a homogeneous mixture of {sup 243}Cm metal and water when JENDL-3.2 was replaced with ENDF/B-VI and JEF-2.2, respectively; and a 44% reduction in the critical mass by changing from JENDL-3.2 to ENDF/B-VI for {sup 246}Cm metal. The present study supplied basic information to the ANSI/ANS-8.15 Working Group for revision of the standard for nuclear criticality control of special actinide elements. The new or revised values of the subcritical mass limits for curium isotopes accepted by the ANSI/ANS-8.15 Working Group were finally summarized. (author)

Present status and plans of JENDL FP data evaluation project

International Nuclear Information System (INIS)

Kawano, Toshihiko

2003-01-01

The Fission Product (FP) data evaluation is one of the key issues for the next revision of JENDL, since many data were carried over from JENDL-3.2 to 3.3. The new FP Working Group (FPWG) was organized in the Japanese Nuclear Data Committee (JNDC) to perform new evaluations of FP data for JENDL. In the present review of FP data evaluation, we show the history and current status of FP nuclear data, requirements of FP data, new measurements, the international cooperation, and the future plans toward the new JENDL. (author)

Detailed benchmark test of JENDL-4.0 iron data for fusion applications

Energy Technology Data Exchange (ETDEWEB)

Konno, Chikara, E-mail: konno.chikara@jaea.go.jp [Japan Atomic Energy Agency, Tokai-Mura, Ibaraki-ken, 319-1195 (Japan); Wada, Masayuki [Japan Computer System, Mito, 310-0805 (Japan); Kondo, Keitaro; Ohnishi, Seiki; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi [Japan Atomic Energy Agency, Tokai-Mura, Ibaraki-ken, 319-1195 (Japan)

2011-10-15

The major revised version of Japanese Evaluated Nuclear Data Library (JENDL), JENDL-4.0, was released in May, 2010. As one of benchmark tests, we have carried out a benchmark test of JENDL-4.0 iron data, which are very important for radiation shielding in fusion reactors, by analyzing the iron fusion neutronics integral experiments (in situ and Time-of-Flight (TOF) experiments) at JAEA/FNS. It is demonstrated that the problems of the iron data in the previous version of JENDL, JENDL-3.3, are solved in JENDL-4.0; the first inelastic scattering cross section data of {sup 57}Fe and the angular distribution of the elastic scattering of {sup 56}Fe. The iron data in JENDL-4.0 are comparable to or are partly better than those in ENDF/B-VII.0 and JEFF-3.1.

Self-shielding effect in unresolved resonance data in JENDL-4.0

International Nuclear Information System (INIS)

Konno, Chikara; Takakura, Kosuke; Ochiai, Kentaro; Sato, Satoshi; Kato, Yoshinari

2012-01-01

At International Conference on Nuclear Data for Science and Technology in 2007 we pointed out that most of unresolved resonance data in JENDL-3.3 have a problem related to self-shielding correction. Here with a simple calculation model we have investigated whether the latest JENDL, JENDL-4.0, was improved for the problem or not. The results suggest that unresolved resonance data in JENDL-4.0 have no problem, but it seems that self-shielding effects for the unresolved resonance data in JENDL-4.0 are too large. New benchmark experiments for unresolved resonance data are strongly recommended in order to verify unresolved resonance data. (author)

Validation of JENDL-3.3 for the HTTR criticality

International Nuclear Information System (INIS)

Goto, Minoru; Nojiri, Naoki; Shimakawa, Satoshi

2004-01-01

Validation of JENDL-3.3 has been performed for the HTTR criticality using the MVP code with a ''lattice-cell'' of infinite models and a ''whole-core'' of finite models. It was found that the keff values calculated with JENDL-3.3 was decreased about 0.2-0.4%Δk from one with JENDL-3.2. The criticality prediction was closed to the experimental data in the critical approach situation of the HTTR. (author)

Summary report of the technical meeting on 'International Reactor Dosimetry File: IRDF-2002'

International Nuclear Information System (INIS)

Greenwood, L.R.; Paviotti-Corcuera, R.

2002-09-01

This report summarizes the presentations, recommendations and conclusions of the Technical Meeting on 'International Reactor Dosimetry File: IRDF-2002.' The purpose of this meeting was to discuss scientific and technical matters related to the subject and coordinate related tasks. Discussions were held and recommendations were given for the preparation of the files on topics related to: reactions to be included, need for new evaluations or revisions, decay data, radiation damage data, integral testing in benchmark fields, and computer codes to be included. Tasks were assigned and deadlines were set. The participants emphasized that accurate and complete knowledge of nuclear data for reactor dosimetry are essential for improving the accuracy of the reactor pressure vessel service life assessment of nuclear power plants as well as in other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. (author)

Jendl-3.1 iron validation on the PCA-REPLICA (H{sub 2}O/Fe) shielding benchmark experiment

Energy Technology Data Exchange (ETDEWEB)

Pescarini, M.; Borgia, M. G. [ENEA, Centro Ricerche ``Ezio Clementel``, Bologna (Italy). Dipt. Energia

1997-03-01

The PCA-REPLICA (H{sub 2}O/Fe) neutron shielding benchmarks experiment is analysed using the SN 2-D DOT 3.5-E code and the 3-D-equivalent flux synthesis method. This engineering benchmark reproduces the ex-core radial geometry of a PWR, including a mild steel reactor pressure vessel (RPV) simulator, and is designed to test the accuracy of the calculation of the in-vessel neutron exposure parameters. This accuracy is strongly dependent on the quality of the iron neutron cross sections used to describe the nuclear reactions within the RPV simulator. In particular, in this report, the cross sections based on the JENDL-3.1 iron data files are tested, through a comparison of the calculated integral and spectral results with the corresponding experimental data. In addition, the present results are compared, on the same benchmark experiment, with those of a preceding ENEA-Bologna validation of the ENDF/B VI iron cross sections. The integral result comparison indicates that, for all the threshold detectors considered (Rh-103 (n, n`) Rh-103m, In-115 (n, n`) In-115m and S-32 (n, p) P-32), the JENDL-3.1 natural iron data produce satisfactory results similar to those obtained with the ENDF/B VI iron data. On the contrary, when the JENDL/3.1 Fe-56 data file is used, strongly underestimated results are obtained for the lower energy threshold detectors, Rh-103 and In-115. This fact, in particular, becomes more evident with increasing the neutron penetration depth in the RPV simulator.

Status of data testing of ENDF/B-V reactor dosimetry file

International Nuclear Information System (INIS)

Magurno, B.A.

1979-01-01

The ENDF/B-V Reactor Dosimetry File was released August 1979, and Phase II data testing started. The results presented here are from Brookhaven National Laboratory only, and are considered preliminary. The tests include calculated spectrum-averaged cross sections using 235 U fission spectrum (Watt), 252 Cf spontaneous fission spectrum (Watt and Maxwellian), and the Coupled Fast Reactor Measurement Facility (CFRMF) spectrum. 6 tables

Integral test of JENDL-3.3 for thermal reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Mori, Takamasa

2003-01-01

Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

Nuclear decay data files of the Dosimetry Research Group

International Nuclear Information System (INIS)

Eckerman, K.F.; Westfall, R.J.; Ryman, J.C.; Cristy, M.

1993-12-01

This report documents the nuclear decay data files used by the Dosimetry Research Group at Oak Ridge National Laboratory and the utility DEXRAX which provides access to the files. The files are accessed, by nuclide, to extract information on the intensities and energies of the radiations associated with spontaneous nuclear transformation of the radionuclides. In addition, beta spectral data are available for all beta-emitting nuclides. Two collections of nuclear decay data are discussed. The larger collection contains data for 838 radionuclides, which includes the 825 radionuclides assembled during the preparation of Publications 30 and 38 of the International Commission on Radiological Protection (ICRP) and 13 additional nuclides evaluated in preparing a monograph for the Medical Internal Radiation Dose (MIRD) Committee of the Society of Nuclear Medicine. The second collection is composed of data from the MIRD monograph and contains information for 242 radionuclides. Abridged tabulations of these data have been published by the ICRP in Publication 38 and by the Society of Nuclear Medicine in a monograph entitled ''MIRD: Radionuclide Data and Decay Schemes.'' The beta spectral data reported here have not been published by either organization. Electronic copies of the files and the utility, along with this report, are available from the Radiation Shielding Information Center at Oak Ridge National Laboratory

Analysis of JUPITER critical experiments by JENDL-3.2

International Nuclear Information System (INIS)

Ishikawa, Makoto

1996-01-01

Applicability of the JENDL-3.2 library to large FBR cores was evaluated using JUPITER experimental data. The nuclear characteristics treated in the present report include criticality, reaction rate ratio, space dependency of C/E values, sodium void reactivity and Doppler reactivity. As a conclusion, JENDL-3.2 is judged to be a well-balanced library for prediction of large FBR core parameters. The unification of integral experimental information from JUPITER and differential nuclear data of JENDL-3.2 will enhance the accuracy and reliability of large FBR core design. (author)

Summary report of the final technical meeting on 'International Reactor Dosimetry File: IRDF-2002'

International Nuclear Information System (INIS)

Griffin, Patrick J.; Paviotti-Corcuera, R.

2003-10-01

Presentations, recommendations and conclusions of the Final Technical Meeting on 'International Reactor Dosimetry File: IRDF-2002' are summarized in this report. The main aims of this meeting were to discuss scientific and technical matters related to reactor dosimetry and to assign responsibilities for the preparation of the final version of the IRDF- 2002 library and the associated TECDOC. Tasks were assigned and deadlines were agreed. Participants emphasized that accurate and complete nuclear data for reactor dosimetry are essential to improve the assessment accuracies for reactor pressure vessel service lifetimes in nuclear power plants, as well as for other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. (author)

Thermal reactor benchmark tests on JENDL-2

International Nuclear Information System (INIS)

Takano, Hideki; Tsuchihashi, Keichiro; Yamane, Tsuyoshi; Akino, Fujiyoshi; Ishiguro, Yukio; Ido, Masaru.

1983-11-01

A group constant library for the thermal reactor standard nuclear design code system SRAC was produced by using the evaluated nuclear data JENDL-2. Furthermore, the group constants for 235 U were calculated also from ENDF/B-V. Thermal reactor benchmark calculations were performed using the produced group constant library. The selected benchmark cores are two water-moderated lattices (TRX-1 and 2), two heavy water-moderated cores (DCA and ETA-1), two graphite-moderated cores (SHE-8 and 13) and eight critical experiments for critical safety. The effective multiplication factors and lattice cell parameters were calculated and compared with the experimental values. The results are summarized as follows. (1) Effective multiplication factors: The results by JENDL-2 are considerably improved in comparison with ones by ENDF/B-IV. The best agreement is obtained by using JENDL-2 and ENDF/B-V (only 235 U) data. (2) Lattice cell parameters: For the rho 28 (the ratio of epithermal to thermal 238 U captures) and C* (the ratio of 238 U captures to 235 U fissions), the values calculated by JENDL-2 are in good agreement with the experimental values. The rho 28 (the ratio of 238 U to 235 U fissions) are overestimated as found also for the fast reactor benchmarks. The rho 02 (the ratio of epithermal to thermal 232 Th captures) calculated by JENDL-2 or ENDF/B-IV are considerably underestimated. The functions of the SRAC system have been continued to be extended according to the needs of its users. A brief description will be given, in Appendix B, to the extended parts of the SRAC system together with the input specification. (author)

Japanese evaluated nuclear data library version 3 revision-3: JENDL-3.3

CERN Document Server

Shibata, K; Kawano, T

2002-01-01

Evaluation for JENDL-3.3 has been performed by considering the accumulated feedback information and various benchmark tests of the previous library JENDL-3.2. The major problems of the JENDL-3.2 data were solved by the new library: overestimation of criticality values for thermal fission reactors was improved by the modifications of fission cross sections and fission neutron spectra for sup 2 sup 3 sup 5 U; incorrect energy distributions of secondary neutrons from important heavy nuclides were replaced with statistical model calculations; the inconsistency between elemental and isotopic evaluations was removed for medium-heavy nuclides. Moreover, covariance data were provided for 20 nuclides. The reliability of JENDL-3.3 was investigated by the benchmark analyses on reactor and shielding performances. The results of the analyses indicate that JENDL-3.3 predicts various reactor and shielding characteristics better than JENDL-3.2. (author)

Analysis of reactivity characteristics of the MONJU initial core using JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Sasaki, Kenji; Suzuki, Takayuki; Suzuki, Norimichi [Power Reactor and Nuclear Fuel Development Corp., Tsuruga, Fukui (Japan). Monju Construction Office; Itagaki, Yoshihiko

1998-03-01

This paper describes the evaluated results of criticality, absorber rod worth and coolant worth in the MONJU initial cores based on the JENDL-3.2 library compared with those of the JENDL-2 library. We confirm that the ratios of calculated and experimental (C/E) values using the JENDL-3.2 library are slightly better than those based on the JENDL-2 library. (author)

Benchmark test of JENDL-3T and -3T/Rev.1

International Nuclear Information System (INIS)

Takano, Hideki; Kaneko, Kunio.

1989-10-01

The fast reactor 70-group constant set JFS-3-J3T has been generated by using the JENDL-3T nuclear data. One-dimensional 21-benchmark cores and the ZPPR-9 core were analysed with the JFS-3-J3T set. The results obtained are summarized as follows: (1) The values of keff are underestimated by 0.6% for Pu-fueled cores and overestimated by 2% for U-fueled cores. (2) The central reaction rate ratio 239 σ f φ/ 235 σ f φ is in a good agreement with the experimental value, though 238 σ c φ/ 239 σ f φ and 238 σ f φ/ 235 σ f φ are overestimated. (3) Doppler and Na-void reactivities are in a good agreement with the measured data. (4) The prediction accuracy of radial reaction rate distributions are improved in the comparison of the results obtained with the JENDL-2 data. Furthermore, the benchmark test of JENDL-3T/Rev. 1 which was revised from JENDL-3T for several important nuclides has been again performed. It was shown that JENDL-3T/Rev. 1 would predict nuclear characteristics more satisfactorily than JENDL-3T. (author)

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

Integral data testing of JENDL-3.2 for fusion reactor and shielding applications

International Nuclear Information System (INIS)

Oyama, Yukio

1995-01-01

Integral data testing of JENDL-3.2 is being performed in the activities of two working groups of the Japanese Nuclear Data Committee. The continuous and group-wise libraries prepared from JENDL-3.2 are planned to be tested by the working groups. In this paper, the continuous library FSXLIB-J3R2 processed from JENDL-3.2 for MCNP was tested for fission and fusion neutrons using data of integral experiments and compared to the results of JENDL-3.1. The results of integral data testing of JENDL-3.2 for fusion and shielding application are reviewed. (author)

Analysis of Doppler effect measurement in FCA cores using JENDL-3.2 library

International Nuclear Information System (INIS)

Okajima, Shigeaki

1996-01-01

For the evaluation of the calculation accuracy of the 238 U Doppler effect using JENDL-3.2 library, the previously measured Doppler reactivity worths in the FCA were systematically analyzed. In the analysis the Doppler reactivity worth was calculated by a first order perturbation theory. The calculated results were compared with those using JENDL-3.1 library. The JENDL-3.2 calculation in MOX fuel mock-up cores agrees well with the experimental values within the experimental error. In U-235/Pu fuel cores, the JENDL-3.2 calculation gives 12-15% larger Doppler reactivity worths than the JENDL-3.1 calculation. (author)

CD-ROM storing JENDL-3.2 plots and data

Energy Technology Data Exchange (ETDEWEB)

Fukahori, Tokio; Iwamoto, Osamu; Nakagawa, Tsuneo; Shibata, Keiichi; Narita, Tsutomu; Katakura, Jun-ichi; Hasegawa, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-01-01

Numerical data and graphs of Japanese Evaluated Nuclear Data Library, Version 3, Revision 2 (JENDL-3.2) have been stored into CD-ROM. JENDL-3.2 is an evaluated nuclear data library which is highly rated in the world, and consists of neutron nuclear data and secondary gamma-ray data for 340 nuclides. JENDL-3.2 has been already released through World Wide Web (WWW, http://cracker.tokai.jaeri.go.jp). For users` convenience and to promote further utilization, this CD-ROM has been produced. The data stored in CD-ROM are the original data and point-wise data (0 K and 300 K) of JENDL-3.2 and graphs of cross sections in gif, PostScript and PDF formats. These data can be retrieved by users` personal computers and engineering workstations, and also can be seen by using WWW browsers. This CD-ROM makes user completely free from downloading, which takes a lot of time, and tedious management such a treating big amount of data. (author)

Summary remarks and recommended reactions for an international data file for dosimetry applications for LWR, FBR, and MFR reactor research, development and testing programs

International Nuclear Information System (INIS)

McElroy, W.N.; Lippincott, E.P.; Grundl, J.A.; Fabry, A.; Dierckx, R.; Farinelli, U.

1979-01-01

The need for the use of an internationally accepted data file for dosimetry applications for light water reactor (LWR), fast breeder reactor (FBR), and magnetic fusion reactor (MFR) research, development, and testing programs continues to exist for the Nuclear Industry. The work of this IAEA meeting, therefore, will be another important step in achieving consensus agreement on an internationally recommended file and its purpose, content, structure, selected reactions, and associated uncertainy files. Summary remarks and a listing of recommended reactions for consideration in the formulation of an ''International Data File for Dosimetry Applications'' are presented in subsequent sections of this report

Toward JENDL-3.3: comments on the problems of JENDL-3.2 for the next version developments

Energy Technology Data Exchange (ETDEWEB)

Hasegawa, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1997-03-01

Revision work for JENDL-3.2 is foreseen and the discussions about the problems of JENDL-3.2 is under going in a small advisory group of JNDC (Japanese Nuclear Data Committee) since this April. In this article main discussion points in that group are presented. Problems from evaluators` point of view, problem from users` point of view and request from anticipated users are summarized. Some discussions on U-235 cross sections in the thermal and epi-thermal energy range are made in relation of the new evaluation of resolved resonance parameters performed by Leal/Derrien. (author)

JENDL-3. The Japanese evaluated nuclear data library

International Nuclear Information System (INIS)

Lemmel, H.D.

1992-01-01

This document summarizes the contents of JENDL-3.1, the Japanese evaluated data library for neutron nuclear data, released in 1989 and revised in Dec. 1990. It also summarizes the JENDL-3 fission-products cross-section data library released in 1990. The entire library or retrievals of selected materials are available on magnetic tape from the IAEA Nuclear Data Section free of charge. (author)

Maxwellian-averaged cross sections calculated from JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Tsuneo; Chiba, Satoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ohsaka, Toshiro; Igashira, Masayuki [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, Tokyo (Japan)

2000-02-01

Maxwellian-averaged cross sections of neutron capture, fission, (n,p) and (n,{alpha}) reactions are calculated from the Japanese Evaluated Nuclear Data Library, JENDL-3.2, for applications in the astrophysics. The calculation was made in the temperature (kT) range from 1 keV to 1 MeV. Results are listed in tables. The Maxwellian-averaged capture cross sections were compared with recommendations of other authors and recent experimental data. Large discrepancies were found among them especially in the light mass nuclides. Since JENDL-3.2 reproduces relatively well the recent experimental data, we conclude that JENDL-3.2 is superior to the others in such a mass region. (author)

MCNP and MATXS cross section libraries based on JENDL-3.3

International Nuclear Information System (INIS)

Kosako, Kazuaki; Konno, Chikara; Fukahori, Tokio; Shibata, Keiichi

2003-01-01

The continuous energy cross section library for the Monte Carlo transport code MCNP-4C, FSXLIB-J33, has been generated from the latest version of JENDL-3.3. The multigroup cross section library with the MATXS format, MATXS-J33, has been generated also from JENDL-3.3. Both libraries contain all nuclides in JENDL-3.3 and are processed at 300 K by the nuclear data processing system NJOY99. (author)

Calibration of a He accumulation fluence monitor for fast reactor dosimetry

Energy Technology Data Exchange (ETDEWEB)

Ito, Chikara [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1997-03-01

The helium accumulation fluence monitor (HAFM) has been developed for a fast reactor dosimetry. The HAFM measurement system was calibrated using He gas and He implanted samples and the measurement accuracy was confirmed to be less than 5%. Based on the preliminary irradiation test in JOYO, the measured He in the {sup 10}B type HAFM agreed well with the calculated values using the JENDL-3.2 library. (author)

Computer program system for evaluation of FP nuclear data for JENDL. Smooth part

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Tsuneo [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Watanabe, Takashi; Iijima, Shungo

1997-12-01

This report describes computer programs used to evaluate nuclear data of fission product (FP) nuclides stored in an evaluated nuclear data library JENDL, especially in the smooth part above the resonance region. Many programs were used for determination of nuclear model parameters, calculation of nuclear data, handling of experimental and/or calculated data, and so on. Among them, reported here are programs for determination of level density parameters (ENSDFRET, LVLPLOT, LEVDES), for making sets of JCL and input data for the theoretical calculation program CASTHY (JOBSETTER, INDES/CASTHY), and for conversion of format of CASTHY output files to the ENDF format (CTOB2). (author). 51 refs.

FAIR-DDX, Double Diffusion Cross-Sections Scattering Matrix Generated from ENDF/B-4 or JENDL-2

International Nuclear Information System (INIS)

Minami, Kazuyoshi; Yamano, Naoki

2001-01-01

1 - Description of program or function: FAIR-DDX produces double differential (energy and angle) cross sections (DDX) in the form of group-to-group scattering matrices using the evaluated nuclear data libraries JENDL-2 or ENDF/B-IV. The DDX form is useful for verification of the evaluated data, such as the inelastic scattering, through comparison with the experimental DDX values. 2 - Method of solution: DDX uses the file 4 data (angular distribution of secondary neutrons) and the energy and momentum conservation laws. For continuum region reactions, file 5 (energy spectrum of secondary neutrons) is used. To express the angular distribution of secondary neutrons in group-to-group scattering matrices FAIR-DDX adopts a direct angular representation method. 3 - Restrictions on the complexity of the problem: The maximum number of energy groups is 200

Computational analysis of Bangladesh 3 MW TRIGA research reactor using MCNP4C, JENDL-3.3 and ENDF/B-Vl data libraries

International Nuclear Information System (INIS)

Huda, M.Q.

2006-01-01

The three-dimensional continuous energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Validation of the JENDL-3.3 and ENDF/BVI continuous energy cross-section data for MCNP4C was performed against some well-known benchmark lattices. For TRIGA analysis, data from JENDL-3.3 and ENDF/B-VI in combination with the JENDL-3.2 and ENDF/B-V data files (for nat Zr, nat Mo, nat Cr, nat Fe, nat Ni, nat Si, and nat Mg) at 300 K evaluations were used. Full S(α, β) scattering functions from ENDF/B-V for Zr in ZrH, H in ZrH and water molecule, and for graphite were used in both cases. The validation of the model was performed against the criticality and reactivity benchmark experiments of the TRIGA reactor. There is ∼20.0% decrease of thermal neutron flux occurs when the thermal library is removed during the calculation. Effect of erbium isotope that is present in the TRIGA fuel was also studied. In addition to the effective multiplication values, the well-known integral parameters: δ 28 , δ 25 , ρ 25 , and C * were calculated and compared for both JENDL3.3 and ENDF/B-VI libraries and were found to be in very good agreement. Results are also reported for most of the analyses performed by JENDL-3.2 and ENDF/B-V data libraries

Sensitivity analysis for reactivity parameter change of the creole experiment caused by the differences between ENDF-BVII and JENDL neutron cross section evaluations

International Nuclear Information System (INIS)

Boulaich, Y.; Bardouni, C.; Elyounoussi, C.; Elbakkari, H.; Boukhal, H.; Erradi, L.; Nacir, B.

2011-01-01

Full text: In this work, we present our analysis of the CREOLE experiment on the parameter by using the three-dimensional continuous energy code (MCNPS) and the last updated nuclear data evaluations. This experiment performed in the EOLE critical facility located at CEA-Cadarache, was dedicated to studies for both UO2 and UO2-PuO2 PWR type lattices covering the whole temperature range from 20 0 C to 300 0 C. We have developed an accurate model of the EOLE reactor to be used by the MCNP5 Monte Carlo code. This model guarantees a high level of fidelity in the description of different configurations at various temperatures taking into account their consequence on neutron cross section data and all thermal expansion effects. In this case, the remaining error between calculation and experiment will be awarded mainly to uncertainties on nuclear data. Our own cross section library was constructed by using NJOY99.259 code with point-wise nuclear data based on ENDF-BVII. JEFF3.1, JENDL3.3 and JENDL4 evaluation files. The MCNP model was validated through the axial and radial fission rate measurements at room and hot temperatures. Calculation-experiment discrepancies of the reactivity parameter were analyzed and the results have shown that the JENDL evaluations give the most consistent values. In order to specify the source of the relatively large difference between experiment and calculation due to ENDF-BVII nuclear data evaluation, the discrepancy in reactivity between ENDF-BVII and JENDL evaluations was decomposed using sensitivity and uncertainty analysis technique

Total cross-sections assessment of neutron reaction with stainless steel SUS-310 contained in various nuclear data files

International Nuclear Information System (INIS)

Suwoto

2002-01-01

The integral testing of neutron cross-sections for Stainless Steel SUS-310 contained in various nuclear data files have been performed. The shielding benchmark calculations for Stainless Steel SUS-310 has been analysed through ORNL-Broomstick Experiment calculation which performed by MAERKER, R.E. at ORNL - USA ( 1) . Assessment with JENDL-3.1, JENDL-3.2, ENDF/B-IV, ENDF/B-VI nuclear data files and data from GEEL have also been carried out. The overall calculation results SUS-310 show in a good agreement with the experimental data, although, underestimate results appear below 3 MeV for all nuclear data files. These underestimation tendencies clearly caused by presented of iron nuclide which more than half in Stainless Steel compound. The total neutron cross-sections of iron nuclide contained in various nuclear data files relatively lower on that energy ranges

Consistency between data from the ENDF/B-V dosimetry file and corresponding experimental data for some fast neutron reference spectra

International Nuclear Information System (INIS)

Nolthenius, H.J.; Zijp, W.L.

1981-11-01

Results are given of a study on the consistency between 'integral' and 'differential' cross sections data for four benchmark neutron spectra and 36 neutron reactions of importance for reactor neutron metrology. The energy dependent cross section data and their uncertainty data are obtained from the ENDF/B-V dosimetry file. The reactions have been considered with respect to the following quantities: 1. the precision of the averaged cross sections, for a specified spectrum; 2. the discrepancy between the measured and the calculated average cross section values; 3. the consistency between the measured and calculated average cross section values, described by the chi 2 -parameter. It was possible to take into account the available cross section covariance information present in the ENDF/B-V dosimetry file. Covariance information on the benchmark flux density spectra was not taken into account in this study

JENDL-4.0 benchmarking for effective delayed neutron fraction of fast neutron systems

International Nuclear Information System (INIS)

Chiba, Go; Tsuji, Masashi; Sugiyama, Ken-ichiro; Narabayashi, Tadashi

2011-01-01

The performance of the latest Japanese evaluated nuclear data library JENDL-4.0 for the prediction of effective delayed neutron fraction β eff is assessed using experimental data of a wide range of fast neutron systems. Covariance data of JENDL-4.0 are used to quantify nuclear-data-induced uncertainties. Calculations with other libraries. JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1, are also carried out for a quantitative comparison. JENDL-4.0 results in good agreement between calculation and experimental values within total uncertainties, and consistency between the differential nuclear data and integral experimental data is confirmed. While the other libraries also show good performance for β eff prediction, there are small differences in the predicted values of β eff among different libraries and ENDF/B-VII.0 gives the most stable results. Furthermore, a simple and convenient procedure to calculate sensitivity profiles of β eff to nuclear data is proposed. (author)

Benchmark comparisons of evaluated nuclear data files

International Nuclear Information System (INIS)

Resler, D.A.; Howerton, R.J.; White, R.M.

1994-05-01

With the availability and maturity of several evaluated nuclear data files, it is timely to compare the results of integral tests with calculations using these different files. We discuss here our progress in making integral benchmark tests of the following nuclear data files: ENDL-94, ENDF/B-V and -VI, JENDL-3, JEF-2, and BROND-2. The methods used to process these evaluated libraries in a consistent way into applications files for use in Monte Carlo calculations is presented. Using these libraries, we are calculating and comparing to experiment k eff for 68 fast critical assemblies of 233,235 U and 239 Pu with reflectors of various material and thickness

Differences among KERMA or DPA data calculated from JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.1b with NJOY

International Nuclear Information System (INIS)

Konno, C.

2016-01-01

KERMA factors and DPA cross-section data below 20 MeV in the official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.1b were compared in detail. As a result, it was found out that the KERMA and DPA data of a lot of nuclei were different among the nuclear data libraries. Reasons of most of the differences were successfully categorized to the nuclear data and NJOY issues. The KERMA factors and DPA cross-section data in the ACE files with the problems should be revised. (author)

Evaluation of covariances for resolved resonance parameters of 235U, 238U, and 239Pu in JENDL-3.2

International Nuclear Information System (INIS)

Kawano, Toshihiko; Shibata, Keiichi

2003-02-01

Evaluation of covariances for resolved resonance parameters of 235 U, 238 U, and 239 Pu was carried out. Although a large number of resolved resonances are observed for major actinides, uncertainties in averaged cross sections are more important than those in resonance parameters in reactor calculations. We developed a simple method which derives a covariance matrix for the resolved resonance parameters from uncertainties in the averaged cross sections. The method was adopted to evaluate the covariance data for some important actinides, and the results were compiled in the JENDL-3.2 covariance file. (author)

Present status and history of nuclear data development for transmutation technology

CERN Document Server

Hasegawa, A

2002-01-01

A history of development of nuclear data from JENDL (Japanese Evaluated Nuclear Data Library)-1 to JENDL-3.3 and JENDL-HF (High energy File) and JENDL- Actinide File are stated. 5 nuclear data such as JENDL-1, JENDL-2, JENDL-3.1, JENDL-3.2 and JENDL-3.3 have been developed by JAERI. JENDL-1 for fast reactor (1977) has 66+6 nuclide and 15 MeV the largest energy. JENDL-2 for fast and light water reactor (1982) has 173+8 nuclide and 20 MeV. All-purpose nuclear data: JENDL-3.1 (1990) with 305+19 nuclide and 59 2nd gamma-ray data, JENDL-3.2 (1994) with 318+22 nuclide and 66 2nd gamma-ray data and JENDL-3.3 (2002) with 335+2 nuclide, 114 2nd gamma-ray data, 60 angular dependence neutron data and 20 MeV have been developed. JENDL-3.3 was opened at JAERI home page in May 2002. JENDL High Energy library consists of JENDL-HF, JENDL-Photonuclear Data File and JENDL-PKA/KERMA File. JENDL-HF includes nuclear reaction data of neutron and proton incidence, for example, total cross section, elastic scattering cross section a...

Analysis of benchmark lattices with endf/b-vi, jef-2.2 and jendl-3 data

International Nuclear Information System (INIS)

Saglam, M.

1995-01-01

The NJOY Nuclear Data Processing System has been used to process the ENDF/B-VI , JEF-2.2 and JENDL-3 Nuclear Cross Section Data Bases into multigroup form. A brief description of the data bases is given and the assumptions made in processing the data from evaluated nuclear data file format to multigroup format are presented. The differences and similarities of the Evaluated Nuclear Data Files have been investigated by producing four group cross sections by using the GROUPIE code and calculating thermal, fission spectrum averaged and 2200 m/s cross sections and resonance integrals using the INTER cale. It has been shown that the evaluated data for U238 in JEF and ENDF/B-VI are principally the same while in case of U235 the same is true for JENDL and ENDF/B-VI. The evaluations for U233 and Th232 are different for all three ENDF files. Several utility codes have been written to convert the multigroup library into a WIMS-D4 compatible binary library. The performance and suitability of the generated libraries have been tested with the use of metal tueled TRX lattices, uranium oxide fueled BAPL lattices and Th232-U233 fueled BNL lattices. The use ot a new thermal scattering matrix for Hydrogen from ENDF/B-VI increased keff for 0.5 o/ while the use of ENDF/B-VI U238 decreased it for 2.5 %. Although the original WIMS library performed well for Ihe effective multiplication factor of the lattices there is an improvement for the epithermal to thermal capture rate of U238 while using new data in the TRX and BAPL lattices. The effect of the fission spectrum is investigated for the BNL lattices and it is shown that using U233 fission spectrum instead of the original U235 spectrum gives a keff which agrees better with the experimental value. The results obtained by using new multigroup data are generally acceptable and in the experimental error range. They especially improve the prediction of the reaction rate dependent benchmark parameters

Neutron nuclear data evaluation of cesium isotopes for JENDL-4.0

International Nuclear Information System (INIS)

Iwamoto, Nobuyuki

2011-01-01

For the development of JENDL-4.0, neutron nuclear data for fission product nuclides, 133,134,135,136,137 Cs, were revised in the incident neutron energy range from 1 eV to 20 MeV by using a coupled-channels optical model (OM), and nuclear reaction models. The OM potential parameters were determined for stable 133 Cs to reproduce the experimental data of total and elastic scattering cross sections and angular distributions of elastically scattered neutrons. The present results reasonably reproduce measured data for (n,2n), (n,p), (n,α), and capture reactions on 133 Cs. Important differences between the present results and JENDL-3.3 are found for the capture cross sections of 134,137 Cs. The cross section obtained for 137 Cs was smaller than that in JENDL-3.3. This result makes the transmutation of medium-lived 137 Cs increasingly difficult. The production probabilities of metastable states for 134,138 Cs via capture reactions on 133,137 Cs are compared with experimental values. The present result for 134m Cs production is marginally consistent with measured data. However, a large discrepancy is recognized for 138m Cs production. The γ-ray emission data were evaluated with available measurements, and newly compiled in JENDL-4.0. Maxwellian-averaged capture cross sections were calculated in the energy range from 1 to 10 3 keV, and are compared with other derived data. (author)

Development of Indian cross section data files for Th-232 and U-233 and integral validation studies

International Nuclear Information System (INIS)

Ganesan, S.

1988-01-01

This paper presents an overview of the tasks performed towards the development of Indian cross section data files for Th-232 and U-233. Discrepancies in various neutron induced reaction cross sections in various available evaluated data files have been obtained by processing the basic data into multigroup form and intercomparison of the latter. Interesting results of integral validation studies for capture, fission and (n,2n) cross sections for Th-232 by analyses of selected integral measurements are presented. In the resonance range, energy regions where significant differences in the calculated self-shielding factors for Th-232 occur have been identified by a comparison of self-shielded multigroup cross sections derived from two recent evaluated data files, viz., ENDF/B-V (Rev.2) and JENDL-2, for several dilutions and temperatures. For U-233, the three different basic data files ENDF/B-IV, JENDL-2 and ENDL-84 were intercompared. Interesting observations on the predictional capability of these files for the criticality of the spherical metal U-233 system are given. The current status of Indian data file is presented. (author) 62 ref

Analyses of criticality and reactivity for TRACY experiments based on JENDL-3.3 data library

International Nuclear Information System (INIS)

Sono, Hiroki; Miyoshi, Yoshinori; Nakajima, Ken

2003-01-01

The parameters on criticality and reactivity employed for computational simulations of the TRACY supercritical experiments were analyzed using a recently revised nuclear data library, JENDL-3.3. The parameters based on the JENDL-3.3 library were compared to those based on two former-used libraries, JENDL-3.2 and ENDF/B-VI. In the analyses computational codes, MVP, MCNP version 4C and TWOTRAN, were used. The following conclusions were obtained from the analyses: (1) The computational biases of the effective neutron multiplication factor attributable to the nuclear data libraries and to the computational codes do not depend the TRACY experimental conditions such as fuel conditions. (2) The fractional discrepancies in the kinetic parameters and coefficients of reactivity are within ∼5% between the three libraries. By comparison between calculations and measurements of the parameters, the JENDL-3.3 library is expected to give closer values to the measurements than the JENDL-3.2 and ENDF/B-VI libraries. (3) While the reactivity worth of transient rods expressed in the $ unit shows ∼5% discrepancy between the three libraries according to their respective β eff values, there is little discrepancy in that expressed in the Δk/k unit. (author)

Evaluation of Some (n,n'), (n,γ), (n,p), (n,2n) and (n,3n) Reaction Excitation Functions for Fission and Fusion Reactor Dosimetry Applications; Evaluation of the Excitation Functions for the 54Fe(n,p)54Mn, 58Ni(n,2n)57Ni, 67Zn(n,p)67Cu, 92Mo(n,p)92mNb, 93Nb(n,γ)94Nb, 113In(n,n')113mIn, 115In(n,γ) 116mIn, and 169Tm(n,3n)167Tm Reactions. Progress Report on Research Contract No 16242

International Nuclear Information System (INIS)

Zolotarev, K.I.; Zolotarev, P.K.

2013-12-01

Cross section data for the 54 Fe(n,p) 54 Mn, 58 Ni(n,2n) 57 Ni, 67 Zn(n,p) 67 Cu, 92 Mo(n,p) 92m Nb, 93 Nb(n,γ) 94 Nb, 113 In(n,n') 113m In, 115 In(n,γ) 116m In, 169 Tm(n,3n) 167 Tm reactions are needed to solve a wide spectrum of scientific and technical tasks. Activation detectors based on these reactions may be used in the field of reactor dosimetry. Furthermore, the 54 Fe(n,p) 54 Mn reaction is often used in experimental nuclear physics as a monitor reaction for measurements of unknown cross sections by means of the activation method over the neutron energy range from 5 to 15 MeV. The 93 Nb(n,γ) 94 Nb reaction is also very promising for using in retrospective neutron dosimetry for determination of total neutron fluence during a campaign of a reactor. In the existing version of the International Reactor Dosimetry File and the new extended version named as IRDFF data for excitation functions of 67 Zn(n,p) 67 Cu, 92 Mo(n,p) 92m Nb, 113 In(n,n') 113m In, and 169 Tm(n,3n) 167 Tm reactions are absent. Data for these reactions are also absent in the JENDL/D-99 dosimetry file. Excitation functions of 67 Zn(n,p) 67 Cu and 169 Tm(n,3n) 167 Tm are presented in the TENDL-2012, EAF-2010, JENDL-4.0, JEFF-3.1/A, MENDL-2 libraries. Cross section data for the 67 Zn(n,p) 67 Cu reaction up to 20 MeV are given also in the JENDL/HE-2007 library. Excitation functions of the 92 Mo(n,p) 92m Nb and 113 In(n,n') 113m In reactions are evaluated in the EAF-2010 and JEFF-3.1/A libraries. Cross section data for the 113 In(n,n') 113m In reaction are given also in the TENDL-2010 library. It is necessary to note that neutron data in the JEFF-3.1/A and JENDL-4.0 libraries were evaluated up to 20 MeV. Neutron data in the TENDL-2012, EAF-2010, MENDL-2 and TENDL-2010 libraries had been evaluated up to 30 MeV, 60 MeV, 100 MeV and 200 MeV, respectively. Neutron cross sections in the MENDL-2, TENDL-2010 and TENDL-2012 libraries had been obtained on the basis of pure theoretical model calculations

Evaluation of the fusion-related neutron nuclear data for JENDL-3

International Nuclear Information System (INIS)

Chiba, Satoshi

1988-01-01

Status of the neutron nuclear data evaluations for JENDL-3 will be described for nuclides important in the development of D-T fusion reactors. In this article, however, only explanation of the evaluations for the very light mass region will be presented to avoid overlapping with what are given in another papers submitted to this seminar. Emphases are placed on the tritium production cross sections, inelastic scattering cross sections including the double-differential neutron emission spectrum (DDX), threshold reaction cross sections and photon production cross sections. The methods employed to prepare JENDL-3T library and their results will be summarized. (author)

Review of nuclear data of major actinides and 56Fe in JENDL-4.0

International Nuclear Information System (INIS)

Iwamoto, Osamu; Iwamoto, Nobuyuki

2014-01-01

JENDL-4.0 was released in 2010. Most of the actinide data were revised from JENDL-3.3, especially at the fast energy region, using the newly developed CCONE code. Fission cross-sections of major actinides were determined using the least-squares method and the SOK code, taking available experimental data into account. In the resonance region around 1 keV for 235 U, capture cross-sections were reduced to improve integral benchmark tests. The reduction was confirmed by Jandel, et al. during a later experiment at Los Alamos. For 56 Fe, cross-sections of inelastic scattering and angular distributions of elastic scattering were revised from JENDL-3.3. Evaluation methods and comparisons with experimental data and other evaluated data will be reviewed. (authors)

Generation and validation of the WIMS-D5 library based on JENDL-3.2

International Nuclear Information System (INIS)

Gil, Choong-Sup; Kim, Jung-Do

2002-01-01

A WIMS-D5 library based on JENDL-3.2 has been prepared and is being tested against benchmark problems. Several sensitivity calculations for stability confirmation of the library were carried out, such as fission spectrum dependency and the self-shielding effects of the elastic scattering cross sections and the self-shielding effects 240 Pu and 242 Pu capture cross sections below 4.0 eV. The results of benchmark calculations with the libraries based on JENDL-3.2, ENDF/B-VI.5, JEF-2.2, and the 1986 WIMS-D library were intercompared. The multiplication factors for the thermal lattices are slightly underpredicted by all libraries (up to 1% with ENDF/B-VI.5). The k eff values with the library based on JENDL-3.2 are slightly higher than those of ENDF/B-VI.5 and JEF-2.2. The spectral indices for the lattices with JENDL-3.2 agree with the measured quantities within the uncertainties of the experiments. The calculated amounts of some isotopes such as 149 Sm, 237 Np, 238 Pu, 242 Cm and 243 Cm show large differences from the measured or reference values. (author)

Present status and benchmark tests of JENDL-2

International Nuclear Information System (INIS)

Kikuchi, Y.

1983-01-01

The second version of Japanese Evaluated Nuclear Data Library (JENDL-2) consists of the evaluated data from 10 -5 eV to 20 MeV for 176 nuclides including 99 fission product nuclei. Complete reevaluation has been made to heavy actinide, fission product and main structural material nuclides. Benchmark tests have been made on JENDL-2 for fast reactor application. Various characteristics in core center have been tested with one-dimensional model for total of 27 assemblies, and more sophisticated problems have been examined for MOZART and ZPPR-3. Furthermore analyses of JUPITER project give useful information. Satisfactory results have been obtained as a whole. However, the spectrum is a little underestimated above a few hundred keV and below a few keV. The positive sodium void reactivity worth is much overestimated. As to the latter, the sensitivity analysis with the generalized perturbation method suggests that the fission cross section of 239 Pu below a few keV has an important role. (Auth.)

Development of the adjusted nuclear cross-section library based on JENDL-3.2 for large FBR

International Nuclear Information System (INIS)

Yokoyama, Kenji; Ishikawa, Makoto; Numata, Kazuyuki

1999-04-01

JNC (and PNC) had developed the adjusted nuclear cross-section library in which the results of the JUPITER experiments were reflected. Using this adjusted library, the distinct improvement of the accuracy in nuclear design of FBR cores had been achieved. As a recent research, JNC develops a database of other integral data in addition to the JUPITER experiments, aiming at further improvement for accuracy and reliability. In 1991, the adjusted library based on JENDL-2, JFS-3-J2 (ADJ91R), was developed, and it has been used on the design research for FBR. As an evaluated nuclear library, however, JENDL-3.2 is recently used. Therefore, the authors developed an adjusted library based on JENDL-3.2 which is called JFS-3-J3.2(ADJ98). It is known that the adjusted library based on JENDL-2 overestimated the sodium void reactivity worth by 10-20%. It is expected that the adjusted library based on JENDL-3.2 solve the problem. The adjusted library JFS-3-J3.2(ADJ98) was produced with the same method as the adjusted library JFS-3-J2(ADJ91R) and used more integral parameters of JUPITER experiments than the adjusted library JFS-3-J2(ADJ91R). This report also describes the design accuracy estimation on a 600 MWe class FBR with the adjusted library JFS-3-J3.2(ADJ98). Its main nuclear design parameters (multiplication factor, burn-up reactivity loss, breeding ratio, etc.) except the sodium void reactivity worth which are calculated with the adjusted library JFS-3-J3.2(ADJ98) are almost the same as those predicted with JFS-3-J2(ADJ91R). As for the sodium void reactivity, the adjusted library JFS-3-J3.2(ADJ98) estimates about 4% smaller than the JFS-3-J2(ADJ91R) because of the change of the basic nuclear library from JENDL-2 to JENDL-3.2. (author)

ZZ ORIGEN2.2-UPJ, A complete package of ORIGEN2 libraries based on JENDL-3.2 and JENDL-3.3

International Nuclear Information System (INIS)

Ishikawa, Makoto; Kataoka, Masaharu; Ohkawachi, Yasushi; Ohki, Shigeo; JIN, Tomoyuki; Katakura, Jun-ich; Suyama, Kenya; Yanagisawa, Hiroshi; Matsumoto, Hideki; ONOUE, Akira; Sasahara, Akihiro

2006-01-01

1 - Description: ORLIBJ32 is a package of the libraries for ORIGEN2 code based on JENDL-3.2(NEA-1642). The one grouped cross section data for PWR and BWR were compiled using the burnup calculation results by SWAT code. The FBR libraries were compiled by the analysis system used at JNC for FBR core calculation. The fission yield and decay constants data were also updated using the second version of the JNDC FP library. In ORLIBJ32, not only one-grouped cross section data but also variable actinide cross section data are prepared, using a code written in FORTRAN77. The routines should be linked to the Original ORIGEN2.1 program. The LWR Libraries are prepared based on the current PWR fuel assembly specification, and the FBR libraries are based on the request by the Japanese FBR researchers. Before compiling the libraries, the specification of fuel assembly was completely reviewed and evaluated by the members of Working Group in the Japanese Nuclear Data Committee, 'working group on the evaluation of the amount of isotope generation'. ORLIBJ33 is a new libraries based on JENDL-3.3 following the release of JENDL-3.3. The parameters used to prepare the library are the same as those of ORLIBJ32. The Original version or ORLIBJ33 is coupled with ORIGEN2.1. But after the release of ORIGEN2.2 from ORNL as CCC-0371 through RSICC, several requests for a combination with ORLIBJ33 and ORIGEN2.2 were received. During the development of ORLIBJ33, released as NEA-1642, authors found a problem in the library maker for FBR libraries, and consequently it was revised and tested in JNC-Oarai. This package 'ORIGEN2.2-UPJ' contains: - updated source code of ORIGEN2.2 of CCC-0371 to use ORLIBJ32 and ORLIBJ33, - all Original libraries in CCC-0371, - ORLIBJ32 in NEA-164/03 (but libraries for FBR are revised), - and ORLIBJ33. In this package, decay data based on the second version of the JNDC FP library and, photon and decay data libraries based on JENDL-3.3 are also included. NLB and NLIB

JENDL-4.0 benchmarking for effective delayed neutron fraction with a continuous-energy Monte Carlo code MVP

International Nuclear Information System (INIS)

Nagaya, Yasunobu

2013-01-01

Benchmark calculations with a continuous-energy Monte Carlo code have been performed for delayed neutron data of JENDL-4.0. JENDL-4.0 gives good prediction for the effective delayed neutron fraction in the present benchmarks but further detailed analysis is required for some cores. (author)

Status of neutron cross sections for reactor dosimetry

International Nuclear Information System (INIS)

Vlasov, M.F.; Fabry, A.; McElroy, W.N.

1977-03-01

The status of current international efforts to develop standardized sets of evaluated energy-dependent (differential) neutron cross sections for reactor dosimetry is reviewed. The status and availability of differential data are considered, some recent results of the data testing of the ENDF/B-IV dosimetry file using 252 Cf and 235 U benchmark reference neutron fields are presented, and a brief review is given of the current efforts to characterize and identify dosimetry benchmark radiation fields

Integral test of International Reactor Dosimetry and Fusion File with Li{sub 2}O assembly and DT neutron source at JAEA/FNS

Energy Technology Data Exchange (ETDEWEB)

Sato, Satoshi, E-mail: sato.satoshi92@jaea.go.jp [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken (Japan); Kwon, Saerom; Ohta, Masayuki [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken (Japan); Ochiai, Kentaro [Japan Atomic Energy Agency, Rokkasho-mura, Kamikita-gun, Aomori-ken (Japan); Konno, Chikara [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki-ken (Japan)

2016-11-01

In order to validate a new library of dosimetry cross section data, International Reactor Dosimetry and Fusion File release 1.0 (IRDFF 1.0), not only for DT neutrons but also for neutrons with energy of less than 14 MeV, we perform an integral test with a Li{sub 2}O rectangular assembly of 60.7 cm in thickness and a DT neutron source at JAEA/FNS. We place a lot of activation foils at depths of 10.1 cm and 30.4 cm for measurements of dosimetry reaction rates in small space along the central axis in the assembly, measure decay gamma-rays from the activation foils with high-purity Ge detectors after the DT neutron irradiation by the foil activation technique, and deduce a variety of dosimetry reaction rates. We calculate the reaction rates by using a Monte Carlo code MCNP5-1.40 and the nuclear data library ENDF/B-VII.1 with the IRDFF-v.1.05 as the response functions for the dosimetry reactions. The calculation results generally show good agreements with the measured ones, and it can be confirmed that most of the data in IRDFF-v.1.05 are valid for the neutron field in the Li{sub 2}O assembly with the DT neutrons.

JENDL reactor constant and its application. The 40th anniversary of Japan nuclear data committee

International Nuclear Information System (INIS)

Zukeran, Atsushi

2004-01-01

The status of reactor constants about 27 years ago is briefly reviewed from the criticality predictions and nuclear data processing codes. In the second section, status of current users of JENDL-3.3 and/or JENDL-3.2 is consulted with the 2003 Fall Meeting of Atomic Energy Society of Japan. In the third section, the reliabilities of JENDL-3.3 and -3.2 are reviewed mainly from the application to light water reactor (LWR) mockup experiments; MISTRAL and BASALA made on EOLE critical facility of Cadarache Laboratory in France, since an extensive evaluation for nuclear data applicability to LWR have been scarcely performed in relative to for FBR. The results of international benchmark cores and criticality safety analyses are briefly reviewed. In the concluding remarks, overall applicability is shown as a summary with respect to all reactor parameters obtained in the LWR mockup experiment and some remarks are noted. (author)

Analysis of the MOZART experiments with JENDL-1

International Nuclear Information System (INIS)

Seki, Yuji; Hojuyama, Takeshi

1979-01-01

Analyses of the MOZART experiments have been made on reaction rate distributions, sodium void reactivity map and control rod worths to confirm the applicability of JENDL-1 to fast reactor core design. Some anomalous C/E-values have been observed in these core parameters at the outer core and the blanket. Further study is needed in both nuclear data evaluation and data processing method. (author)

Analyses of iron and concrete shielding experiments at JAEA/TIARA with JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0

International Nuclear Information System (INIS)

Konno, Chikara; Ochiai, Kentaro; Sato, Satoshi; Ohta, Masayuki

2015-01-01

IAEA released a new Fusion Evaluated Nuclear Data Library, FENDL-3.0, in 2012. FENDL-3.0 extends the neutron energy range from 20 MeV to greater than 60 MeV. Now there is increasing interest in nuclear data above 20 MeV. Thus we have analyzed the iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in Japan Atomic Energy Agency with the latest high-energy nuclear data libraries, JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0. The Monte Carlo code MCNP-5 and ACE files of JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0, which are supplied from JAEA, LANL and IAEA, respectively, were used for this analysis. The collimated neutron beam and test shields were modeled in the analysis. The measured source neutron data were adopted in the analysis. The followings are found out from the results; (1) Iron experiments: The calculation result with FENDL-3.0 agrees with the measured one best. That with JENDL/HE-2007 fairly agrees with the measured one. On the contrary that with ENDF/B-VII.1 drastically overestimates the measured one. It is confirmed that this overestimation is due to the smaller non-elastic scattering data of "5"6Fe in ENDF/B-VII.1. (2) Concrete experiments: The calculation result with ENDL/HE-2007 agrees with the measured one best, while those with FENDL-3.0 and ENDF/B-VII.1 drastically overestimate the measured one. It is confirmed that this overestimation is due to both the larger elastic and smaller non-elastic scattering data of "1"6O in FENDL-3.0 and ENDF/B-VII.1.

Benchmark test of evaluated nuclear data files for fast reactor neutronics application

International Nuclear Information System (INIS)

Chiba, Go; Hazama, Taira; Iwai, Takehiko; Numata, Kazuyuki

2007-07-01

A benchmark test of the latest evaluated nuclear data files, JENDL-3.3, JEFF-3.1 and ENDF/B-VII.0, has been carried out for fast reactor neutronics application. For this benchmark test, experimental data obtained at fast critical assemblies and fast power reactors are utilized. In addition to comparing of numerical solutions with the experimental data, we have extracted several cross sections, in which differences between three nuclear data files affect significantly numerical solutions, by virtue of sensitivity analyses. This benchmark test concludes that ENDF/B-VII.0 predicts well the neutronics characteristics of fast neutron systems rather than the other nuclear data files. (author)

Inter-comparison of High Energy Files (neutron-induced, from 20 to 150 MeV)

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Ouk; Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-11-01

Recent new applications using accelerator-driven system require well-tested nuclear data when modeling the interaction of neutrons above 20 MeV. This work is aimed to review evaluation methods applied in currently available neutron high energy files above 20 to 150 MeV, to inter-compare their evaluated cross sections on some important isotopes, and to analyze resulting discrepancies. Through out these, integrities and consistencies of the high energy files are checked, applicability of physics models and evaluation methodologies are assessed, and some directions are derived to improve and expand current JENDL High Energy File. (author)

Activation cross section data file, (1)

International Nuclear Information System (INIS)

Yamamuro, Nobuhiro; Iijima, Shungo.

1989-09-01

To evaluate the radioisotope productions due to the neutron irradiation in fission of fusion reactors, the data for the activation cross sections ought to be provided. It is planning to file more than 2000 activation cross sections at final. In the current year, the neutron cross sections for 14 elements from Ni to W have been calculated and evaluated in the energy range 10 -5 to 20 MeV. The calculations with a simplified-input nuclear cross section calculation system SINCROS were described, and another method of evaluation which is consistent with the JENDL-3 were also mentioned. The results of cross section calculation are in good agreement with experimental data and they were stored in the file 8, 9 and 10 of ENDF/B format. (author)

Summary report of consultants' meeting to review the requirements to improve and extend the IRDF library (International Reactor Dosimetry File (IRDF-2002))

International Nuclear Information System (INIS)

Greenwood, L.R.; Nichols, A.L.

2007-01-01

Presentations, recommendations and conclusions of a Consultants' Meeting to 'Review the Requirements to Improve and Extend the IRDF library (International Reactor Dosimetry File (IRDF-2002))' are summarized is this report. The main aims of this meeting were to discuss scientific and technical matters related to reactor dosimetry and to consider the needs for improvements to the existing data in IRDF-2002 and possible extensions to other higher neutron energy applications. Specific tasks were assigned and deadlines agreed. The requirements for fusion studies are particularly challenging (up to 60 MeV) and should include adequate covariance data - the provision of these neutron cross sections will require additional effort and assessment prior to initiating any work programme, and specific participants agreed to undertake preliminary exercises. (author)

Revision of heavy nuclei data in JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Kawano, Toshihiko [Kyushu Univ., Advanced Energy Engineering Science, Kasuga, Fukuoka (Japan)

2000-03-01

In order to deal with problems concerning the data of heavy nuclides in JENDL-3.2, a working group was organized to update the evaluated nuclear data of Uranium, Plutonium, and Thorium isotopes. The current status of the working group is reviewed, and some results about resonance parameters, secondary neutron energy spectra, fission cross sections, and direct/semidirect capture process are shown. (author)

Revision of heavy nuclei data in JENDL-3.2

International Nuclear Information System (INIS)

Kawano, Toshihiko

2000-01-01

In order to deal with problems concerning the data of heavy nuclides in JENDL-3.2, a working group was organized to update the evaluated nuclear data of Uranium, Plutonium, and Thorium isotopes. The current status of the working group is reviewed, and some results about resonance parameters, secondary neutron energy spectra, fission cross sections, and direct/semidirect capture process are shown. (author)

Analysis of MISTRAL experiments with JENDL-3.2

International Nuclear Information System (INIS)

Umano, Takuya; Yamamoto, Toru; Kanda, Ryoji; Sasagawa, Masaru; Kan, Taro; Ishii, Kazuya; Ando, Yoshihira; Tatsumi, Masahiro

2003-01-01

NUPEC and CEA have launched and extensive experimental program called MISTRAL to study highly moderated MOX cores for the advanced LWRs. The analyses using the SRAC system and the MVP code with the JENDL-3.2 library are in progress on the experiments of the MISTRAL program and also the EPICURE program that was carried out by CEA before the MISTRAL program. Various comparisons have been made between the calculation results and the measurement values. (author)

Update of WIMS-D libraries using JENDL-3.2, ENDF/B-VI.5 and JEF-2.2

International Nuclear Information System (INIS)

Gil, Choong Sup; Kim, Jung Do; Chang, Jong Wha

2001-01-01

The WIMS-D5 Libraries based on JENDL-3.2, ENDF/B-VI.5, and JEF-2.2 have been prepared and are being tested against the benchmark problems. Several sensitivity calculations for stabililty confirmation of the libraries were carried out such as the fission spectrum dependency, the self shielding effects of the elastic scattering cross sections, the self shielding effects of Pu -240 and Pu -242 capture cross sections below 4.0 eV, etc. The results of benchmark calculations with the libraries based on JENDL-3.2, ENDF/B-VI.5, JEF-2.2, and the '1986 library were intercompared. The predictions of criticalities and isotopic compositions with the updated libraries show good agreements with the measurements or the reference results. The multiplication factors with the library based on JENDL-3.2 are slightly higher than those of ENDF/B-VI.5 and JEF-2.2

The JENDL-3 sublibrary for gas production. Summary of contents

International Nuclear Information System (INIS)

Nakagawa, T.; Narita, T.

1993-01-01

This document summarizes the contents of the JENDL-3 Sublibrary for Gas Production. This nuclear data library contains neutron-induced production cross-sections of hydrogen and helium nuclei for elements from 3-Li to 41-Nb. The library or retrievals of selected materials are available on magnetic tape from the IAEA Nuclear Data Section upon request. (author)

Benchmarking the new JENDL-4.0 library on criticality experiments of a research reactor with oxide LEU (20 w/o) fuel, light water moderator and beryllium reflectors

International Nuclear Information System (INIS)

Liem, Peng Hong; Sembiring, Tagor Malem

2012-01-01

Highlights: ► Benchmark calculations of the new JENDL-4.0 library. ► Thermal research reactor with oxide LEU fuel, H 2 O moderator and Be reflector. ► JENDL-4.0 library shows better C/E values for criticality evaluations. - Abstract: Benchmark calculations of the new JENDL-4.0 library on the criticality experiments of a thermal research reactor with oxide low enriched uranium (LEU, 20 w/o) fuel, light water moderator and beryllium reflector (RSG GAS) have been conducted using a continuous energy Monte Carlo code, MVP-II. The JENDL-4.0 library shows better C/E values compared to the former library JENDL-3.3 and other world-widely used latest libraries (ENDF/B-VII.0 and JEFF-3.1).

Integral test of JENDL-3.3 for fast reactors

International Nuclear Information System (INIS)

Chiba, Gou

2003-01-01

An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

Compilation of MCNP data library based on JENDL-3T and test through analysis of benchmark experiment

International Nuclear Information System (INIS)

Sakurai, K.; Sasamoto, N.; Kosako, K.; Ishikawa, T.; Sato, O.; Oyama, Y.; Narita, H.; Maekawa, H.; Ueki, K.

1989-01-01

Based on an evaluated nuclear data library JENDL-3T, a temporary version of JENDL-3, a pointwise neutron cross section library for MCNP code is compiled which involves 39 nuclides from H-1 to Am-241 which are important for shielding calculations. Compilation is performed with the code system which consists of the nuclear data processing code NJOY-83 and library compilation code MACROS. Validity of the code system and reliability of the library are certified by analysing benchmark experiments. (author)

Integral test of JENDL-3.3 on fast reactors

International Nuclear Information System (INIS)

Chiba, Gou; Hazama, Taira

2003-05-01

An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

Integral test for Np237 and Am241 cross sections in JENDL, ENDF and JEF libraries

International Nuclear Information System (INIS)

Iwasaki, Tomohiko; Unesaki, Hironobu; Kitada, Takanori

2002-01-01

Experiments using Kyoto University critical assembly (KUCA) were performed for measuring the capture and fission reaction rates of 237 Np and 241 Am. A back-to-back fission chamber was employed for the measurement of the fission rate of 237 Np and 241 Am relative to 235 U. The capture rate of 237 Np relative to 197 Au was measured by using activation method. Eleven cores, of which the spectra were changed systematically, were mocked up for the present measurements. Five cores among the eleven were utilized for the fission reaction rate measurement. The experiment was analyzed using the Monte Carlo code MVP, the transport code TWOTRAN and the diffusion code CITATION using the libraries based on JENDL3.2, ENDF/B-VI and JEF2.2. As the results, for 237 Np, JENDL3.2 showed good agreement for both capture and fission. However, for the fission rate of 241 Am, JENDL3.2 underestimates 15-20%. On the other hand, ENDF/B-VI and JEF2.2 show different C/Es for 237 Np and 241 Am. (author)

Skyshine analysis using various nuclear data files

International Nuclear Information System (INIS)

Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N.; Nomura, Y.; Tsubosaka, A.

2000-01-01

The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)

Use of the nuclear data in the reactor dosimetry: limitations, improvements and perspectives

International Nuclear Information System (INIS)

Destouches, C.; Domergue, C.; Beretz, D.; Gregoire, G.; Habert, B.

2008-01-01

This paper has briefly presented the main steps of the reactor dosimetry process. Each step, activity measurement, reaction rate derivation and neutron spectrum unfolding process has been analysed from a dosimetry point of view and improvements are asked in the knowledge of the following issues. First, for activity measurements: attenuation coefficients (μ/ρ) especially for X measurements, emission intensity I γ for 115m In and I X for 103m Rh. Secondly, for cross sections: improvements are useful but before, a new in-depth analysis of the recent releases of Endf-B7, JEFF3, EAF2005, JENDL3.3 should be done with a focus on the covariance matrices. Results of this study should give place to a new release of the IRDF2002 evaluation. Thirdly, for fission spectrum modelling: a study of the fission spectra modelling would be interesting to improve the accuracy of both the experimental benchmarks and of neutron core spectrum calculations by neutron codes. Fourthly, for benchmarks: an extension of the database with new reactions and integral measurements in fission spectra ( 252 Cf, 235 U, or mixed 239 Pu- 235 U) should be performed

Development of libraries for ORIGEN2 code based on JENDL-3.2

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Ishikawa, Makoto; Ohkawachi, Yasushi

1998-03-01

The working Group of JNDC `Nuclide Generation Evaluation` has launched a project to make libraries for ORIGEN2 code based on the latest nuclear data library `JENDL-3.2` for current design of LWR and FBR fuels. Many of these libraries are under validation. (author)

Generation, Testing, and Validation of a WIMS-D/4 Multigroup Cross-Section Library Based on the JENDL-3.2 Nuclear Data

International Nuclear Information System (INIS)

Rahman, Mafizur; Takano, Hideki

2001-01-01

A new 69-group library of multigroup constants for the lattice code WIMS-D/4 has been generated with an improved resonance treatment, processing nuclear data from JENDL-3.2 by NJOY91.108. A parallel ENDF/B-VI based library has also been constructed for intercomparison of results. Benchmark calculations for a number of thermal reactor critical assemblies of both uranium and plutonium fuels have been performed with the code WIMS-D/4.1 with its three different libraries: the original WIMS library (NEA-0329/10) and the new ENDF/B-VI and JENDL-3.2 based libraries. The results calculated with both ENDF and JENDL based libraries show a similar tendency and are found in better agreement with the experimental values. Benchmark parameters are further calculated with the comprehensive lattice code SRAC95. The results from SRAC95 and WIMS-D/4.1 (both using JENDL-3.2 based libraries) agree well with each other. The new library is also verified for its applicability to mixed-oxide cores of varying plutonium contents

Software tool for portal dosimetry research.

Science.gov (United States)

Vial, P; Hunt, P; Greer, P B; Oliver, L; Baldock, C

2008-09-01

This paper describes a software tool developed for research into the use of an electronic portal imaging device (EPID) to verify dose for intensity modulated radiation therapy (IMRT) beams. A portal dose image prediction (PDIP) model that predicts the EPID response to IMRT beams has been implemented into a commercially available treatment planning system (TPS). The software tool described in this work was developed to modify the TPS PDIP model by incorporating correction factors into the predicted EPID image to account for the difference in EPID response to open beam radiation and multileaf collimator (MLC) transmitted radiation. The processes performed by the software tool include; i) read the MLC file and the PDIP from the TPS, ii) calculate the fraction of beam-on time that each point in the IMRT beam is shielded by MLC leaves, iii) interpolate correction factors from look-up tables, iv) create a corrected PDIP image from the product of the original PDIP and the correction factors and write the corrected image to file, v) display, analyse, and export various image datasets. The software tool was developed using the Microsoft Visual Studio.NET framework with the C# compiler. The operation of the software tool was validated. This software provided useful tools for EPID dosimetry research, and it is being utilised and further developed in ongoing EPID dosimetry and IMRT dosimetry projects.

Evaluation of resonance parameters of Mo, Tc, Te, Ba, La, Ce, Pr, Nd, Pm, Sm and Eu isotopes for JENDL-2 fission product file

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki; Togawa, Orihiko; Nakagawa, Tsuneo

1986-03-01

The resonance parameters of 39 fission product nuclides have been evaluated. The present work is a part of the evaluation of 100 fission product nuclei for JENDL-2 by Japanese Nuclear Data Committee. All the available experimental data were collected, stored in REPSTOR system and compared with one another. The evaluation was made on the basis of the experimental data. The precise description of the evaluation is given in this report. The presently evaluated resonance parameters are tabulated in Appendix with the experimental data. (author)

Evaluated neutronic file for indium

International Nuclear Information System (INIS)

Smith, A.B.; Chiba, S.; Smith, D.L.; Meadows, J.W.; Guenther, P.T.; Lawson, R.D.; Howerton, R.J.

1990-01-01

A comprehensive evaluated neutronic data file for elemental indium is documented. This file, extending from 10 -5 eV to 20 MeV, is presented in the ENDF/B-VI format, and contains all neutron-induced processes necessary for the vast majority of neutronic applications. In addition, an evaluation of the 115 In(n,n') 116m In dosimetry reaction is presented as a separate file. Attention is given in quantitative values, with corresponding uncertainty information. These files have been submitted for consideration as a part of the ENDF/B-VI national evaluated-file system. 144 refs., 10 figs., 4 tabs

General activities of JAERI nuclear data center and Japanese nuclear data committee

International Nuclear Information System (INIS)

Fukahori, Tokio

1999-01-01

The nuclear data center of Japan Atomic Energy Research Institute (JAERI/NDC) is playing the role of Japanese domestic nuclear data center and gateway to foreign data centers. As the domestic nuclear data center, activities of JAERI/NDC are 1) compiling the Japanese Evaluated Nuclear Data Library (JENDL) for both general and special purposes, 2) importing and exporting nuclear data, 3) nuclear data services for the domestic users, and 4) organizing japanese Nuclear Data Committee (JNDC) as a secretariat. Compiled JENDL General Purpose Files up to now are JENDL-1, 2, 3, 3.1 and 3.2. The data for 340 nuclei in the energy range from 10 -5 eV to 20 MeV are available in JENDL-3.2. JENDL Special Purpose Files were also prepared in order to meet the requests from the specified application fields. JNDC has about 140 members. JNDC consists of Main Committee, Steering Committee, Subcommittee on Nuclear Data, Subcommittee on Reactor Constants, Subcommittee on Nuclear Fuel Cycle and Standing Groups. Above subcommittees are performing essential evaluation for the files described above, checking the JENDL files through the benchmark and integral testing as well as considering the standard group constant, and considering about evaluation of decay heat and nuclide generation/depletion and fission product yields. (author)

Skyshine analysis using various nuclear data files

Energy Technology Data Exchange (ETDEWEB)

Zharkov, V.P.; Dikareva, O.F.; Kartashev, I.A.; Kiselev, A.N. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation); Nomura, Y.; Tsubosaka, A. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2000-03-01

The calculations of the spacial distributions of dose rate for neutron and secondary photons, thermal neutron fluxes and space-energy distributions of neutron and photons near the air-ground interface were performed by MCNP and DORT codes. Different nuclear data files were used (ENDF/B-IV, ENDF/B-VI, FENDL-2, JENDL-3.2). Either the standard pointwise libraries (MCNP) or special libraries prepared by NJOY code from ENDF/B and others' files were used. Prepared multigroup coupled neutron and photon cross sections libraries for DORT code had CASK-40 group energy structures. The libraries contain pointwise or multigroup cross sections data for all elements included in the atmosphere and ground composition. The validation of the calculated results was performed with using the experimental data obtained for the series of measurements at RA reactor. (author)

Benchmark test of nuclear data with pulsed sphere experiment using OKTAVIAN

International Nuclear Information System (INIS)

Ichihara, Ch.; Hayashi, S.; Yamamoto, J.; Kimura, I.; Tkahashi, A.

1996-01-01

Nuclear data files such as JENDL - Fusion File, JENDL - 3.2, ENDF / B -VI, BROND - 2 have been compared to pulsed sphere experiments on 14 samples modeled with the MCNP4A code for the purpose of bench-marking the data libraries. The results are in good agreement for Li, Cr, Mn, Cu and Mo. Satisfying results have been obtained for Zr and Nb with JENDL - Fusion and JENDL - 3.2. For W, quite good results were obtained with ENDF / B - VI. There is a disagreement for the other samples such as LiF, TEFLON, Si, Ti and Co

JENDL-4.0: A new library for nuclear science and engineering

International Nuclear Information System (INIS)

Shibata, Keiichi; Iwamoto, Osamu; Nakagawa, Tsuneo; Iwamoto, Nobuyuki; Ichihara, Akira; Kunieda, Satoshi; Chiba, Satoshi; Furutaka, Kazuyoshi; Katakura, Jun-ichi; Otuka, Naohiko; Ohsawa, Takaaki; Murata, Toru; Matsunobu, Hiroyuki; Zukeran, Atsushi; Kamada, So

2011-01-01

The fourth version of the Japanese Evaluated Nuclear Data Library has been produced in cooperation with the Japanese Nuclear Data Committee. In the new library, much emphasis is placed on the improvements of fission product and minor actinoid data. Two nuclear model codes were developed in order to evaluate the cross sections of fission products and minor actinoids. Coupled-channel optical model parameters, which can be applied to wide mass and energy regions, were obtained for nuclear model calculations. Thermal cross sections of actinoids were carefully examined by considering experimental data or by the systematics of neighboring nuclei. Most of the fission cross sections were derived from experimental data. A simultaneous evaluation was performed for the fission cross sections of important uranium and plutonium isotopes above 10 keV. New evaluations were performed for the thirty fission-product nuclides that had not been contained in the previous library JENDL-3.3. The data for light elements and structural materials were partly reevaluated. Moreover, covariances were estimated mainly for actinoids. The new library was released as JENDL-4.0, and the data can be retrieved from the Web site of the JAEA Nuclear Data Center. (author)

Comparison of 235U fission cross sections in JENDL-3.3 and ENDF/B-VI

International Nuclear Information System (INIS)

Kawano, Toshihiko; Carlson, Allan D.; Matsunobu, Hiroyuki; Nakagawa, Tsuneo; Shibata, Keiichi

2002-01-01

Comparisons of evaluated fission cross sections for 235 U in JENDL-3.3 and ENDF/B-VI are carried out. The comparisons are made for both the differential and integral data. The fission cross sections as well as the fission ratios are compared with the experimental data in detail. Spectrum averaged cross sections are calculated and compared with the measurements. The employed spectra are the 235 U prompt fission neutron spectrum, the 252 Cf spontaneous fission neutron spectrum, and the neutron spectrum produced by a 9 Be(d, xn) reaction. For 235 U prompt fission neutron spectrum, the ENDF/B-VI evaluation reproduces experimental averaged cross sections. For 252 Cf and 9 Be(d, xn) neutron spectra, the JENDL-3.3 evaluation gives better results than ENDF/B-VI. (author)

Evaluated nuclear-data file for niobium

International Nuclear Information System (INIS)

Smith, A.B.; Smith, D.L.; Howerton, R.J.

1985-03-01

A comprehensive evaluated nuclear-data file for elemental niobium is provided in the ENDF/B format. This file, extending over the energy range 10 -11 -20 MeV, is suitable for comprehensive neutronic calculations, particulary those dealing with fusion-energy systems. It also provides dosimetry information. Attention is given to the internal consistancy of the file, energy balance, and the quantitative specification of uncertainties. Comparisons are made with experimental data and previous evaluated files. The results of integral tests are described and remaining outstanding problem areas are cited. 107 refs

Radiation Litigation and Internal Dosimetry

International Nuclear Information System (INIS)

Jose, D.E.

1987-01-01

Radiation Litigation refers to those lawsuits filed by individuals who claim to have been injured by some past exposure to ionizing radiation. Law classifies these cases as personal injury or tort cases. However, they are a new breed of such cases and the law is presently struggling with whether these cases can be resolved using the traditional methods of legal analysis or whether new forms of analysis, such as probability of causation, need to be applied. There are no absolutely certain rules concerning how these particular lawsuits will be tried and analyzed. The United States presently is defending cases filed by approximately 7000 plaintiffs. The private nuclear industry is defending cases filed by over 2000 plaintiffs. While not all of these cases will actually be tried on their merits, at least some will and internal dosimetry will play a very important part in many of these trials

Measurement of 89Y(n,2n) spectral averaged cross section in LR-0 special core reactor spectrum

Science.gov (United States)

Košťál, Michal; Losa, Evžen; Baroň, Petr; Šolc, Jaroslav; Švadlenková, Marie; Koleška, Michal; Mareček, Martin; Uhlíř, Jan

2017-12-01

The present paper describes reaction rate measurement of 89Y(n,2n)88Y in a well-defined reactor spectrum of a special core assembled in the LR-0 reactor and compares this value with results of simulation. The reaction rate is derived from the measurement of activity of 88Y using gamma-ray spectrometry of irradiated Y2O3 sample. The resulting cross section value averaged in spectrum is 43.9 ± 1.5 μb, averaged in the 235U spectrum is 0.172 ± 0.006 mb. This cross-section is important as it is used as high energy neutron monitor and is therefore included in the International Reactor Dosimetry and Fusion File. Calculations of reaction rates were performed with the MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JEFF-3.2, JENDL-3.3, JENDL-4, ROSFOND-2010, CENDL-3.1 and IRDFF nuclear data libraries. The agreement with uranium description by CIELO library is very good, while in ENDF/B-VII.0 description of uranium, underprediction about 10% in average can be observed.

Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO2 and MOX spent fuels

International Nuclear Information System (INIS)

Sasahara, Akihiro; Matsumura, Tetsuo; Nicolaou, G.; Betti, M.; Walker, C.T.

1997-01-01

The post irradiation examinations (PIE) were carried out for high burn-up UO 2 spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, a) isotopic analysis, b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

Impact of up-to-date evaluated nuclear data files on the Monte-Carlo analysis results of metallic fueled BFS critical assemblies

International Nuclear Information System (INIS)

Yoo, Jaewoon; Kim, Do-Heon; Kim, Sang-Ji; Kim, Yeong-Il

2009-01-01

Three metallic fueled BFS critical assemblies, BFS-73-1, BFS-75-1, and BFS-55-1 were analyzed by using the Monte-Carlo analysis code MCNP4C with five different evaluated data files, ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, JENDL-AC and ENDF/B-VI.6. The impacts of microscopic cross sections in the up-to-date evaluated nuclear data files were clarified by the analyses. The update of Zr cross section leads to the calculated k-effective lower than that of ENDF/B-VI.6. The revision of U-238 inelastic scattering cross section makes large difference in the predicted k-effectives between the libraries, which depends on the amount of the contribution of the inelastic cross sections change and the compensation of other reaction types. The results of the spectral indices and reaction rate ratios shows the improvement of the up-to-date evaluated nuclear data files for the U-238, Np-237, Pu-240 fission reactions, however, there are still need of further improvement for other minor actinide cross sections. The heterogeneity effects involved on the k-effective and relative fission rate distribution were evaluated in this study, which can be used as the correction factor for constructing the homogeneous benchmark configuration while keeping the consistency with the actual critical experiment. (author)

Dosimetry; La dosimetrie

Energy Technology Data Exchange (ETDEWEB)

Le Couteulx, I.; Apretna, D.; Beaugerie, M.F. [Electricite de France (EDF), 75 - Paris (France)] [and others

2003-07-01

Eight articles treat the dosimetry. Two articles evaluate the radiation doses in specific cases, dosimetry of patients in radiodiagnosis, three articles are devoted to detectors (neutrons and x and gamma radiations) and a computer code to build up the dosimetry of an accident due to an external exposure. (N.C.)

Benchmark validation by means of pulsed sphere experiment at OKTAVIAN

Energy Technology Data Exchange (ETDEWEB)

Ichihara, Chihiro [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst.; Hayashi, S.A.; Kimura, Itsuro; Yamamoto, Junji; Takahashi, Akito

1997-03-01

The new version of Japanese nuclear data library JENDL-3.2 has recently been released. JENDL Fusion File which adopted DDX representations for secondary neutrons was also improved with the new evaluation method. On the other hand, FENDL nuclear data project to compile nuclear data library for fusion related research has been conducted partly under auspices of International Atomic Energy Agency (IAEA). The first version FENDL-1 consists of JENDL-3.1, ENDF/B-VI, BROND-2 and EFF-1 and has been released in 1995. The work for the second version FENDL-2 is now ongoing. The Bench mark validation of the nuclear data libraries have been performed to help selecting the candidate for the FENDL-2. The benchmark experiment have been conducted at OKTAVIAN of Osaka university. The sample spheres were constructed by filling the spherical shells with sample. The leakage neutron spectra from sphere piles were measured with a time-of-flight method. The measured spectra were compared with the theoretical calculation using MCNP 4A and the processed libraries from JENDL-3.1, JENDL-3.2, JENDL Fusion File, and FENDL-1. JENDL Fusion File and JENDL-3.2 gave almost the same prediction for the experiment. And both prediction are almost satisfying for Li, Cr, Mn, Cu, Zr, Nb and Mo, whereas for Al, LiF, CF2, Si, Ti, Co and W there is some discrepancy. However, they gave better prediction than the calculations using the library from FENDL-1, except for W. (author)

Isotopic analyses and calculation by use of JENDL-3.2 for high burn-up UO{sub 2} and MOX spent fuels

Energy Technology Data Exchange (ETDEWEB)

Sasahara, Akihiro; Matsumura, Tetsuo [Central Research Inst. of Electric Power Industry, Komae, Tokyo (Japan). Komae Research Lab.; Nicolaou, G.; Betti, M.; Walker, C.T.

1997-03-01

The post irradiation examinations (PIE) were carried out for high burn-up UO{sub 2} spent fuel (3.8%U235, average burn-up:60GWd/t) and mixed oxide (MOX) spent fuel (5.07%Pu, average burn-up:45GWd/t). The PIE includes, (a) isotopic analysis, (b) electron probe microanalysis (EPMA) in pellet cross section and so on. The results of isotopic analyses and EPMA were compared with ORIGEN2/82 and VIM-BURN calculation results. In VIM-BURN calculation, the nuclear data of actinides were proceeded from new data file, JENDL-3.2. The sensitivities of power history and moderator density to nuclides composition were investigated by VIM-BURN calculation and consequently power history mainly effected on Am241 and Am242m and moderator density effected on fissile nuclides. From EPMA results of U and Pu distribution in pellet, VIM-BURN calculation showed reasonable distribution in pellet cross section. (author)

Automated four-dimensional Monte Carlo workflow using log files and real-time motion monitoring

DEFF Research Database (Denmark)

Sibolt, Patrik; Cronholm, R.O.; Heath, E.

2017-01-01

been developed in python. The workflow has been tested and verified using an in-house developed dosimetry system comprised of a dynamic thorax phantom constructed for plastic scintillator dosimetry. The workflow is directly compatible with any treatment planning system and can also be triggered...... by the appearance of linac log files. It has minimum user interaction and, with the use of linac log files, it provides a method for verification of the actually delivered dose in the patient geometry....

Comparison of {sup 235}U fission cross sections in JENDL-3.3 and ENDF/B-VI

Energy Technology Data Exchange (ETDEWEB)

Kawano, Toshihiko [Kyushu Univ., Fukuoka (Japan); Carlson, Allan D. [National Institute of Standards and Technology (United States); Matsunobu, Hiroyuki [Data Engineering, Inc., Fujisawa, Kanagawa (Japan); Nakagawa, Tsuneo; Shibata, Keiichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Talou, Patrick; Young, Philip G.; Chadwick, Mark B. [Los Alamos National Laboratory, Los Alamos, NM (United States)

2002-01-01

Comparisons of evaluated fission cross sections for {sup 235}U in JENDL-3.3 and ENDF/B-VI are carried out. The comparisons are made for both the differential and integral data. The fission cross sections as well as the fission ratios are compared with the experimental data in detail. Spectrum averaged cross sections are calculated and compared with the measurements. The employed spectra are the {sup 235}U prompt fission neutron spectrum, the {sup 252}Cf spontaneous fission neutron spectrum, and the neutron spectrum produced by a {sup 9}Be(d, xn) reaction. For {sup 235}U prompt fission neutron spectrum, the ENDF/B-VI evaluation reproduces experimental averaged cross sections. For {sup 252}Cf and {sup 9}Be(d, xn) neutron spectra, the JENDL-3.3 evaluation gives better results than ENDF/B-VI. (author)

Development of a common nuclear group constants library system: JSSTDL-295n-104γ based on JENDL-3 nuclear data library

International Nuclear Information System (INIS)

Hasegawa, A.

1992-01-01

JSSTDL 295n-104γ: A common group cross-section library system has been developed in JAERI to be used in fairly wide range of applications in nuclear industry. This system is composed of a common 295n-104γ group cross-section library based on JENDL-3 nuclear data file and its utility codes. Target of this system is focused to the criticality or shielding calculations in fast and fusion reactors using ANISN, DOT, or MORSE code. Specifications of the common group constants were decided responding to the request from various nuclear data users, particularly from nuclear design group in Japan. Group structure is decided so as to cover almost all group structures currently used in our country. This library includes self-shielding factor tables for primary reactions. A routine for generating macro-scopic cross-section using the self-shielding factor table is also provided. Neutron cross-sections and photon production cross-sections are processed by Prof. GROUCH-G/B code system and γ ray transport cross-sections are generated by GAMLEG-JR. In this paper, outline and present status of the JSSTDL library system is described along with two examples adopted in JENDL-3 benchmark test. One is for shielding calculation, where effects of self-shielding factor (f-table) is shown in conjunction with the analysis of the ASPIS natural iron deep penetration experiment. Without considering resonance self-shielding effect in resonance energy region for resonant nuclides like iron, the results is completely missled in the attenuation profile calculation in the shields. The other example is fast rector criticality calculations of very small critical assemblies with very high enrichment fuel materials where some basic characteristics of this library is presented. (orig.)

JENDL. Nuclear databases for science and technology

International Nuclear Information System (INIS)

Shibata, Keiichi

2013-01-01

It is exactly 50 years since the Japanese Nuclear Data Committee was founded both in the Atomic Energy Society of Japan and in the former Japan Atomic Energy Research Institute. The committee promoted the development of Japan's own evaluated nuclear data libraries. As a result, we managed to produce a series of Japanese Evaluated Nuclear Data Libraries (JENDLs) to be used in various fields for science and technology. The libraries are categorized into general-purpose and special-purpose ones. The general-purpose libraries have been updated periodically by considering the latest knowledge on experimental and theoretical nuclear physics that was available at the time of the updates. On the other hand, the special-purpose libraries have been issued in order to meet the needs for particular application fields. This paper reviews the research and development for those libraries. (author)

Titanium-II: an evaluated nuclear data file

International Nuclear Information System (INIS)

Philis, C.; Howerton, R.; Smith, A.B.

1977-06-01

A comprehensive evaluated nuclear data file for elemental titanium is outlined including definition of the data base, the evaluation procedures and judgments, and the final evaluated results. The file describes all significant neutron-induced reactions with elemental titanium and the associated photon-production processes to incident neutron energies of 20.0 MeV. In addition, isotopic-reaction files, consistent with the elemental file, are separately defined for those processes which are important to applied considerations of material-damage and neutron-dosimetry. The file is formulated in the ENDF format. This report formally documents the evaluation and, together with the numerical file, is submitted for consideration as a part of the ENDF/B-V evaluated file system. 20 figures, 9 tables

CASMO5 JENDL-4.0 and ENDF/B-VII.1beta4 libraries

International Nuclear Information System (INIS)

Rhodes, J.; Gheorghiu, N.; Ferrer, R.

2012-01-01

This paper details the generation of neutron data libraries for the CASMO5 lattice physics code based on the recently released JENDL-4.0 and ENDF/B-VII.1beta4 nuclear data evaluations. This data represents state-of-the-art nuclear data for late-2011. The key features of the new evaluations are briefly described along with the procedure for processing of this data into CASMO5, 586-energy group neutron data libraries. Finally some CASMO5 results for standard UO 2 and MOX critical experiments for the two new libraries and the current ENDF/B-VII.0 CASMO5 library are presented including the B and W 1810 series, DIMPLE S06A, S06B, TCA reflector criticals with iron plates and the PNL-30-35 MOX criticals. The results show that CASMO5 with the new libraries is performing well for these criticals with a very slight edge in results to the JENDL-4.0 nuclear data evaluation over the ENDF/B-VII.1beta4 evaluation. Work is currently underway to generate a CASMO5 library based on the final ENDF/B-VII.R1 evaluation released Dec. 22, 2011. (authors)

Evaluation of the 115In(n,n)/sup 115m/In reaction for the ENDF/B-V dosimetry file

International Nuclear Information System (INIS)

Smith, D.L.

1976-12-01

An evaluation of the 115 In(n,n')/sup 115m/In reaction for the ENDF/B-V Dosimetry File is presented. This evaluation is based entirely on reported experimental differential data. Several data sets were renormalized prior to the evaluation in order to take into account recent adjustments in corresponding standard cross sections and in other nuclear parameters used for derivation of cross sections. The present evaluation is compared with the corresponding ENDF/B-IV evaluation. The value of the spectrum-average cross section for the standard neutron field resulting from thermal-neutron fission of 235 U has been computed for this reaction using cross section values from the present evaluation. This computed cross section compares favorably with the result of a recent evaluation of integral data

Evaluation of nuclear data for neutron dosimetry

International Nuclear Information System (INIS)

Tardelli, Tiago Cardoso

2013-01-01

Absorbed dose and Effective dose are usually calculated using radiation transport computer codes. The quality of the calculations of absorbed dose depends on nuclear data utilized, however, there are rare information about the differences in dose caused by the use of different libraries. The objective of this study is to compare dose values obtained using different nuclear data libraries due to external source of neutrons in the energy range from 10-11 to 20 MeV. The nuclear data libraries used are: JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. Dose calculations were carried out with the MCNPX code considering the anthropomorphic ICRP 110 model. The differences in the absorbed dose values using JEFF 3.3.1 and ENDF/B.VII libraries are small, around 1%, but the results obtained with JENDL 4.0 presented differences up to 85% compared to ENDF and JEFF results. Differences in effective dose values are around 1.5% between ENDF and JEFF and 11% between ENDF/B.VII and JENDL 4.0. (author)

ORIGEN-2 libraries based on JENDL-3.2 for PWR-MOX fuel

Energy Technology Data Exchange (ETDEWEB)

Matsumoto, Hideki; Onoue, Masaaki; Tahara, Yoshihisa [Mitsubishi Heavy Industries Ltd., Tokyo (Japan)

2001-08-01

A set of ORIGEN-2 libraries for PWR MOX fuel was developed based on JENDL-3.2 in the Working Group on Evaluation of Nuclide Production, Japanese Nuclear Data Committee. The calculational model generating ORIGEN-2 libraries of PWR MOX is explained here in detail. The ORIGEN-2 calculation with the new ORIGEN-2 MOX library can predict the nuclides contents within 10% for U and Pu isotopes and 20% for both minor actinides and main FPs. (author)

Topics in radiation dosimetry radiation dosimetry

CERN Document Server

1972-01-01

Radiation Dosimetry, Supplement 1: Topics in Radiation Dosimetry covers instruments and techniques in dealing with special dosimetry problems. The book discusses thermoluminescence dosimetry in archeological dating; dosimetric applications of track etching; vacuum chambers of radiation measurement. The text also describes wall-less detectors in microdosimetry; dosimetry of low-energy X-rays; and the theory and general applicability of the gamma-ray theory of track effects to various systems. Dose equivalent determinations in neutron fields by means of moderator techniques; as well as developm

Acute ingestion dosimetry using the ICRP 30 gastrointestinal tract model

International Nuclear Information System (INIS)

Cassels, B.M.

1987-01-01

This paper examines the gastrointestinal (GI) tract model used for dosimetry as outlined in ICRP30, to allow quick calculations of effective dose equivalents for acute radionuclide ingestion. A computer program has been developed to emulate the GI tract model. The program and associated data files are structured so that the GI tract model parameters can be varied, while the file structure and algorithm for the GI tract model should require minimal modification to allow the same theories that apply in this model to be used for other dosimetric models

Analysis of Sodium-23 Data Cross-Sections for Coolant on Generation IV Reactor - SFR

International Nuclear Information System (INIS)

Suwoto; Zuhair

2009-01-01

The integral tests of sodium-23 neutron cross-sections for coolant contained in JENDL-3.3, ENDF/B-VII.0, BROND-2.2 and JEFF-3.1 files have been performed. Cross-sections analysis of sodium-23 such as total cross-sections, elastic scattering, in-elastic scattering and radiative capture cross-sections for several temperature i.e. 300K, 800K and 1500K. The sodium-23 total cross-sections analysis based on MAEKER, R.E. experimental result through Broomstick experiment calculation. Differences between among other evaluated nuclear data file for elastic scattering, in-elastic scattering and radiative capture cross-sections were done analyzed and compared to ENDF/B-VII.0 as standard reference. Analysis of total cross-sections sodium-23 through broomstick calculation show JENDL-3.3 file give the best result on C/E statistical average value is 1.1043 compared to another nuclear data files. Differences sodium-23 total cross-sections on JEFF-3.1 file for all temperature work specially for energy 40keV and 1MeV-2MeV is about 0.2%. Meanwhile, relative small differences on in-elastic total scattering cross-sections are shown for all temperatures are about ±0.1% in JENDL-3.3. On the other hand, BROND-2.2 file give ±6% higher on sodium-23 in-elastic total scattering cross sections for energy range 450keV-550keV. Clearly significant differences on sodium-23 radiative capture cross sections for BROND-2.2 file especially in energy 109.659keV is somewhat higher than 446%, otherwise JENDL-3.3 and JEFF-3.1 give 16% higher than ENDF/B-VII.0 file. Overall result show that JENDL-3.3, ENDFB-VII.0, BROND-2.2 and JEFF-3.1 have little bit differences in total, elastic scattering, in-elastic scattering total cross sections, except BROND-2.2 file due to radiative capture cross-sections with larger discrepancies. (author)

Evaluation and improvement of cross section accuracy for most important dosimetry reactions 27Al(n,p), 56Fe(n,p) and 237Np(n,f) including covariance data

International Nuclear Information System (INIS)

Zolotarev, K.I.

2004-02-01

New evaluations of cross sections and their uncertainties for dosimetry reactions 27 Al(n,p) , 56 Fe(n,p) and 237 Np(n,f) have been carried out in the frame work of IAEA Research Contract No. 11372/RB. Data files prepared for this reactions in the ENDF-6 format may be consider as candidates for the new International Reactor Dosimetry File: IRDF-2002. (author)

[Automatic Extraction and Analysis of Dosimetry Data in Radiotherapy Plans].

Science.gov (United States)

Song, Wei; Zhao, Di; Lu, Hong; Zhang, Biyun; Ma, Jun; Yu, Dahai

To improve the efficiency and accuracy of extraction and analysis of dosimetry data in radiotherapy plans for a batch of patients. With the interface function provided in Matlab platform, a program was written to extract the dosimetry data exported from treatment planning system in DICOM RT format and exported the dose-volume data to an Excel file with the SPSS compatible format. This method was compared with manual operation for 14 gastric carcinoma patients to validate the efficiency and accuracy. The output Excel data were compatible with SPSS in format, the dosimetry data error for PTV dose interval of 90%-98%, PTV dose interval of 99%-106% and all OARs were -3.48E-5 ± 3.01E-5, -1.11E-3 ± 7.68E-4, -7.85E-5 ± 9.91E-5 respectively. Compared with manual operation, the time required was reduced from 5.3 h to 0.19 h and input error was reduced from 0.002 to 0. The automatic extraction of dosimetry data in DICOM RT format for batch patients, the SPSS compatible data exportation, quick analysis were achieved in this paper. The efficiency of clinical researches based on dosimetry data analysis of large number of patients will be improved with this methods.

Validation of actinide nuclear data from ENDF/B-V, INDL/A-83 and JENDL-2

International Nuclear Information System (INIS)

Paviotti Corcuera, R.; Moraes, M. de

1988-11-01

Resonance integrals and fission spectrum averaged cross sections are calculated for the actinides of ENDF/B-V, INDL/A-83 and JENDL-2. The results are compared with each other and with experimental data when available. The experimental data are scarce and there exist large differences among data from different libraries. (author). 16 refs, 4 tabs

Evaluation and improvement of cross section accuracy for most important dosimetry reactions 27Al(n,p), 56Fe(n,p) and 237Np(n,f) including covariance data

Energy Technology Data Exchange (ETDEWEB)

Zolotarev, K I [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

2004-02-01

New evaluations of cross sections and their uncertainties for dosimetry reactions {sup 27}Al(n,p) , {sup 56}Fe(n,p) and {sup 237}Np(n,f) have been carried out in the frame work of IAEA Research Contract No. 11372/RB. Data files prepared for this reactions in the ENDF-6 format may be consider as candidates for the new International Reactor Dosimetry File: IRDF-2002. (author)

JENDL-3.2 performance in analyses of MISTRAL critical experiments for high-moderation MOX cores

International Nuclear Information System (INIS)

Takada, Naoyuki; Hibi, Koki; Ishii, Kazuya; Ando, Yoshihira; Yamamoto, Toru; Ueji, Masao; Iwata, Yutaka

2001-01-01

NUPEC and CEA have launched an extensive experimental program called MISTRAL to study highly moderated MOX cores for the advanced LWRs. The analyses using SRAC system and MVP code with JENDL-3.2 library are in progress on the experiments of the MISTRAL and the former EPICURE programs. Various comparisons have been made between calculation results and measurement values. (author)

Individual dosimetry and calibration

International Nuclear Information System (INIS)

Otto, T.

1997-01-01

In 1996, the Dosimetry and Calibration Section was, as in previous years, mainly engaged in routine tasks: the distribution of over 6000 dosimeters (with a total of more than 10,000 films) every two months and the calibration of about 900 fixed and mobile instruments used in the radiation survey sections of RP group. These tasks were, thanks to an experienced team, well mastered. Special efforts had to be made in a number of areas to modernize the service or to keep it in line with new prescriptions. The Individual Dosimetry Service had to assure that CERN's contracting firms comply with the prescriptions in the Radiation Safety Manual (1996) that had been inspired by the Swiss Ordinance of 1994: Companies must file for authorizations with the Swiss Federal Office for Public Health requiring that in every company an 'Expert in Radiation Protection' be nominated and subsequently trained. CERN's Individual Dosimetry Service is accredited by the Swiss Federal Authorities and works closely together with other, similar services on a rigorous quality assurance programme. Within this framework, CERN was mandated to organize this year the annual Swiss 'Intercomparison of Dosimeters'. All ten accredited dosimetry services - among others those of the Paul Scherrer Institute (PSI) in Villigen and of the four Swiss nuclear power stations - sent dosimeters to CERN, where they were irradiated in CERN's calibration facility with precise photon doses. After return to their origin they were processed and evaluated. The results were communicated to CERN and were compared with the originally given doses. A report on the results was subsequently prepared and submitted to the Swiss 'Group of Experts on Personal Dosimetry'. Reference monitors for photon and neutron radiation were brought to standard laboratories to assure the traceability of CERN's calibration service to the fundamental quantities. For photon radiation, a set of ionization chambers was calibrated in the reference field

JAERI's activities on photon production data

International Nuclear Information System (INIS)

Shibata, Keiichi; Maekawa, Fujio; Niita, Koji

1996-01-01

Summarized are activities on photon production data at JAERI. The activities consists of evaluation of photon production data for JENDL Fusion File, benchmark tests of JENDL and FENDL-1 data, and calculation of photon production data in the framework of the Quantum Molecular Dynamics. The capture cross sections of 12 C and 16 O were evaluated for JENDL Fusion File by taking account of the direct radiative capture calculations obtained by A. Mengoni (ENEA). The presently evaluated data are in good agreement with the measurements of Igashira et al. in the keV region, describing the behaviour of p-wave capture which is in proportion to υ. Photon production data on Fe and Ni were updated for JENDL Fusion File by using a statistical-model calculations. According to the results of benchmark tests, the calculations with the updated data reproduce the integral measurements on gamma-ray heating. Benchmark tests of evaluated photon production data have been continued by analyzing the integral experimental performed at OKTAVIAN and FNS. The calculations with JENDL Fusion File are in good agreement with the integral measurements. Preliminary calculation of photon production data in the high energy region has been done in the framework of the Quantum Molecular Dynamics approach. The quasi-deuteron model was used to describe photon absorption in the low energy region. Above pion production threshold, pion production channels were included in the calculation. The neutron-proton bremsstrahlung obtained with the one-boson-exchange model was incorporated into QMD codes. (Abstract only)

Status of the ENDF/B special applications files

International Nuclear Information System (INIS)

Stewart, L.

1977-01-01

The newly formed SAFE Subcommittee of the Cross Section Evaluation Working Group is charged with the responsibility for providing, reviewing, and testing several ENDF/B special purpose evaluated files. This responsibility currently encompasses dosimetry, activation, hydrogen and helium production, and radioactive decay data required by a variety of users. New formats have been approved by CSEWG for the inclusion of the activation and hydrogen and helium production cross-section libraries. The decay data will be in the same format as that already employed by the Fission Product and Actinide Subcommittee of CSEWG. While an extensive dosimetry file was available on the ENDF/B-IV library for fast reactor applications, other data are needed to extend the range of applications, especially to higher incident neutron energies. This Subcommittee has long-range plans to provide evaluated neutron interaction data that can be recommended for use in many specialized applications. 1 figure, 3 tables

Evaluations of the 58Fe(n,γ)59Fe, and 54Fe(n,p)54Mn reactions for the ENDF/B-V dosimetry file

International Nuclear Information System (INIS)

Schenter, R.E.; Schmittroth, F.; Mann, F.M.

1979-10-01

A generalized least-squares adjustment procedure was used to evaluate two important dosimetry reactions for the ENDF/B-V files. Calculations for the cross section adjustments were made with the computer code FERRET, where input data included both integral and differential experimental data results. For the 54 Fe reaction, important ratio measurements were renormalized to ENDF/B-V evaluations of 235 U(n,f), 238 U(n,f), and 56 Fe(n,p). A priori curves which are required for the calculations were obtained using Hauser-Feshbach calculations from the codes NCAP ( 58 Fe) and HAUSER-5 ( 54 Fe). Covariance matrices were also calculated and are included in the evaluations. 8 figures, 1 table

ZZ SNLRML, Dosimetry Cross-Section Recommendations

International Nuclear Information System (INIS)

1996-01-01

Description of program or function: Format: SAND-II; Number of groups: 640 group SAND-II group structure. Nuclides: Cd, B, Au, S, Ni, Li, F, Na, Mg, Al, Si, P, Sc, Ti, Mn, Fe, Co, Cu, Zn, Zr, Nb, Mo, Rh, Ag, In, I, Th, U, Np, Pu, Am. Origin: ENDF/B-VI, ENDF/B-V, IRDF-90, JENDL-3, JEF 2.2 and GLUCS data with special modifications from private communications. Weighting spectrum: flat. SNLRML is a reactor dosimetry library that draws upon all available evaluated cross section libraries and selects the best evaluation for application to research reactor spectrum determinations. Many of the components of the SNLRML come from the ENDF/B-VI and IRDF-90 (DLC-0161) libraries. The library format was selected for easy interface with spectrum determination codes such as SAND-II (CCC-0112 and LSL-M2 (PSR-233) and the new PSR-0345/SNL/SAND-II has been enhanced to interface with SNLRML. The data is recommended for spectrum determination applications and for the prediction of neutron activation of typical radiation sensor materials. The library has been tested for consistency of the cross section in wide variety of neutron environments. The results and cautions from this testing have been documented. The data has been interfaced with radiation transport codes, such as TWODANT-SYS (CCC-0547) and MCNP (CCC-0200), in order to compare calculated and measured activities for benchmark reactor experiments

Chemical dosimetry principles in high dose dosimetry

International Nuclear Information System (INIS)

Mhatre, Sachin G.V.

2016-01-01

In radiation processing, activities of principal concern are process validation and process control. The objective of such formalized procedures is to establish documentary evidence that the irradiation process has achieved the desired results. The key element of such activities is inevitably a well characterized reliable dosimetry system that is traceable to recognized national and international dosimetry standards. Only such dosimetry systems can help establish the required documentary evidence. In addition, industrial radiation processing such as irradiation of foodstuffs and sterilization of health careproducts are both highly regulated, in particular with regard to dose. Besides, dosimetry is necessary for scaling up processes from the research level to the industrial level. Thus, accurate dosimetry is indispensable

Neutron metrology file NMF-90. An integrated database for performing neutron spectrum adjustment calculations

International Nuclear Information System (INIS)

Kocherov, N.P.

1996-01-01

The Neutron Metrology File NMF-90 is an integrated database for performing neutron spectrum adjustment (unfolding) calculations. It contains 4 different adjustment codes, the dosimetry reaction cross-section library IRDF-90/NMF-G with covariances files, 6 input data sets for reactor benchmark neutron fields and a number of utility codes for processing and plotting the input and output data. The package consists of 9 PC HD diskettes and manuals for the codes. It is distributed by the Nuclear Data Section of the IAEA on request free of charge. About 10 MB of diskspace is needed to install and run a typical reactor neutron dosimetry unfolding problem. (author). 8 refs

Quantitative evaluation of patient-specific quality assurance using online dosimetry system

Science.gov (United States)

Jung, Jae-Yong; Shin, Young-Ju; Sohn, Seung-Chang; Min, Jung-Whan; Kim, Yon-Lae; Kim, Dong-Su; Choe, Bo-Young; Suh, Tae-Suk

2018-01-01

In this study, we investigated the clinical performance of an online dosimetry system (Mobius FX system, MFX) by 1) dosimetric plan verification using gamma passing rates and dose volume metrics and 2) error-detection capability evaluation by deliberately introduced machine error. Eighteen volumetric modulated arc therapy (VMAT) plans were studied. To evaluate the clinical performance of the MFX, we used gamma analysis and dose volume histogram (DVH) analysis. In addition, to evaluate the error-detection capability, we used gamma analysis and DVH analysis utilizing three types of deliberately introduced errors (Type 1: gantry angle-independent multi-leaf collimator (MLC) error, Type 2: gantry angle-dependent MLC error, and Type 3: gantry angle error). A dosimetric verification comparison of physical dosimetry system (Delt4PT) and online dosimetry system (MFX), gamma passing rates of the two dosimetry systems showed very good agreement with treatment planning system (TPS) calculation. For the average dose difference between the TPS calculation and the MFX measurement, most of the dose metrics showed good agreement within a tolerance of 3%. For the error-detection comparison of Delta4PT and MFX, the gamma passing rates of the two dosimetry systems did not meet the 90% acceptance criterion with the magnitude of error exceeding 2 mm and 1.5 ◦, respectively, for error plans of Types 1, 2, and 3. For delivery with all error types, the average dose difference of PTV due to error magnitude showed good agreement between calculated TPS and measured MFX within 1%. Overall, the results of the online dosimetry system showed very good agreement with those of the physical dosimetry system. Our results suggest that a log file-based online dosimetry system is a very suitable verification tool for accurate and efficient clinical routines for patient-specific quality assurance (QA).

100 years of solid state dosimetry and radiation protection dosimetry

International Nuclear Information System (INIS)

Bartlett, David T.

2008-01-01

The use of solid state detectors in radiation dosimetry has passed its 100th anniversary. The major applications of these detectors in radiation dosimetry have been in personal dosimetry, retrospective dosimetry, dating, medical dosimetry, the characterization of radiation fields, and also in microdosimetry and radiobiology research. In this introductory paper for the 15th International Conference, I shall speak of the history of solid state dosimetry and of the radiation measurement quantities that developed at the same time, mention some landmark developments in detectors and applications, speak a bit more about dosimetry and measurement quantities, and briefly look at the past and future

Neutron integral test of graphite cross sections in MeV energy region for the JENDL-3T through an analysis of WINFRITH shielding experiment

International Nuclear Information System (INIS)

Ueki, Kohtaro; Sakurai, Kiyoshi.

1988-01-01

The neutron integral tests of graphite cross sections in MeV neutron energy region for the ENDF/B-IV, JENDL-2, JENDL-3PR1 and -3T were performed through the Monte Carlo analysis of the graphite shielding experiment at the WINFRITH. The measured values were on the reaction rates of 115 In(n,n') 115m In, 27 Al(n,α) 24 Na, 32 S(n,p) 32 P, and 103 Rh(n,n') 103m Rh threshold detectors located in the graphite slabs, so that the experiment on the graphite was good at the integral test of neutron cross sections in MeV energy resion. (author)

A graphical user interface for diagnostic radiology dosimetry using Monte Carlo (MCNP) simulation

International Nuclear Information System (INIS)

Collins, P.J.; Gorbatkov, D.; Schultz, F.W.

2000-01-01

Monte Carlo methods (for example, MCNP, EGGS4) are the 'gold standard' for both external and internal dosimetry in humans. These powerful simulation tools are, however, general-purpose codes and consequently do not provide a simple user interface for specific dosimetry tasks. We have developed a graphical user interface, for external radiation dosimetry (diagnostic radiology) using MCNP and an anthropomorphic mathematical phantom (Adam/Eva), which enables convenient modification and processing of the MCNP input and output files. The input form displays a colour coded, 3D representation of the phantom with a superimposed 'beam' for the required x-ray projection. The phantom can be rotated through 360 degrees and a transverse section at the level of the mid-point of the beam is also displayed. Text fields enable entry of input data (beam dimensions, source position, kVp, total filtration, focus-to-skin distance). A pull-down menu enables the user to select from 22 standard radiographic views. A standard projection can be modified, or new projection data entered if required. The input program modifies the MCNP input file and initiates processing. An output form displays the organ doses, normalised to unit skin entrance dose (with backscatter) (SED). The user can also enter the SED (calculated or measured) for a particular machine, to obtain the effective dose. To validate the program, the results for a PA Chest study (80 kVp, 2.5 mm Al total filtration) were compared with NRPB data (Jones and Wall, 1985). In conclusion, a convenient and reliable graphical user interface has been developed for MCNP, which enables dosimetry calculation for a full range of diagnostic radiological studies. (author)

Verification OFENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 nuclear data libraries for criticality calculations using NEA/NSC benchmarks

International Nuclear Information System (INIS)

Bouhaddane, A.; Farkas, G.; Hascik, J.; Slugen, V.

2015-01-01

The paper presents verification of selected nuclear data libraries with the aim to apply them to fast reactor calculations. More precise results were achieved for thermal neutrons calculations. This corresponds with the demand for more precise nuclear data for fast reactors. However, fast neutron calculations show some consistency, in particular between ENDF-B/VII.1 and JENDL-4.0 nuclear data libraries. The results support the idea to prefer using newer ENDF-B/VII.1 instead of the previous version ENDF-B/VII.0. Certainly, there are still some issues to be addressed and there is potential to gain more conclusive results. Although, application of ENDF-B/VII.1 and JENDL-4.0 is expected for further calculations. (authors)

Evaluation of neutron nuclear data of 9Be for JENDL-3

International Nuclear Information System (INIS)

Shibata, Keiichi

1984-12-01

Neutron nuclear data of 9 Be have been evaluated for JENDL-3 in the energy range from 10 -5 eV to 20 MeV. Evaluated quantities are the total, elastic and inelastic scattering, photon-production, (n,γ), (n,p), (n,d), (n,t) and (n,α) reaction cross sections and the angular and energy distributions of neutrons. The total cross section below 830 keV was calculated with the R-matrix theory. The statistical model was applied to the prediction of the inelastic scattering and charged-particle emission cross sections. The (n,2n) reaction cross section is given by the sum of the inelastic scattering and (n,α 1 ) reaction cross sections. (author)

Evaluation of neutron nuclear data of 241pu for JENDL-2

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki; Sekine, Nobuo

1984-06-01

Neutron nuclear data of 241 Pu were newly evaluated for JENDL-2. Evaluated quantities are the total, elastic and inelastic scattering, fission, capture, (n,2n), (n,3n) and (n,4n) reaction cross sections, the resolved and unresolved resonance parameters, the angular and energy distributions of emitted neutrons, and the average number of neutrons emitted per fission. The simultaneous evaluation method was adopted for the fission cross section so as to keep the consistency among the main fissile and fertile material nuclides. The theoretical calculations based on the spherical optical model and the statistical model were also used, when the experimental data were not sufficient. Discussion is given on the evaluation method. (author)

Dosimetry

International Nuclear Information System (INIS)

Anon.

1990-01-01

The purpose of ionizing radiation dosimetry is the measurement of the physical and biological consequences of exposure to radiation. As these consequences are proportional to the local absorption of energy, the dosimetry of ionizing radiation is based on the measurement of this quantity. Owing to the size of the effects of ionizing radiation on materials in all of these area, dosimetry plays an essential role in the prevention and the control of radiation exposure. Its use is of great importance in two areas in particular where the employment of ionizing radiation relates to human health: radiation protection, and medical applications. Dosimetry is different for various reasons: owing to the diversity of the physical characteristics produced by different kinds of radiation according to their nature (X- and γ-photons, electrons, neutrons,...), their energy (from several keV to several MeV), the orders of magnitude of the doses being estimated (a factor of about 10 5 between diagnostic and therapeutic applications); and the temporal and spatial variation of the biological parameters entering into the calculations. On the practical level, dosimetry poses two distinct yet closely related problems: the determination of the absorbed dose received by a subject exposed to radiation from a source external to his body (external dosimetry); and the determination of the absorbed dose received by a subject owing to the presence within his body of some radioactive substance (internal dosimetry)

Dosimetry Service

CERN Multimedia

2006-01-01

Cern Staff and Users can now consult their dose records for an individual or an organizational unit with HRT. Please see more information on our web page: http://cern.ch/rp-dosimetry Dosimetry Service is open every morning from 8.30 - 12.00. Closed in the afternoons. We would like to remind you that dosimeters cannot be sent to customers by internal mail. Short-term dosimeters (VCT's) must always be returned to the Service after the use and must not be left on the racks in the experimental areas or in the secretariats. Dosimetry Service Tel. 7 2155 Dosimetry.service@cern.ch http://cern.ch/rp-dosimetry

TestDose: A nuclear medicine software based on Monte Carlo modeling for generating gamma camera acquisitions and dosimetry

Energy Technology Data Exchange (ETDEWEB)

Garcia, Marie-Paule, E-mail: marie-paule.garcia@univ-brest.fr; Villoing, Daphnée [UMR 1037 INSERM/UPS, CRCT, 133 Route de Narbonne, 31062 Toulouse (France); McKay, Erin [St George Hospital, Gray Street, Kogarah, New South Wales 2217 (Australia); Ferrer, Ludovic [ICO René Gauducheau, Boulevard Jacques Monod, St Herblain 44805 (France); Cremonesi, Marta; Botta, Francesca; Ferrari, Mahila [European Institute of Oncology, Via Ripamonti 435, Milano 20141 (Italy); Bardiès, Manuel [UMR 1037 INSERM/UPS, CRCT, 133 Route de Narbonne, Toulouse 31062 (France)

2015-12-15

Purpose: The TestDose platform was developed to generate scintigraphic imaging protocols and associated dosimetry by Monte Carlo modeling. TestDose is part of a broader project (www.dositest.com) whose aim is to identify the biases induced by different clinical dosimetry protocols. Methods: The TestDose software allows handling the whole pipeline from virtual patient generation to resulting planar and SPECT images and dosimetry calculations. The originality of their approach relies on the implementation of functional segmentation for the anthropomorphic model representing a virtual patient. Two anthropomorphic models are currently available: 4D XCAT and ICRP 110. A pharmacokinetic model describes the biodistribution of a given radiopharmaceutical in each defined compartment at various time-points. The Monte Carlo simulation toolkit GATE offers the possibility to accurately simulate scintigraphic images and absorbed doses in volumes of interest. The TestDose platform relies on GATE to reproduce precisely any imaging protocol and to provide reference dosimetry. For image generation, TestDose stores user’s imaging requirements and generates automatically command files used as input for GATE. Each compartment is simulated only once and the resulting output is weighted using pharmacokinetic data. Resulting compartment projections are aggregated to obtain the final image. For dosimetry computation, emission data are stored in the platform database and relevant GATE input files are generated for the virtual patient model and associated pharmacokinetics. Results: Two samples of software runs are given to demonstrate the potential of TestDose. A clinical imaging protocol for the Octreoscan™ therapeutical treatment was implemented using the 4D XCAT model. Whole-body “step and shoot” acquisitions at different times postinjection and one SPECT acquisition were generated within reasonable computation times. Based on the same Octreoscan™ kinetics, a dosimetry

Dosimetry system 1986

International Nuclear Information System (INIS)

Woolson, William A.; Egbert, Stephen D.; Gritzner, Michael L.

1987-01-01

In May 1983, the authors proposed a dosimetry system for use by the Radiation Effects Research Foundation (RERF) that would incorporate the new findings and calculations of the joint United States - Japan working groups on the reassessment of A-bomb dosimetry. The proposed dosimetry system evolved from extensive discussions with RERF personnel, numerous meetings of the scientists from Japan and the United States involved in the dosimetry reassessment research, and requirements expressed by epidemiologists and radiobiologists on the various review panels. The dosimetry system proposed was based on considerations of the dosimetry requirements for the normal work of RERF and for future research in radiobiology, the computerized input data on A-bomb survivors available in the RERF data base, the level of detail, precision, and accuracy of various components of the dosimetric estimates, and the computer resources available at RERF in Hiroshima. These discussions and our own experience indicated that, in light of the expansion of computer and radiation technologies and the desire for more detail in the dosimetry, an entirely new approach to the dosimetry system was appropriate. This resulted in a complete replacement of the T65D system as distinguished from a simpler approach involving a renormalization of T65D parameters to reflect the new dosimetry. The proposed dosimetry system for RERF and the plan for implementation was accepted by the Department of Energy (DOE) Working Group on A-bomb Dosimetry chaired by Dr. R.F. Christy. The dosimetry system plan was also presented to the binational A-bomb dosimetry review groups for critical comment and was discussed at joint US-Japan workshop. A prototype dosimetry system incorporating preliminary dosimetry estimates and applicable to only a limited set of A-bomb survivors was installed on the RERF computer system in the fall of 1984. This system was successfully operated at RERF and provided an initial look at the impact of

Integral test of JENDL-3.2 data by re-analysis of sample reactivity measurements at SEG and STEK facilities

International Nuclear Information System (INIS)

Dietze, Klaus

2001-01-01

Sample reactivity measurements, which have been performed at the fast-thermal coupled facilities RRR/SEG and STEK, have been re-analyzed using the JNC route for reactor calculation JENDL-3.2 // SLAROM / CITATION / PERKY. C/E-values of central reactivity worths (CRW) of FP nuclides, structural materials, and standards are given. (author)

Benchmarks solutions of the coupled systems NJOY/AMPX-II/HAMMER-TECHNION and the library JENDL-3

International Nuclear Information System (INIS)

Santos, A. dos

1991-01-01

Benchmarks calculations were performed with the newest released japanese nuclear data library JENDL-3. The calculations were performed with the methodology developed at IPEN/CNEN-SP based on a coupled NJOY/AMPX-II/HAMMER-TECHNION system. The analyses show that the long-standing problem of the overprediction of the epithermal U-238 neutron capture still remains. Besides that there is an indication that the fission cross section of U-238 might be underestimated. (author)

SU-F-T-295: MLCs Performance and Patient-Specific IMRT QA Using Log File Analysis

Energy Technology Data Exchange (ETDEWEB)

Osman, A [King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia); American University of Biuret Medical Center, Biuret (Lebanon); Maalej, N [King Fahd University of Petroleum and Minerals, Dhahran (Saudi Arabia); Jayesh, K; Abdel-Rahman, W [King Fahad Specialist Hospital-Dammam, Eastern Province (Saudi Arabia)

2016-06-15

Purpose: To analyze the performance of the multi-leaf collimators (MLCs) from the log files recorded during the intensity modulated radiotherapy (IMRT) treatment and to construct the relative fluence maps and do the gamma analysis to compare the planned and executed MLCs movement. Methods: We developed a program to extract and analyze the data from dynamic log files (dynalog files) generated from sliding window IMRT delivery treatments. The program extracts the planned and executed (actual or delivered) MLCs movement, calculates and compares the relative planned and executed fluences. The fluence maps were used to perform the gamma analysis (with 3% dose difference and 3 mm distance to agreement) for 3 IMR patients. We compared our gamma analysis results with those obtained from portal dose image prediction (PDIP) algorithm performed using the EPID. Results: For 3 different IMRT patient treatments, the maximum difference between the planned and the executed MCLs positions was 1.2 mm. The gamma analysis results of the planned and delivered fluences were in good agreement with the gamma analysis from portal dosimetry. The maximum difference for number of pixels passing the gamma criteria (3%/3mm) was 0.19% with respect to portal dosimetry results. Conclusion: MLC log files can be used to verify the performance of the MLCs. Patientspecific IMRT QA based on MLC movement log files gives similar results to EPID dosimetry results. This promising method for patient-specific IMRT QA is fast, does not require dose measurements in a phantom, can be done before the treatment and for every fraction, and significantly reduces the IMRT workload. The author would like to thank King Fahd University of petroleum and Minerals for the support.

ORIGEN2 libraries based on JENDL-3.2 for LWR-MOX fuels

Energy Technology Data Exchange (ETDEWEB)

Suyama, Kenya; Katakura, Jun-ichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Onoue, Masaaki; Matsumoto, Hideki [Mitsubishi Heavy Industries Ltd., Tokyo (Japan); Sasahara, Akihiro [Central Research Inst. of Electric Power Industry, Tokyo (Japan)

2000-11-01

A set of ORIGEN2 libraries for LWR MOX fuels was developed based on JENDL-3.2. The libraries were compiled with SWAT using the specification of MOX fuels that will be used in nuclear power reactors in Japan. The verification of the libraries were performed by the analyses of post irradiation examinations for the fuels from European PWR. By the analysis of PIE data from PWR in United States, the comparison was made between calculation and experimental results in the case of that parameters for making the libraries are different from irradiation conditions. These new libraries for LWR MOX fuels are packaged in ORLIBJ32, the libraries released in 1999. (author)

Using NJOY to Create MCNP ACE Files and Visualize Nuclear Data

Energy Technology Data Exchange (ETDEWEB)

Kahler, Albert Comstock [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-10-14

We provide lecture materials that describe the input requirements to create various MCNP ACE files (Fast, Thermal, Dosimetry, Photo-nuclear and Photo-atomic) with the NJOY Nuclear Data Processing code system. Input instructions to visualize nuclear data with NJOY are also provided.

Neutron personnel dosimetry

International Nuclear Information System (INIS)

Griffith, R.V.

1981-01-01

The current state-of-the-art in neutron personnel dosimetry is reviewed. Topics covered include dosimetry needs and alternatives, current dosimetry approaches, personnel monitoring devices, calibration strategies, and future developments

Automated four-dimensional Monte Carlo workflow using log files and real-time motion monitoring

International Nuclear Information System (INIS)

Sibolt, P; Andersen, C E; Cronholm, R O; Heath, E; Behrens, C F

2017-01-01

With emerging techniques for tracking and gating methods in radiotherapy of lung cancer patients, there is an increasing need for efficient four-dimensional Monte Carlo (4DMC) based quality assurance (QA). An automated and flexible workflow for 4DMC QA, based on the 4DdefDOSXYZnrc user code, has been developed in python. The workflow has been tested and verified using an in-house developed dosimetry system comprised of a dynamic thorax phantom constructed for plastic scintillator dosimetry. The workflow is directly compatible with any treatment planning system and can also be triggered by the appearance of linac log files. It has minimum user interaction and, with the use of linac log files, it provides a method for verification of the actually delivered dose in the patient geometry. (paper)

Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

International Nuclear Information System (INIS)

Altaf, M.H.; Badrun, N.H.; Chowdhury, M.T.

2015-01-01

Highlights: • SRAC-PIJ code and SRAC-CITATION have been utilized to model the core. • Most of the simulated results show no significant differences with references. • Thermal peak flux varies a bit due to up condition of TRIGA. • ENDF/B-VII.0 and JENDL-3.3 libraries perform well for neutronics analysis of TRIGA. - Abstract: Important kinetic parameters such as effective multiplication factor, k eff , excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7 W/cc and 170.1 W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor

Dosimetry

International Nuclear Information System (INIS)

Rezende, D.A.O. de

1976-01-01

The fundamental units of dosimetry are defined, such as exposure rate, absorbed dose and equivalent dose. A table is given of relative biological effectiveness values for the different types of radiation. The relation between the roentgen and rad units is calculated and the concepts of physical half-life, biological half-life and effective half-life are discussed. Referring to internal dosimetry, a mathematical treatment is given to β particle-and γ radiation dosimetry. The absorbed dose is calculated and a practical example is given of the calculation of the exposure and of the dose rate for a gama source [pt

Film dosimetry for IMRT: sensitivity corrections

International Nuclear Information System (INIS)

Suchowerska, N.; Hoban, P.; Davison, A.; Metcalfe, P.

2000-01-01

Full text: The trend towards conformal, dynamic and intensity modulated radiotherapy treatments has furthered the need for true integrating dosimetry. In traditional radiotherapy, film dosimetry is commonly used. The accuracy and reproducibility of film optical density as an indicator of dose, has been associated with several variables. These include the effects of film specific sensitivity, direction of exposure, chemical processing and film scanner sensitivity. In this study, a procedure is developed to account for these variables, with a particular view to film being used as a dosimeter for conformal treatments. An effective sensitometric curve was established by exposing part of a single sheet of film to known doses. All films were processed together and scanned using a DuoscanT1200 transmission scanner, resulting in 12 bit image files. The images were analysed using Osiris software and the results fitted to the modified Williamson equation: P P s (l - 10 αD ) This yields values of α [film sensitivity], and P s [saturation pixel value], allowing individual dosimetry films to be normalised to this sensitometric calibration curve. For validation, a piece of Kodak X Omat-V film was sealed in a head phantom and exposed to a total of 51 IMRT fields, delivered from 6 gantry angles. The rest of the sheet of film was resealed and exposed to four known doses, providing sensitometric data, specific to this exposure. All films were then processed, scanned and analysed as described above. Observed variations in serial films exposed to 50cGy is in the order of 9% [mean 25.0,standard deviation = 3.2]. The automatic gain of the scanner system typically contributed 4% variation and needs to be carefully monitored. Results indicate that by using the sensitometric data from each exposure, the collective errors can be minimised. The IMRT exposure results confirm that the above process is viable for use in dosimetry for conformal radiation therapy. Copyright (2000) Australasian

Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1

International Nuclear Information System (INIS)

Van Der Marck, S. C.

2012-01-01

Three nuclear data libraries have been tested extensively using criticality safety benchmark calculations. The three libraries are the new release of the US library ENDF/B-VII.1 (2011), the new release of the Japanese library JENDL-4.0 (2011), and the OECD/NEA library JEFF-3.1 (2006). All calculations were performed with the continuous-energy Monte Carlo code MCNP (version 4C3, as well as version 6-beta1). Around 2000 benchmark cases from the International Handbook of Criticality Safety Benchmark Experiments (ICSBEP) were used. The results were analyzed per ICSBEP category, and per element. Overall, the three libraries show similar performance on most criticality safety benchmarks. The largest differences are probably caused by elements such as Be, C, Fe, Zr, W. (authors)

Calculation of dosimetry parameters for fast neutron radiotherapy

Energy Technology Data Exchange (ETDEWEB)

Wells, A.H.

1978-05-01

A computer simulation of the interactions of 50 MeV d/sup +/ on Be and 42 MeV p/sup +/ on Be neutron spectra with ICRU muscle tissue and Shonka A-150 tissue equivalent plastic was performed to allow computation of the charged particle spectra that result. Nuclear data were obtained from the Evaluated Nuclear Data File (ENDF) whenever possible and from the Intranuclear Cascade and Evaporation models otherwise. The dosimetry parameters calculated are: the kerma ratio, K/sub A-150//K/sub tissue/; the energy required to form an ion pair, W; and the stopping power ratio, S/sub g//sup W/.

Calculation of dosimetry parameters for fast neutron radiotherapy

International Nuclear Information System (INIS)

Wells, A.H.

1978-05-01

A computer simulation of the interactions of 50 MeV d + on Be and 42 MeV p + on Be neutron spectra with ICRU muscle tissue and Shonka A-150 tissue equivalent plastic was performed to allow computation of the charged particle spectra that result. Nuclear data were obtained from the Evaluated Nuclear Data File (ENDF) whenever possible and from the Intranuclear Cascade and Evaporation models otherwise. The dosimetry parameters calculated are: the kerma ratio, K/sub A-150//K/sub tissue/; the energy required to form an ion pair, W; and the stopping power ratio, S/sub g//sup W/

Nuclear data evaluation for medium and heavy nuclei

International Nuclear Information System (INIS)

Mizumoto, Motoharu

1988-01-01

Present status of nuclear data evaluation works for medium and heavy nuclei is described in this paper. These data are being prepared for JENDL-3 (Japanese Evaluated Nuclear Data Library-Version 3). At present, about a half of the data files, which are expected to be stored in the final library, has been brought into a temporary library called JENDL-3T. The remaining works and additional revisions are still needed to be made in order to finalize the data library as JENDL-3. Special emphases have been put on the high energy neutron data for which the previous JENDL-2 had some inadequacies, and gamma-ray production cross sections have been newly evaluated. Systematic and consistent evaluations have been intended for the new evaluations. (author)

Japanese Evaluated Nuclear Data Library, version-3

International Nuclear Information System (INIS)

Shibata, Keiichi; Nakagawa, Tsuneo; Asami, Tetsuo

1990-06-01

The general purpose file of the third version of Japanese Evaluated Nuclear Data Library, JENDL-3, has been compiled by the JAERI Nuclear Data Center in cooperation with the Japanese Nuclear Data Committee. It contains neutron nuclear data for 171 nuclides which are needed for design of fission and fusion reactors and for shielding calculation. In the JENDL-3 evaluation, much effort was devoted to improve reliability of high-energy data for fusion application and to include gamma-ray production data. Theoretical calculations played an important role in achieving these purposes. A special method called simultaneous evaluation was adopted to determine important cross sections of fissile and fertile nuclides. This report presents a general description for the evaluation of light, medium-heavy and heavy nuclide data. Also given are the descriptive data for each nuclide contained in the File 1 part of JENDL-3. (author)

Average cross sections calculated in various neutron fields

International Nuclear Information System (INIS)

Shibata, Keiichi

2002-01-01

Average cross sections have been calculated for the reactions contained in the dosimetry files, JENDL/D-99, IRDF-90V2, and RRDF-98 in order to select the best data for the new library IRDF-2002. The neutron spectra used in the calculations are as follows: 1) 252 Cf spontaneous fission spectrum (NBS evaluation), 2) 235 U thermal fission spectrum (NBS evaluation), 3) Intermediate-energy Standard Neutron Field (ISNF), 4) Coupled Fast Reactivity Measurement Facility (CFRMF), 5) Coupled thermal/fast uranium and boron carbide spherical assembly (ΣΣ), 6) Fast neutron source reactor (YAYOI), 7) Experimental fast reactor (JOYO), 8) Japan Material Testing Reactor (JMTR), 9) d-Li neutron spectrum with a 2-MeV deuteron beam. The items 3)-7) represent fast neutron spectra, while JMTR is a light water reactor. The Q-value for the d-Li reaction mentioned above is 15.02 MeV. Therefore, neutrons with energies up to 17 MeV can be produced in the d-Li reaction. The calculated average cross sections were compared with the measurements. Figures 1-9 show the ratios of the calculations to the experimental data which are given. It is found from these figures that the 58 Fe(n, γ) cross section in JENDL/D-99 reproduces the measurements in the thermal and fast reactor spectra better than that in IRDF-90V2. (author)

Application for internal dosimetry using biokinetic distribution of photons based on nuclear medicine images.

Science.gov (United States)

Leal Neto, Viriato; Vieira, José Wilson; Lima, Fernando Roberto de Andrade

2014-01-01

This article presents a way to obtain estimates of dose in patients submitted to radiotherapy with basis on the analysis of regions of interest on nuclear medicine images. A software called DoRadIo (Dosimetria das Radiações Ionizantes [Ionizing Radiation Dosimetry]) was developed to receive information about source organs and target organs, generating graphical and numerical results. The nuclear medicine images utilized in the present study were obtained from catalogs provided by medical physicists. The simulations were performed with computational exposure models consisting of voxel phantoms coupled with the Monte Carlo EGSnrc code. The software was developed with the Microsoft Visual Studio 2010 Service Pack and the project template Windows Presentation Foundation for C# programming language. With the mentioned tools, the authors obtained the file for optimization of Monte Carlo simulations using the EGSnrc; organization and compaction of dosimetry results with all radioactive sources; selection of regions of interest; evaluation of grayscale intensity in regions of interest; the file of weighted sources; and, finally, all the charts and numerical results. The user interface may be adapted for use in clinical nuclear medicine as a computer-aided tool to estimate the administered activity.

Validation of the Continuous-Energy Monte Carlo Criticality-Safety Analysis System MVP and JENDL-3.2 Using the Internationally Evaluated Criticality Benchmarks

International Nuclear Information System (INIS)

Mitake, Susumu

2003-01-01

Validation of the continuous-energy Monte Carlo criticality-safety analysis system, comprising the MVP code and neutron cross sections based on JENDL-3.2, was examined using benchmarks evaluated in the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments'. Eight experiments (116 configurations) for the plutonium solution and plutonium-uranium mixture systems performed at Valduc, Battelle Pacific Northwest Laboratories, and other facilities were selected and used in the studies. The averaged multiplication factors calculated with MVP and MCNP-4B using the same neutron cross-section libraries based on JENDL-3.2 were in good agreement. Based on methods provided in the Japanese nuclear criticality-safety handbook, the estimated criticality lower-limit multiplication factors to be used as a subcriticality criterion for the criticality-safety evaluation of nuclear facilities were obtained. The analysis proved the applicability of the MVP code to the criticality-safety analysis of nuclear fuel facilities, particularly to the analysis of systems fueled with plutonium and in homogeneous and thermal-energy conditions

Polymer gel dosimetry

Energy Technology Data Exchange (ETDEWEB)

Baldock, C [Institute of Medical Physics, School of Physics, University of Sydney (Australia); De Deene, Y [Radiotherapy and Nuclear Medicine, Ghent University Hospital (Belgium); Doran, S [CRUK Clinical Magnetic Resonance Research Group, Institute of Cancer Research, Surrey (United Kingdom); Ibbott, G [Radiation Physics, UT M D Anderson Cancer Center, Houston, TX (United States); Jirasek, A [Department of Physics and Astronomy, University of Victoria, Victoria, BC (Canada); Lepage, M [Centre d' imagerie moleculaire de Sherbrooke, Departement de medecine nucleaire et de radiobiologie, Universite de Sherbrooke, Sherbrooke, QC (Canada); McAuley, K B [Department of Chemical Engineering, Queen' s University, Kingston, ON (Canada); Oldham, M [Department of Radiation Oncology, Duke University Medical Center, Durham, NC (United States); Schreiner, L J [Cancer Centre of South Eastern Ontario, Kingston, ON (Canada)], E-mail: c.baldock@physics.usyd.edu.au, E-mail: yves.dedeene@ugent.be

2010-03-07

Polymer gel dosimeters are fabricated from radiation sensitive chemicals which, upon irradiation, polymerize as a function of the absorbed radiation dose. These gel dosimeters, with the capacity to uniquely record the radiation dose distribution in three-dimensions (3D), have specific advantages when compared to one-dimensional dosimeters, such as ion chambers, and two-dimensional dosimeters, such as film. These advantages are particularly significant in dosimetry situations where steep dose gradients exist such as in intensity-modulated radiation therapy (IMRT) and stereotactic radiosurgery. Polymer gel dosimeters also have specific advantages for brachytherapy dosimetry. Potential dosimetry applications include those for low-energy x-rays, high-linear energy transfer (LET) and proton therapy, radionuclide and boron capture neutron therapy dosimetries. These 3D dosimeters are radiologically soft-tissue equivalent with properties that may be modified depending on the application. The 3D radiation dose distribution in polymer gel dosimeters may be imaged using magnetic resonance imaging (MRI), optical-computerized tomography (optical-CT), x-ray CT or ultrasound. The fundamental science underpinning polymer gel dosimetry is reviewed along with the various evaluation techniques. Clinical dosimetry applications of polymer gel dosimetry are also presented. (topical review)

Internal sources dosimetry

International Nuclear Information System (INIS)

Savio, Eduardo

1994-01-01

The absorbed dose, need of estimation in risk evaluation in the application of radiopharmaceuticals in Nuclear Medicine practice,internal dosimetry,internal and external sources. Calculation methodology,Marinelli model,MIRD system for absorbed dose calculation based on biological parameters of radiopharmaceutical in human body or individual,energy of emitted radiations by administered radionuclide, fraction of emitted energy that is absorbed by target body.Limitation of the MIRD calculation model. A explanation of Marinelli method of dosimetry calculation,β dosimetry. Y dosimetry, effective dose, calculation in organs and tissues, examples. Bibliography .

Summary Report of Consultants' Meeting on Improvements and Extensions to IRDF (International Reactor Dosimetry File (IRDF-2002))

International Nuclear Information System (INIS)

Kellett, M.A.; Greenwood, L.R.

2010-12-01

The main aim of this Consultants' Meeting was to discuss the appropriate manner for implementing improvements and extensions to the current IRDF-2002 reactor dosimetry library. It was important to assess the applications requiring a dosimetry library, to discuss if a library that would meet the requirements of these varied applications could be produced and, if so, to define an approach for producing such an updated version. This report summarises the presentations and discussions undertaken in order to achieve these goals, followed by the recommendations and conclusions resulting from the meeting. (author)

US Army Radiological Bioassay and Dosimetry: The RBD software package

International Nuclear Information System (INIS)

Eckerman, K.F.; Ward, R.C.; Maddox, L.B.

1993-01-01

The RBD (Radiological Bioassay and Dosimetry) software package was developed for the U. S. Army Material Command, Arlington, Virginia, to demonstrate compliance with the radiation protection guidance 10 CFR Part 20 (ref. 1). Designed to be run interactively on an IBM-compatible personal computer, RBD consists of a data base module to manage bioassay data and a computational module that incorporates algorithms for estimating radionuclide intake from either acute or chronic exposures based on measurement of the worker's rate of excretion of the radionuclide or the retained activity in the body. In estimating the intake,RBD uses a separate file for each radionuclide containing parametric representations of the retention and excretion functions. These files also contain dose-per-unit-intake coefficients used to compute the committed dose equivalent. For a given nuclide, if measurements exist for more than one type of assay, an auxiliary module, REPORT, estimates the intake by applying weights assigned in the nuclide file for each assay. Bioassay data and computed results (estimates of intake and committed dose equivalent) are stored in separate data bases, and the bioassay measurements used to compute a given result can be identified. The REPORT module creates a file containing committed effective dose equivalent for each individual that can be combined with the individual's external exposure

Techniques for radiation measurements: Micro-dosimetry and dosimetry

International Nuclear Information System (INIS)

Waker, A. J.

2006-01-01

Experimental Micro-dosimetry is concerned with the determination of radiation quality and how this can be specified in terms of the distribution of energy deposition arising from the interaction of a radiation field with a particular target site. This paper discusses various techniques that have been developed to measure radiation energy deposition over the three orders of magnitude of site-size; nano-meter, micrometer and millimetre, which radiation biology suggests is required to fully account for radiation quality. Inevitably, much of the discussion will concern the use of tissue-equivalent proportional counters and variants of this device, but other technologies that have been studied, or are under development, for their potential in experimental Micro-dosimetry are also covered. Through an examination of some of the quantities used in radiation metrology and dosimetry the natural link with Micro-dosimetric techniques will be shown and the particular benefits of using Micro-dosimetric methods for dosimetry illustrated. (authors)

Thermoluminescence albedo-neutron dosimetry

International Nuclear Information System (INIS)

Strand, T.; Storruste, A.

1986-10-01

The report discusses neutron detection with respect to dosimetry and compares different thermoluminescent dosimetry materials for neutron dosimetry. Construction and calibration of a thermoluminescence albedo neutron dosemeter, developed by the authors, is described

Applications of gel dosimetry

International Nuclear Information System (INIS)

Ibbott, Geoffrey S

2004-01-01

Gel dosimetry has been examined as a clinical dosimeter since the 1950s. During the last two decades, however, a rapid increase in the number of investigators has been seen, and the body of knowledge regarding gel dosimetry has expanded considerably. Gel dosimetry is still considered a research project, and the introduction of this tool into clinical use is proceeding slowly. This paper will review the characteristics of gel dosimetry that make it desirable for clinical use, the postulated and demonstrated applications of gel dosimetry, and some complications, set-backs, and failures that have contributed to the slow introduction into routine clinical use

Benchmark experiment on vanadium assembly with D-T neutrons. Leakage neutron spectrum measurement

Energy Technology Data Exchange (ETDEWEB)

Kokooo; Murata, I.; Nakano, D.; Takahashi, A. [Osaka Univ., Suita (Japan); Maekawa, F.; Ikeda, Y.

1998-03-01

The fusion neutronics benchmark experiments have been done for vanadium and vanadium alloy by using the slab assembly and time-of-flight (TOF) method. The leakage neutron spectra were measured from 50 keV to 15 MeV and comparison were done with MCNP-4A calculations which was made by using evaluated nuclear data of JENDL-3.2, JENDL-Fusion File and FENDL/E-1.0. (author)

MO-B-BRB-04: 3D Dosimetry in End-To-End Dosimetry QA

Energy Technology Data Exchange (ETDEWEB)

Ibbott, G. [UT MD Anderson Cancer Center (United States)

2016-06-15

Full three-dimensional (3D) dosimetry using volumetric chemical dosimeters probed by 3D imaging systems has long been a promising technique for the radiation therapy clinic, since it provides a unique methodology for dose measurements in the volume irradiated using complex conformal delivery techniques such as IMRT and VMAT. To date true 3D dosimetry is still not widely practiced in the community; it has been confined to centres of specialized expertise especially for quality assurance or commissioning roles where other dosimetry techniques are difficult to implement. The potential for improved clinical applicability has been advanced considerably in the last decade by the development of improved 3D dosimeters (e.g., radiochromic plastics, radiochromic gel dosimeters and normoxic polymer gel systems) and by improved readout protocols using optical computed tomography or magnetic resonance imaging. In this session, established users of some current 3D chemical dosimeters will briefly review the current status of 3D dosimetry, describe several dosimeters and their appropriate imaging for dose readout, present workflow procedures required for good dosimetry, and analyze some limitations for applications in select settings. We will review the application of 3D dosimetry to various clinical situations describing how 3D approaches can complement other dose delivery validation approaches already available in the clinic. The applications presented will be selected to inform attendees of the unique features provided by full 3D techniques. Learning Objectives: L. John Schreiner: Background and Motivation Understand recent developments enabling clinically practical 3D dosimetry, Appreciate 3D dosimetry workflow and dosimetry procedures, and Observe select examples from the clinic. Sofie Ceberg: Application to dynamic radiotherapy Observe full dosimetry under dynamic radiotherapy during respiratory motion, and Understand how the measurement of high resolution dose data in an

Application for internal dosimetry using biokinetic distribution of photons based on nuclear medicine images*

Science.gov (United States)

Leal Neto, Viriato; Vieira, José Wilson; Lima, Fernando Roberto de Andrade

2014-01-01

Objective This article presents a way to obtain estimates of dose in patients submitted to radiotherapy with basis on the analysis of regions of interest on nuclear medicine images. Materials and Methods A software called DoRadIo (Dosimetria das Radiações Ionizantes [Ionizing Radiation Dosimetry]) was developed to receive information about source organs and target organs, generating graphical and numerical results. The nuclear medicine images utilized in the present study were obtained from catalogs provided by medical physicists. The simulations were performed with computational exposure models consisting of voxel phantoms coupled with the Monte Carlo EGSnrc code. The software was developed with the Microsoft Visual Studio 2010 Service Pack and the project template Windows Presentation Foundation for C# programming language. Results With the mentioned tools, the authors obtained the file for optimization of Monte Carlo simulations using the EGSnrc; organization and compaction of dosimetry results with all radioactive sources; selection of regions of interest; evaluation of grayscale intensity in regions of interest; the file of weighted sources; and, finally, all the charts and numerical results. Conclusion The user interface may be adapted for use in clinical nuclear medicine as a computer-aided tool to estimate the administered activity. PMID:25741101

Application for internal dosimetry using biokinetic distribution of photons based on nuclear medicine images

International Nuclear Information System (INIS)

Leal Neto, Viriato; Vieira, Jose Wilson; Lima, Fernando Roberto de Andrade

2014-01-01

Objective: this article presents a way to obtain estimates of dose in patients submitted to radiotherapy with basis on the analysis of regions of interest on nuclear medicine images. Materials and methods: a software called DoRadIo (Dosimetria das Radiacoes Ionizantes [Ionizing Radiation Dosimetry]) was developed to receive information about source organs and target organs, generating graphical and numerical results. The nuclear medicine images utilized in the present study were obtained from catalogs provided by medical physicists. The simulations were performed with computational exposure models consisting of voxel phantoms coupled with the Monte Carlo EGSnrc code. The software was developed with the Microsoft Visual Studio 2010 Service Pack and the project template Windows Presentation Foundation for C ⧣ programming language. Results: with the mentioned tools, the authors obtained the file for optimization of Monte Carlo simulations using the EGSnrc; organization and compaction of dosimetry results with all radioactive sources; selection of regions of interest; evaluation of grayscale intensity in regions of interest; the file of weighted sources; and, finally, all the charts and numerical results. Conclusion: the user interface may be adapted for use in clinical nuclear medicine as a computer-aided tool to estimate the administered activity. (author)

Application for internal dosimetry using biokinetic distribution of photons based on nuclear medicine images

Energy Technology Data Exchange (ETDEWEB)

Leal Neto, Viriato, E-mail: viriatoleal@yahoo.com.br [Instituto Federal de Educacao, Ciencia e Tecnologia de Pernambuco (IFPE), Recife, PE (Brazil); Vieira, Jose Wilson [Universidade Federal de Pernambuco (UPE), Recife, PE (Brazil); Lima, Fernando Roberto de Andrade [Centro Regional de Ciencias Nucleares (CRCN-NE/CNEN-PE), Recife, PE (Brazil)

2014-09-15

Objective: this article presents a way to obtain estimates of dose in patients submitted to radiotherapy with basis on the analysis of regions of interest on nuclear medicine images. Materials and methods: a software called DoRadIo (Dosimetria das Radiacoes Ionizantes [Ionizing Radiation Dosimetry]) was developed to receive information about source organs and target organs, generating graphical and numerical results. The nuclear medicine images utilized in the present study were obtained from catalogs provided by medical physicists. The simulations were performed with computational exposure models consisting of voxel phantoms coupled with the Monte Carlo EGSnrc code. The software was developed with the Microsoft Visual Studio 2010 Service Pack and the project template Windows Presentation Foundation for C ⧣ programming language. Results: with the mentioned tools, the authors obtained the file for optimization of Monte Carlo simulations using the EGSnrc; organization and compaction of dosimetry results with all radioactive sources; selection of regions of interest; evaluation of grayscale intensity in regions of interest; the file of weighted sources; and, finally, all the charts and numerical results. Conclusion: the user interface may be adapted for use in clinical nuclear medicine as a computer-aided tool to estimate the administered activity. (author)

Dosimetry Service

CERN Multimedia

2005-01-01

Please remember to read your dosimeter at least once a month. Regular read-outs are vital to ensure that your personal dose is periodically monitored. Dosimeters should be read even if you have not visited the controlled areas. Dosimetry Service - Tel. 72155 http://cern.ch/rp-dosimetry

Thermoluminescence in medical dosimetry

International Nuclear Information System (INIS)

Rivera, T.

2011-10-01

The dosimetry by thermoluminescence (Tl) is applied in the entire world for the dosimetry of ionizing radiations specially to personal and medical dosimetry. This dosimetry method has been very interesting for measures in vivo because the Tl dosimeters have the advantage of being very sensitive in a very small volume and they are also equivalent to tissue and they do not need additional accessories (for example, cable, electrometer, etc.) The main characteristics of the diverse Tl materials to be used in the radiation measures and practical applications are: the Tl curve, the share homogeneity, the signal stability after the irradiation, precision and exactitude, the response in function with the dose and the energy influence. In this work a brief summary of the advances of the radiations dosimetry is presented by means of the thermally stimulated luminescence and its application to the dosimetry in radiotherapy. (Author)

Neutron dosimetry - A review

Energy Technology Data Exchange (ETDEWEB)

Baum, J W

1955-03-29

This review summarizes information on the following subjects: (1) physical processes of importance in neutron dosimetry; (2) biological effects of neutrons; (3) neutron sources; and (4) instruments and methods used in neutron dosimetry. Also, possible improvements in dosimetry instrumentation are outlined and discussed. (author)

Dosimetry Service

CERN Multimedia

2005-01-01

Please remember to read your dosimeter at least once a month. Regular read-outs are vital to ensure that your personal dose is periodically monitored. Dosimeters should be read even if you have not visited the controlled areas. Dosimetry Service - Tel. 7 2155 http://cern.ch/rp-dosimetry

Dosimetry Service

CERN Multimedia

Dosimetry Service

2005-01-01

Please remember to read your dosimeter at least once a month. Regular read-outs are vital to ensure that your personal dose is periodically monitored. Dosimeters should be read even if you have not visited the controlled areas. Dosimetry Service Tel. 7 2155 http://cern.ch/rp-dosimetry

Dosimetry for radiation processing

DEFF Research Database (Denmark)

Miller, Arne

1986-01-01

During the past few years significant advances have taken place in the different areas of dosimetry for radiation processing, mainly stimulated by the increased interest in radiation for food preservation, plastic processing and sterilization of medical products. Reference services both...... and sterilization dosimetry, optichromic dosimeters in the shape of small tubes for food processing, and ESR spectroscopy of alanine for reference dosimetry. In this paper the special features of radiation processing dosimetry are discussed, several commonly used dosimeters are reviewed, and factors leading...

Alanine dosimetry for clinical applications. Proceedings

International Nuclear Information System (INIS)

Anton, M.

2006-05-01

The following topics are dealt with: Therapy level alanine dosimetry at the UK Nationational Physical Laboratory, alanine as a precision validation tool for reference dosimetry, composition of alanine pellet dosimeters, the angular dependence of the alanine ESR spectrum, the CIAE alanine dosimeter for radiotherapy level, a correction for temporal evolution effects in alanine dosimetry, next-generation services foe e-traceability to ionization radiation national standards, establishing e-traceability to HIST high-dose measurement standards, alanine dosimetry of dose delivery from clinical accelerators, the e-scan alanine dosimeter reader, alanine dosimetry at ISS, verification of the integral delivered dose for IMRT treatment in the head and neck region with ESR/alanine dosimetry, alanine dosimetry in helical tomotherapy beams, ESR dosimetry research and development at the University of Palermo, lithium formate as a low-dose EPR radiation dosimeter, sensitivity enhancement of alanine/EPR dosimetry. (HSI)

Developments in physical dosimetry and radiation protection; Entwicklungen in der physikalischen Dosimetrie im Strahlenschutz

Energy Technology Data Exchange (ETDEWEB)

Fiebich, Martin [Technische Hochschule Mittelhessen, Giessen (Germany). Inst. fuer Medizinische Physik und Strahlenschutz

2017-07-01

In the frame of physical dosimetry new dose units have been defined: the depth personal dose (equivalent dose in 10 mm depth) and the surface personal dose (equivalent dose in 0.07 mm depth). Physical dosimetry is applied for the determination of occupational radiation exposure, the radiation protected area control, the estimation of radiation exposure of patients during radiotherapy, for quality assurance and in research projects and optimization challenges. Developments have appeared with respect to punctual measuring chambers, eye lens dosimetry, OSL (optically stimulated luminescence) dosimetry, real-time dosimetry and Monte Carlo methods. New detection limits of about 1 micro Gy were reached.

Textbook of dosimetry. 4. ed.

International Nuclear Information System (INIS)

Ivanov, V.I.

1999-01-01

This textbook of dosimetry is devoted to the students in physics and technical physics of high education institutions, confronted with different application of atomic energy as well as with protection of population and environment against ionizing radiations. Atomic energy is highly beneficial for man but unfortunately incorporates potential dangers which manifest in accidents, the source of which is either insufficient training of the personnel, a criminal negligence or insufficient reliability of the nuclear facilities. The majority of the incident and accident events have had as origin the personnel errors. This was the case with both the 'Three Miles Island' (1979) and Chernobyl (1986) NPP accidents. The dosimetry science acquires a vital significance in accident situations since the data obtained by its procedures are essential in choosing the correct immediate actions, behaviour tactics, orientation of liquidation of accident consequences as well as in ensuring the health of population. An important accent is placed in this manual on clarification of the nature of physical processes taken place in dosimetric detectors, in establishing the relation between radiation field characteristics and the detector response as well as in defining different dosimetric quantities. The terminology and the units of physical quantities is based on the international system of units. The book contains the following 15 chapters: 1. Ionizing radiation field; 2. Radiation doses; 3. Physical bases of gamma radiation dosimetry; 4. Ionization dosimetric detectors; 5. Semiconductor dosimetric detectors; 6. Scintillation detection in the gamma radiation dosimetry; 7. Luminescent methods in dosimetry; 8. The photographic and chemical methods of gamma radiation dosimetry; 9. Neutron dosimetry; 10. Dosimetry of high intensity radiation; 11. Dosimetry of high energy Bremsstrahlung; 12. Measurement of the linear energy transfer; 13. Microdosimetry; 14. Dosimetry of incorporated

New remarks on KERMA factors and DPA cross section data in ACE files

International Nuclear Information System (INIS)

Konno, Chikara; Sato, Satoshi; Ohta, Masayuki; Kwon, Saerom; Ochiai, Kentaro

2016-01-01

KERMA factors and DPA cross section data are essential for nuclear heating and material damage estimation in fusion reactor designs. Recently we compared KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 and it was found out that the KERMA factors and DPA cross section data of a lot of nuclei did not always agree among the nuclear data libraries. We investigated the nuclear data libraries and the nuclear data processing code NJOY and specified new reasons for the discrepancies; (1) incorrect nuclear data and NJOY bugs, (2) huge helium production cross section data, (3) gamma production data format in the nuclear data, (4) no detailed secondary particle data (energy–angular distribution data). These problems should be resolved based on this study.

New remarks on KERMA factors and DPA cross section data in ACE files

Energy Technology Data Exchange (ETDEWEB)

Konno, Chikara, E-mail: konno.chikara@jaea.go.jp; Sato, Satoshi; Ohta, Masayuki; Kwon, Saerom; Ochiai, Kentaro

2016-11-01

KERMA factors and DPA cross section data are essential for nuclear heating and material damage estimation in fusion reactor designs. Recently we compared KERMA factors and DPA cross section data in the latest official ACE files of JENDL-4.0, ENDF/B-VII.1, JEFF-3.2 and FENDL-3.0 and it was found out that the KERMA factors and DPA cross section data of a lot of nuclei did not always agree among the nuclear data libraries. We investigated the nuclear data libraries and the nuclear data processing code NJOY and specified new reasons for the discrepancies; (1) incorrect nuclear data and NJOY bugs, (2) huge helium production cross section data, (3) gamma production data format in the nuclear data, (4) no detailed secondary particle data (energy–angular distribution data). These problems should be resolved based on this study.

Evaluation of cross sections of 56Fe up to 3 GeV and integral benchmark calculation for thick target yield

International Nuclear Information System (INIS)

Yoshizawa, Nobuaki; Meigo, Shin-ichiro

2001-01-01

The neutron and proton cross sections of 56 Fe were evaluated up to 3 GeV. JENDL High Energy File of 56 Fe were developed for use in transport calculation. For neutrons, the high-energy data are merged with JENDL3.3-file. Integral benchmark calculations for thick target neutron yields (TTY) for 113 MeV and 256 MeV proton bombardment of Fe targets were performed using the evaluated libraries. Calculated TTY neutron spectra were compared with experimental data. For 113 MeV, calculated TTY at 7.5 degree underestimated in the emitted neutron energy range above 10 MeV. For 256 MeV, calculated TTY well agree with experimental data except below 10 MeV. (author)

Personnel neutron dosimetry

International Nuclear Information System (INIS)

Hankins, D.

1982-04-01

This edited transcript of a presentation on personnel neutron discusses the accuracy of present dosimetry practices, requirements, calibration, dosemeter types, quality factors, operational problems, and dosimetry for a criticality accident. 32 figs

Skin dosimetry - radiological protection aspects of skin dosimetry

International Nuclear Information System (INIS)

Dennis, J.A.

1991-01-01

Following a Workshop in Skin Dosimetry, a summary of the radiological protection aspects is given. Aspects discussed include routine skin monitoring and dose limits, the need for careful skin dosimetry in high accidental exposures, techniques for assessing skin dose at all relevant depths and the specification of dose quantities to be measured by personal dosemeters and the appropriate methods to be used in their calibration. (UK)

BASACF, Integral Neutron Spectra Adjustment and Dosimetry

International Nuclear Information System (INIS)

Tichy, Milos

1996-01-01

1 - Description of program or function: Adjustment of a neutron spectrum based on integral detector measurements and calculation of an integral dosimetric quantity (integral flux, d.p.a., dose equivalent) and its variance. The program requires measured data (activities and their covariance matrix) and a priori information (spectrum, dosimetry cross sections, integral quantity conversion factor and their covariance matrices). All a priori covariance matrices can be read in from a file prepared by some other code or can be generated by means of three different methods (by subroutines included in the program). A subroutine which can normalize the a priori flux to measured data is also included. The program provides also adjusted dosimetry cross sections (with covariance matrix) so that it can be used for an adjustment of cross sections (or response functions of e.g. Bonner balls) by measurements in well-known neutron spectra. 2 - Method of solution: Bayesian theorem on conditional probability applied to linearized relation between activities, dosimetry cross sections and flux. All probability distributions are supposed to be normal and this supposition leads to minimizing of the same functional as least squares method (STAY'SL). This task is solved by a covariance filter method which avoids any matrix inversion and is numerically robust and stable. 3 - Restrictions on the complexity of the problem: This version can use 45 energy groups and 5 detectors and occupies 310 kB of main memory. This restriction can be modified according to available memory. The covariance matrix of activities is supposed diagonal. A solution is produced for any set of input data but in the case of non-consistent data, when measured activities do not match the a priori flux, the solution is not very meaningful

ESR Dosimetry

International Nuclear Information System (INIS)

Baffa, Oswaldo; Rossi, Bruno; Graeff, Carlos; Kinoshita, Angela; Chen Abrego, Felipe; Santos, Adevailton Bernardo dos

2004-01-01

ESR dosimetry is widely used for several applications such as dose assessment in accidents, medical applications and sterilization of food and other materials. In this work the dosimetric properties of natural and synthetic Hydroxyapatite, Alanine, and 2-Methylalanine are presented. Recent results on the use of a K-Band (24 GHz) ESR spectrometer in dosimetry are also presented

Benchmark experiment on vanadium assembly with D-T neutrons. In-situ measurement

Energy Technology Data Exchange (ETDEWEB)

Maekawa, Fujio; Kasugai, Yoshimi; Konno, Chikara; Wada, Masayuki; Oyama, Yukio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Murata, Isao; Kokooo; Takahashi, Akito

1998-03-01

Fusion neutronics benchmark experimental data on vanadium were obtained for neutrons in almost entire energies as well as secondary gamma-rays. Benchmark calculations for the experiment were performed to investigate validity of recent nuclear data files, i.e., JENDL Fusion File, FENDL/E-1.0 and EFF-3. (author)

Monte Carlo analysis of KRITZ-2 critical benchmarks on the reactivity temperature coefficient using ENDF/B-VII.1 and JENDL-4.0 nuclear data libraries

International Nuclear Information System (INIS)

El Ouahdani, S.; Boukhal, H.; Erradi, L.; Chakir, E.; El Bardouni, T.; Hajjaji, O.; Boulaich, Y.; Benaalilou, K.; Kaddour, M.

2016-01-01

Highlights: • A set of KRITZ-2 experiments with UO 2 and MOX LWR lattices, at room and elevated temperatures, have been analysed using the MCNP6.1 code with the libraries: JENDL-4 and ENDF/B-VII.1. • The detailed comparisons of the calculations and measurements demonstrate a good agreement between calculations and measurements. • To investigate better the influence of cross sections differences on the reactivity temperature coefficient, we break it down into its components using a pin cell model. - Abstract: A set of KRITZ-2 experiments light water moderated lattices with uranium oxide and mixed-oxide fuel rods, at room and elevated temperatures, performed in the early 1970’s have been assessed. Using the MCNP6.1 code with the most recent cross section libraries: JENDL-4 and ENDF/B-VII.1, the critical experiments KRITZ: 2-1, KRITZ: 2-13, and KRITZ: 2-19 achieved in the Sweden reactor KRITZ were analyzed. We have used the ENDF/B-VII.1 data provided with the MCNP6.1.1 version in ACE format and the Makxsf utility to handle the data in the specific temperatures not available in the MCNP6.1.1 original data. The JENDL-4 evaluations were processed using NJOY99 (update 364) to the temperatures of interest. The detailed comparisons of the calculated and measured (Benchmark, 2005) effective multiplication factors and pin power distributions for UO2 and MOX fuelled cores presented in this work demonstrate a good agreement between calculation and measurements. The maximum deviation of the calculation from the experimental data for k eff , is 0.58% (absolute value) obtained for the KRITZ 2:1 at 248.5 °C using ENDF/B-VII.1 data. To investigate better the influence of cross sections differences on the reactivity and temperature coefficient, we break down the infinite multiplication factor into its components using a pin cell model. Using this simple model we evaluated the temperature effect on the infinite multiplication factor and the effect on its components. We have

Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

International Nuclear Information System (INIS)

Mori, Takamasa; Nakagawa, Masayuki; Kaneko, Kunio.

1996-05-01

A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author)

Proceedings of the 2000 symposium on nuclear data

Energy Technology Data Exchange (ETDEWEB)

Yamano, Naoki [Sumitomo Atomic Energy Industries, Ltd., Tokyo (Japan); Fukahori, Tokio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-03-01

The 2000 Symposium on Nuclear Data was held at Tokai Research Establishment, Japan Atomic Energy Research Institute (JAERI), on 16th and 17th of November 2000. Japanese Nuclear Data Committee and Nuclear Data Center, JAERI organized this symposium. In the oral sessions, presented were 18 papers on topics of recent experiments, status of JENDL-3.3 and JENDL High Energy File, benchmark tests of JENDL-3.2, international activities and other subjects. In the poster session, presented were 40 papers concerning experiments, evaluations, benchmark tests and on-line database on nuclear data. Those presented papers are compiled in the proceedings. The 58 of the presented papers are indexed individually. (J.P.N.)

Neutron cross section library production code system for continuous energy Monte Carlo code MVP. LICEM

Energy Technology Data Exchange (ETDEWEB)

Mori, Takamasa; Nakagawa, Masayuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kaneko, Kunio

1996-05-01

A code system has been developed to produce neutron cross section libraries for the MVP continuous energy Monte Carlo code from an evaluated nuclear data library in the ENDF format. The code system consists of 9 computer codes, and can process nuclear data in the latest ENDF-6 format. By using the present system, MVP neutron cross section libraries for important nuclides in reactor core analyses, shielding and fusion neutronics calculations have been prepared from JENDL-3.1, JENDL-3.2, JENDL-FUSION file and ENDF/B-VI data bases. This report describes the format of MVP neutron cross section library, the details of each code in the code system and how to use them. (author).

Dosimetry of ionizing radiation

International Nuclear Information System (INIS)

Musilek, L.; Seda, J.; Trousil, J.

1992-01-01

The publication deals with a major field of ionizing radiation dosimetry, viz., integrating dosimetric methods, which are the basic means of operative dose determination. It is divided into the following sections: physical and chemical effects of ionizing radiation; integrating dosimetric methods for low radiation doses (film dosimetry, nuclear emulsions, thermoluminescence, radiophotoluminescence, solid-state track detectors, integrating ionization dosemeters); dosimetry of high ionizing radiation doses (chemical dosimetric methods, dosemeters based on the coloring effect, activation detectors); additional methods applicable to integrating dosimetry (exoelectron emission, electron spin resonance, lyoluminescence, etc.); and calibration techniques for dosimetric instrumentation. (Z.S.). 422 refs

Comparative studies of JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B-6.8 data libraries on the Monte Carlo continuous energy modeling of the gas turbine-modular helium reactor operating with thorium fuels

International Nuclear Information System (INIS)

Talamo, Alberto; Gudowski, Waclaw

2005-01-01

One of the major benefits of the Gas Turbine-Modular Helium Reactor is the capability to operate with several different types of fuel; either Light Water Reactors waste, military plutonium or thorium represent valid candidates as possible types of fuel. In the present studies, we performed a comparison of various nuclear data libraries by the Monte Carlo Continuous Energy Burnup Code MCB applied to the Gas Turbine-Modular Helium Reactor operating on a thorium fuel. A thorium fuel offers valuable attractive advantages: low fuel cost, high reduction of actinides production and the possibility to enable the reactor to act as a breeder of fuel by the neutron capture of fertile 232 Th. We evaluated the possibility to mix thorium with small quantities, about 3% in atomic composition, of 239 Pu, 233 U and 235 U. The mass of thorium must be much larger than that one of plutonium or uranium because of the low capture cross section of thorium compared to the fission one of the fissile nuclides; at the same time, the quantity of the fissile isotopes must grant the criticality condition. These two simultaneous constraints force to load a huge mass of fuel in the reactor; consequently, we propose to allocate the fuel in TRISO particles with a large radius of the kernel. For each of the three different fuels we calculated the evolution of the fuel composition by the MCB code equipped with five different nuclear data libraries: JENDL-3.3, JENDL-3.2, JEFF-3, JEF-2.2 and ENDF/B. (author)

Clinical dosimetry

International Nuclear Information System (INIS)

Rassow, J.

1973-01-01

The main point of this paper on clinical dosimetry which is to be understood here as application of physical dosimetry on accelerators in medical practice, is based on dosimetric methodics. Following an explanation of the dose parameters and description of the dose distribution important for clinical practice as well as geometric irradiation parameters, the significance of a series of physical parameters such as accelerator energy, surface energy of average stopping power etc. is dealt with in detail. Following a section on field homogenization with bremsstrahlung and electron radiation, details on dosimetry in clinical practice are given. Finally, a few problems of dosemeter or monitor calibration on accelerators are described. The explanations are supplemented by a series of diagrams and tables. (ORU/LH) [de

Fast neutron spectrometry and dosimetry; Spectrometrie et dosimetrie des neutrons rapides

Energy Technology Data Exchange (ETDEWEB)

Blaize, S; Ailloud, J; Mariani, J; Millot, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

We have studied fast neutron spectrometry and dosimetry through the recoil protons they produce in hydrogenated samples. In spectrometric, we used nuclear emulsions, in dosimetric, we used polyethylene coated with zinc sulphide and placed before a photomultiplier. (author)Fren. [French] Nous avons etudie la spectrometrie et la dosimetrie des neutrons rapides en utilisant les protons de recul qu'ils produisent dans une matiere hydrogenee. En spectrometrie, nous avons employe des emulsions nucleaires, en dosimetrie, du polyethylene recouvert de sulfure de zinc place devant un photomultiplicateur. (auteur)

Dosimetry Control: Technic and methods. Proceedings of the international workshop 'Actual problems of dosimetry'

International Nuclear Information System (INIS)

Lyutsko, A.M.; Nesterenko, V.B.; Chudakov, V.A.; Konoplya, E.F.; Milyutin, A.A.

1997-10-01

There is a number of unsolved problems of both dosimetric and radiometric control, questions of the biological dosimetry, reconstruction of dozes of irradiation of the population at radiation incidents, which require coordination of efforts of scientists in various areas of a science. The submitted materials are grouped on five units: dosimetry engineering, biological dosimetry and markers of radiation impact, dosimetry of a medical irradiation, normative and measurement assurance of the dosimetric control, monitoring and reconstruction of dozes at radiation incidents

Non-conventional personal dosimetry techniques

International Nuclear Information System (INIS)

Regulla, D.F.

1984-01-01

Established dosimetry has achieved a high standard in personnel monitoring. This applies particularly to photon dosimetry. Nevertheless, even in photon dosimetry, improvements and changes are being made. The reason may be technological progress, or the introduction of new tasks on the basis of the recommendations of international bodies (e.g. the new ICRU measurement unit) of national legislation. Since we are restricting ourselves here to technical trends the author would like to draw attention to various activities of current interest, e.g. the computation of receptor-related conversion coefficients from personal dose to organ or body doses, taking into account the conditions of exposure with respect to differential energy and angular distribution of the radiation field. Realistic data on exposure geometry are taken from work place analyses. Furthermore, the data banks of central personal dosimetry services are subject to statistical evaluation and radiation protection trend analysis. Technological progress and developments are considered from the point of view of personal dosimetry, partial body or extremity dosimetry and accidental dosimetry

Dosimetry and Calibration Section

International Nuclear Information System (INIS)

Otto, T.

1999-01-01

The Dosimetry and Calibration Section fulfils two tasks within CERN's Radiation Protection Group: the Individual Dosimetry Service monitors more than 5000 persons potentially exposed to ionizing radiation on the CERN sites, and the Calibration Laboratory verifies throughout the year, at regular intervals, over 1000 instruments, monitors, and electronic dosimeters used by RP Group. The establishment of a Quality Assurance System for the Individual Dosimetry Service, a requirement of the new Swiss Ordinance for personal dosimetry, put a considerable workload on the section. Together with an external consultant it was decided to identify and then describe the different 'processes' of the routine work performed in the dosimetry service. The resulting Quality Manual was submitted to the Federal Office for Public Health in Bern in autumn. The CERN Individual Dosimetry Service will eventually be officially endorsed after a successful technical test in March 1999. On the technical side, the introduction of an automatic development machine for gamma films was very successful. It processes the dosimetric films without an operator being present, and its built-in regeneration mechanism keeps the concentration of the processing chemicals at a constant level

Dosimetry for radiation processing

International Nuclear Information System (INIS)

McLaughlin, W.L.; Boyd, A.W.; Chadwick, K.H.; McDonald, J.C.; Miller, A.

1989-01-01

Radiation processing is a relatively young industry with broad applications and considerable commercial success. Dosimetry provides an independent and effective way of developing and controlling many industrial processes. In the sterilization of medical devices and in food irradiation, where the radiation treatment impacts directly on public health, the measurements of dose provide the official means of regulating and approving its use. In this respect, dosimetry provides the operator with a means of characterizing the facility, of proving that products are treated within acceptable dose limits and of controlling the routine operation. This book presents an up-to-date review of the theory, data and measurement techniques for radiation processing dosimetry in a practical and useful way. It is hoped that this book will lead to improved measurement procedures, more accurate and precise dosimetry and a greater appreciation of the necessity of dosimetry for radiation processing. (author)

Thin film tritium dosimetry

Science.gov (United States)

Moran, Paul R.

1976-01-01

The present invention provides a method for tritium dosimetry. A dosimeter comprising a thin film of a material having relatively sensitive RITAC-RITAP dosimetry properties is exposed to radiation from tritium, and after the dosimeter has been removed from the source of the radiation, the low energy electron dose deposited in the thin film is determined by radiation-induced, thermally-activated polarization dosimetry techniques.

DOSCROS81. ECN Cross-Section Library for neutron dosimetry. Summary of contents and documentation

International Nuclear Information System (INIS)

Lemmel, H.D.

1982-01-01

This document summarizes the contents and documentation of the Cross Section Library DOSCROS81 (640 groups in an extended SAND-II format). The library is based on ENDF/B-5 dosimetry file, supplemented with some other evaluations. The total number of reaction cross section sets incorporated in this library is 70 (+3 cover cross section sets). The entire library can be obtained free of charge from the IAEA Nuclear Data Section. A revised version is called DOSCROS81A. (author)

Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

Energy Technology Data Exchange (ETDEWEB)

Nakagawa, Tsuneo [ed.

1992-06-15

Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10{sup {minus}5} eV to 20 MeV. Almost of the cross section data reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in order tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum.

Curves and tables of neutron cross sections of fission product nuclei in JENDL-3

International Nuclear Information System (INIS)

Nakagawa, Tsuneo

1992-06-01

Neutron cross sections of 172 nuclei in the fission product region stored in JENDL-3 are shown in graphs and tables. The evaluation work of these nuclei was made by the Fission Product Nuclear Data Working Group of the Japanese Nuclear Data Committee, in the neutron energy region from 10 -5 eV to 20 MeV. Almost all the cross section data are reproduced in graphs in this report. The cross section averaged over 38 energy intervals are listed in a table. Shown in other tables are thermal cross sections, resonance integrals, Maxwellian neutron flux average cross sections, fission spectrum average cross sections, 14-MeV cross sections, one group average cross sections in neutron flux of typical types of fission reactors and average cross sections in the 30-keV Maxwellian spectrum. (author)

NEA Data Bank report to the NRDC, Vienna, 28-30 May 2001. P12

International Nuclear Information System (INIS)

2001-01-01

Request List. This involves loading the List into an ORACLE database so that online searches can be performed, but more importantly specific users can add comments to the list concerning the quoted requests. JANIS (the Java Nuclear Information System) is a computer program written to display different types of nuclear data (experimental, evaluated and processed data). The Data Bank answered one hundred and eighty (180) manual data requests in 2000. Most of these were either for copies of the CD-ROM of the JENDL-3.2 general-purpose file (90 copies) or of the JENDL/D-99 dosimetry file (70 copies), made available to the Data Bank by the JAERI Nuclear Data Centre

Fundamentals of x-ray dosimetry

International Nuclear Information System (INIS)

Roesch, W.C.

1976-01-01

Fundamental information about x-ray dosimetry is presented. Definitions are given and expanded on for dose, absorbed dose including microdosimetry, radiation physics (properties of the radiation that are important to dosimetry), and dosimetry (how the properties are dealt with in determining dose). 5 figs, 12 refs

Thermoluminescent dosimetry in veterinary diagnostic radiology

International Nuclear Information System (INIS)

Hernández-Ruiz, L.; Jimenez-Flores, Y.; Rivera-Montalvo, T.; Arias-Cisneros, L.; Méndez-Aguilar, R.E.; Uribe-Izquierdo, P.

2012-01-01

This paper presents the results of Environmental and Personnel Dosimetry made in a radiology area of a veterinary hospital. Dosimetry was realized using thermoluminescent (TL) materials. Environmental Dosimetry results show that areas closer to the X-ray equipment are safe. Personnel Dosimetry shows important measurements of daily workday in some persons near to the limit established by ICRP. TL results of radiation measurement suggest TLDs are good candidates as a dosimeter to radiation dosimetry in veterinary radiology. - Highlights: ► Personnel dosimetry in laboratory veterinary diagnostic was determined. ► Student workplaces are safe against radiation. ► Efficiency value of apron lead was determined. ► X-ray beams distribution into veterinarian laboratory was measured.

Sixth symposium on neutron dosimetry

International Nuclear Information System (INIS)

1987-01-01

This booklet contains all abstracts of papers presented in 13 sessions. Main topics: Cross sections and Kerma factors; analytical radiobiology; detectors for personnel monitoring; secondary charged particles and microdosimetric basis of q-value for neutrons; personnel dosimetry; concepts for radiation protection; ambient monitoring; TEPC and ion chambers in radiation protection; beam dosimetry; track detectors (CR-39); dosimetry at biomedical irradiation facilities; health physics at therapy facilities; calibration for radiation protection; devices for beam dosimetry (TLD and miscellaneous); therapy and biomedical irradiation facilities; treatment planning. (HP)

Environmental dosimetry

International Nuclear Information System (INIS)

Gold, R.

1977-01-01

For more than 60 years, natural radiation has offered broad opportunities for basic research as evidenced by many fundamental discoveries. Within the last decade, however, dramatic changes have occurred in the motivation and direction of this research. The urgent need for economical energy sources entailing acceptably low levels of environmental impact has compelled the applied aspects of our radiation environment to become overriding considerations. It is within this general framework that state-of-the-art environmental dosimetry techniques are reviewed. Although applied motivation and relevance underscores the current milieu for both reactor and environmental dosimetry, a perhaps even more unifying force is the broad similarity of reactor and environmental radiation fields. In this review, a comparison of these two mixed radiation fields is presented stressing the underlying similarities that exist. On this basis, the evolution of a strong inner bond between dosimetry methods for both reactor and environmental radiation fields is described. The existence of this bond will be illustrated using representative examples of observed spectra. Dosimetry methods of particularly high applicability for both of these fields are described. Special emphasis is placed on techniques of high sensitivity and absolute accuracy which are capable of resolving the components of these mixed radiation fields

Nuclear accident dosimetry

International Nuclear Information System (INIS)

1982-01-01

The film presents statistical data on criticality accidents. It outlines past IAEA activities on criticality accident dosimetry and the technical documents that resulted from this work. The film furthermore illustrates an international comparison study on nuclear accident dosimetry conducted at the Atomic Energy Research Establishment, Harwell, United Kingdom

Nuclear accident dosimetry

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-12-31

The film presents statistical data on criticality accidents. It outlines past IAEA activities on criticality accident dosimetry and the technical documents that resulted from this work. The film furthermore illustrates an international comparison study on nuclear accident dosimetry conducted at the Atomic Energy Research Establishment, Harwell, United Kingdom

Performance testing of dosimetry processors, status of NRC rulemaking for improved personnel dosimetry processing, and some beta dosimetry and instrumentation problems observed by NRC regional inspectors

International Nuclear Information System (INIS)

Dennis, N.A.; Kinneman, J.D.; Costello, F.M.; White, J.R.; Nimitz, R.L.

1983-01-01

Early dosimetry processor performance studies conducted between 1967 and 1979 by several different investigators indicated that a significant percentage of personnel dosimetry processors may not be performing with a reasonable degree of accuracy. Results of voluntary performance testing of US personnel dosimetry processors against the final Health Physics Society Standard, Criteria for Testing Personnel Dosimetry Performance by the University of Michigan for the Nuclear Regulatory Commission (NRC) will be summarized with emphasis on processor performance in radiation categories involving beta particles and beta particles and photon mixtures. The current status of the NRC's regulatory program for improved personnel dosimetry processing will be reviewed. The NRC is proposing amendments to its regulations, 10 CFR Part 20, that would require its licensees to utilize specified personnel dosimetry services from processors accredited by the National Voluntary Laboratory Accreditation Program of the National Bureau of Standards. Details of the development and schedule for implementation of the program will be highlighted. Finally, selected beta dosimetry and beta instrumentation problems observed by NRC Regional Staff during inspections of NRC licensed facilities will be discussed

Quantitative imaging for clinical dosimetry

Energy Technology Data Exchange (ETDEWEB)

Bardies, Manuel [INSERM U601, 9 Quai Moncousu, 44093 Nantes (France)]. E-mail: manu@nantes.inserm.fr; Flux, Glenn [Department of Physics, Royal Marsden NHS Trust, Sutton (United Kingdom); Lassmann, Michael [Department of Nuclear Medicine, Julis-Maximilians University, Wuerzburg (Germany); Monsieurs, Myriam [Department of Health Physics, University of Ghent, 9000 Ghent (Belgium); Savolainen, Sauli [Department of Physical Sciences, University of Helsinki and HUS, Helsinki Medical Imaging Center, Helsinki University Central Hospital (Finland); Strand, Sven-Erik [Medical Radiation Physics, Department of Clinical Sciences Lund, Lund University (Sweden)

2006-12-20

Patient-specific dosimetry in nuclear medicine is now a legal requirement in many countries throughout the EU for targeted radionuclide therapy (TRT) applications. In order to achieve that goal, an increased level of accuracy in dosimetry procedures is needed. Current research in nuclear medicine dosimetry should not only aim at developing new methods to assess the delivered radiation absorbed dose at the patient level, but also to ensure that the proposed methods can be put into practice in a sufficient number of institutions. A unified dosimetry methodology is required for making clinical outcome comparisons possible.

Comparisons of the MCNP criticality benchmark suite with ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0

International Nuclear Information System (INIS)

Kim, Do Heon; Gil, Choong-Sup; Kim, Jung-Do; Chang, Jonghwa

2003-01-01

A comparative study has been performed with the latest evaluated nuclear data libraries ENDF/B-VI.8, JENDL-3.3, and JEFF-3.0. The study has been conducted through the benchmark calculations for 91 criticality problems with the libraries processed for MCNP4C. The calculation results have been compared with those of the ENDF60 library. The self-shielding effects of the unresolved-resonance (UR) probability tables have also been estimated for each library. The χ 2 differences between the MCNP results and experimental data were calculated for the libraries. (author)

Bibliographical database of radiation biological dosimetry and risk assessment: Part 2

International Nuclear Information System (INIS)

Straume, T.; Ricker, Y.; Thut, M.

1990-09-01

This is part 11 of a database constructed to support research in radiation biological dosimetry and risk assessment. Relevant publications were identified through detailed searches of national and international electronic databases and through our personal knowledge of the subject. Publications were numbered and key worded, and referenced in an electronic data-retrieval system that permits quick access through computerized searches on authors, key words, title, year, journal name, or publication number. Photocopies of the publications contained in the database are maintained in a file that is numerically arranged by our publication acquisition numbers. This volume contains 1048 additional entries, which are listed in alphabetical order by author. The computer software used for the database is a simple but sophisticated relational database program that permits quick information access, high flexibility, and the creation of customized reports. This program is inexpensive and is commercially available for the Macintosh and the IBM PC. Although the database entries were made using a Macintosh computer, we have the capability to convert the files into the IBM PC version. As of this date, the database cites 2260 publications. Citations in the database are from 200 different scientific journals. There are also references to 80 books and published symposia, and 158 reports. Information relevant to radiation biological dosimetry and risk assessment is widely distributed within the scientific literature, although a few journals clearly predominate. The journals publishing the largest number of relevant papers are Health Physics, with a total of 242 citations in the database, and Mutation Research, with 185 citations. Other journals with over 100 citations in the database, are Radiation Research, with 136, and International Journal of Radiation Biology, with 132

Alanine EPR dosimetry of therapeutic irradiators

International Nuclear Information System (INIS)

Bugay, O.; Bartchuk, V.; Kolesnik, S.; Mazin, M.; Gaponenko, H.

1999-01-01

The high-dose alanine EPR dosimetry is a very precise method in the dose range 1-100 kGy. The system is used generally as the standard high-dose transfer dosimetry in many laboratories. This is comparatively expensive technique so it is important to use it as a more universal dosimetry system also in the middle and low dose ranges. The problems of the middle-dose alanine dosimetry are discussed and the solution of several problems is proposed. The alanine EPR dosimetry has been applied to the dose measurements of medical irradiators in the Kiev City Oncology Center. (author)

Comparison of Real-Time Intraoperative Ultrasound-Based Dosimetry With Postoperative Computed Tomography-Based Dosimetry for Prostate Brachytherapy

International Nuclear Information System (INIS)

Nag, Subir; Shi Peipei; Liu Bingren; Gupta, Nilendu; Bahnson, Robert R.; Wang, Jian Z.

2008-01-01

Purpose: To evaluate whether real-time intraoperative ultrasound (US)-based dosimetry can replace conventional postoperative computed tomography (CT)-based dosimetry in prostate brachytherapy. Methods and Materials: Between December 2001 and November 2002, 82 patients underwent 103 Pd prostate brachytherapy. An interplant treatment planning system was used for real-time intraoperative transrectal US-guided treatment planning. The dose distribution was updated according to the estimated seed position to obtain the dose-volume histograms. Postoperative CT-based dosimetry was performed a few hours later using the Theraplan-Plus treatment planning system. The dosimetric parameters obtained from the two imaging modalities were compared. Results: The results of this study revealed correlations between the US- and CT-based dosimetry. However, large variations were found in the implant-quality parameters of the two modalities, including the doses covering 100%, 90%, and 80% of the prostate volume and prostate volumes covered by 100%, 150%, and 200% of the prescription dose. The mean relative difference was 38% and 16% for doses covering 100% and 90% of the prostate volume and 10% and 21% for prostate volumes covered by 100% and 150% of the prescription dose, respectively. The CT-based volume covered by 200% of the prescription dose was about 30% greater than the US-based one. Compared with CT-based dosimetry, US-based dosimetry significantly underestimated the dose to normal organs, especially for the rectum. The average US-based maximal dose and volume covered by 100% of the prescription dose for the rectum was 72 Gy and 0.01 cm 3 , respectively, much lower than the 159 Gy and 0.65 cm 3 obtained using CT-based dosimetry. Conclusion: Although dosimetry using intraoperative US-based planning provides preliminary real-time information, it does not accurately reflect the postoperative CT-based dosimetry. Until studies have determined whether US-based dosimetry or

Dosimetry for radiation processing

International Nuclear Information System (INIS)

Miller, Arne

1986-01-01

During the past few years significant advances have taken place in the different areas of dosimetry for radiation processing, mainly stimulated by the increased interest in radiation for food preservation, plastic processing and sterilization of medical products. Reference services both by international organizations (IAEA) and national laboratories have helped to improve the reliability of dose measurements. In this paper the special features of radiation processing dosimetry are discussed, several commonly used dosimeters are reviewed, and factors leading to traceable and reliable dosimetry are discussed. (author)

Radiochromic film dosimetry

International Nuclear Information System (INIS)

Xu Zhiyong

2002-01-01

Radiochromic film dosimetry was developed to measure ionization irradiation dose for industry and medicine. At this time, there are no comprehensive guideline on the medical application, calibration method and densitometer system for medicine. The review gives update on Radiochromic film dosimetry used for medicine, including principles, film model and material, characteristics, calibration method, scanning densitometer system and medical application

11. International conference on solid radiation dosimetry

International Nuclear Information System (INIS)

Krylova, I.V.

1996-01-01

The main problems discussed during the international conference on solid radiation dosimetry which took place in June 1995 in Budapest are briefly considered. These are the basic physical processes, materials applied for dosimetry, special techniques, personnel monitoring, monitoring of environmental effects, large-dose dosimetry, clinic dosimetry, track detector used for dosimetry, dosimetry in archaeology and geology, equipment and technique for dosimetric measurements. The special attention was paid to superlinearity in the TLD-100 (LiF, Mg, Ti) response function when determining doses of gamma radiation, heavy charged particles, low-energy particle fluxes in particular. New theoretical models were considered

A microcomputer controlled thermoluminescence dosimetry system

International Nuclear Information System (INIS)

Huyskens, C.J.; Kicken, P.J.H.

1980-01-01

Using a microcomputer, an automatic thermoluminescence dosimetry system for personal dosimetry and thermoluminescence detector (TLD) research was developed. Process automation, statistical computation and dose calculation are provided by this microcomputer. Recording of measurement data, as well as dose record keeping for radiological workers is carried out with floppy disk. The microcomputer also provides a human/system interface by means of a video display and a printer. The main features of this dosimetry system are its low cost, high degree of flexibility, high degree of automation and the feasibility for use in routine dosimetry as well as in TLD research. The system is in use for personal dosimetry, environmental dosimetry and for TL-research work. Because of its modular set-up several components of the system are in use for other applications, too. The system seems suited for medium sized health physics groups. (author)

Characterising an aluminium oxide dosimetry system.

Science.gov (United States)

Conheady, Clement F; Gagliardi, Frank M; Ackerly, Trevor

2015-09-01

In vivo dosimetry is recommended as a defence-in-depth strategy in radiotherapy treatments and is currently employed by clinics around the world. The characteristics of a new optically stimulated luminescence dosimetry system were investigated for the purpose of replacing an aging thermoluminescence dosimetry system for in vivo dosimetry. The stability of the system was not sufficient to satisfy commissioning requirements and therefore it has not been released into clinical service at this time.

Solution of the CSEWG - STD-9 shielding standard problem using the coupled systems NJOY/AMPX-II and the JENDL-3 library

International Nuclear Information System (INIS)

Mendoncca, A.G.; Santos, A dos

1993-01-01

The CSEWG STD-9 benchmark has been analyzed to evaluate the shielding and out-of-core methodologies using the coupled systems NJOY/AMPX-II and the JENDL-3 library as well as the DOT 3.5 for the transport calculations. Good agreement between calculated and experimental values for the spatial and energetic neutron flux distributions have been found. However, the results for the reaction rates of the activation foils were discrepant even though for same cases they come into a good agreement. (author)

Dosimetry optimization at COGEMA-La Hague

International Nuclear Information System (INIS)

Kalimbadjian, J.

2000-01-01

At the present time, the la Hague site strives to apply international recommendations together with national regulations concerning radiation protection, and especially the respect of limitation and optimization principles. The application of these principles is based on the implementation of a passive dosimetry and an active dosimetry. The monthly passive dosimetry is monitored by means of a photographic dosimetry film, completed with lithium fluorine thermoluminescent film badges. This personal dosimetry common to X, β, γ and neutron radiations is carried out in close relationship between the Radiation Protection Department, the Occupational Medical Department and the staff running the Plant. The application or ALARA's principle as well as that of radiation protection optimization implies to implement a complementary active dosimetry enabling to gain in real time, the personal dosimetry of each intervening person, either they be COGEMA's workers or external companies'. This active dosimetry provides with following information: This preventive dosimetry is based on the knowledge of doses integration in real time and is fitted with alarm thresholds according to the total amount of doses and dose rates. Thresholds on the dose rate are also set relatively to the radiological environment. This knowledge of doses and dose rates allows a stricter management of the works, while analyzing them according to the nature of the work, to the location and to the skills of the intervening people. This dosimetry allows to analyze and optimize doses integration according to the works nature for the whole intervening staff. The la Hague Site has developed an active personal dosimetry system, common to every intervening person, COGEMA or external companies. The DOSICARD was thus elaborated, shaped as an electronic dosimeter fitted with an alarm and a smart card. The access to controlled areas is conditioned to information given by the DOSICARD concerning medical aptitudes and

The dosimetry of ionizing radiation

CERN Document Server

1990-01-01

A continuation of the treatise The Dosimetry of Ionizing Radiation, Volume III builds upon the foundations of Volumes I and II and the tradition of the preceeding treatise Radiation Dosimetry. Volume III contains three comprehensive chapters on the applications of radiation dosimetry in particular research and medical settings, a chapter on unique and useful detectors, and two chapters on Monte Carlo techniques and their applications.

Secondary standard dosimetry laboratory (SSDL)

International Nuclear Information System (INIS)

Md Saion bin Salikin.

1983-01-01

A secondary Standard Dosimetry Laboratory has been established in the Tun Ismail Research Centre, Malaysia as a national laboratory for reference and standardization purposes in the field of radiation dosimetry. This article gives brief accounts on the general information, development of the facility, programmes to be carried out as well as other information on the relevant aspects of the secondary standard dosimetry laboratory. (author)

Dosimetry in nuclear power plants

International Nuclear Information System (INIS)

Lastra B, J. A.

2008-12-01

To control the occupationally exposed personnel dose working at the Laguna Verde nuclear power plant, two types of dosemeters are used, the thermoluminescent (TLD) which is processed monthly, and the direct reading dosemeter that is electronic and works as daily control of personal dose. In the case of the electronic dosemeters of direct reading conventional, the readings and dose automatic registers and the user identity to which he was assigned to each dosemeter was to carry out the restricted area exit. In activities where the ionizing radiation sources are not fully characterized, it is necessary to relocate the personal dosemeter or assigned auxiliary dosemeters (TLDs and electronics) to determine the dose received by the user to both whole body and in any specific area of it. In jobs more complicated are used a tele dosimetry system where the radiation protection technician can be monitoring the user dose to remote control, the data transmission is by radio. The dosimetry activities are documented in procedures that include dosemeter inventories realization, the equipment and dosemeters calibration, the dosimetry quality control and the discrepancies investigation between the direct reading and TLD systems. TLD dosimetry to have technical expertise in direct and indirect dosimetry and two technicians in TLD dosimetry; electronic dosimetry to have 4 calibration technicians. For the electronic dosemeters are based on a calibrator source of Cesium-137. TLD dosemeters to have an automatic radiator, an automatic reader which can read up to 100 TLD dosemeters per hour and a semiautomatic reader. To keep the equipment under a quality process was development a process of initial entry into service and carried out a periodic verification of the heating cycles. It also has a maintenance contract for the equipment directly with the manufacturer to ensure their proper functioning. The vision in perspective of the dosimetry services of Laguna Verde nuclear power plant

Modern methods of personnel dosimetry

International Nuclear Information System (INIS)

Kraus, W.; Herrmann, D.; Kiesewetter, W.

The physical properties of radiation detectors for personnel dosimetry are described and compared. The suitability of different types of dosimeters for operational and central monitoring of normal occupational exposure, for accident and catastrophe dosimetry and for background and space-flight dosimetry is discussed. The difficulties in interpreting the dosimeter reading with respect to the dose in individual body organs are discussed briefly. 430 literature citations (up to Spring 1966) are given

Theoretical basis for dosimetry

International Nuclear Information System (INIS)

Carlsson, G.A.

1985-01-01

Radiation dosimetry is fundamental to all fields of science dealing with radiation effects and is concerned with problems which are often intricate as hinted above. A firm scientific basis is needed to face increasing demands on accurate dosimetry. This chapter is an attempt to review and to elucidate the elements for such a basis. Quantities suitable for radiation dosimetry have been defined in the unique work to coordinate radiation terminology and usage by the International Commission on Radiation Units and Measurements, ICRU. Basic definitions and terminology used in this chapter conform with the recent ''Radiation Quantities and Units, Report 33'' of the ICRU

Personal dosimetry in Kazakhstan

International Nuclear Information System (INIS)

Khvoshnyanskaya, I.R.; Vdovichenko, V.G.; Lozbin, A.Yu.

2003-01-01

KATEP-AE Radiation Laboratory is the first organization in Kazakhstan officially licensed by the Kazakhstan Atomic Energy Committee to provide individual dosimetry services. The Laboratory was established according to the international standards. Nowadays it is the largest company providing personal dosimetry services in the Republic of Kazakhstan. (author)

MO-B-BRB-00: Three Dimensional Dosimetry

Energy Technology Data Exchange (ETDEWEB)

NONE

2016-06-15

Full three-dimensional (3D) dosimetry using volumetric chemical dosimeters probed by 3D imaging systems has long been a promising technique for the radiation therapy clinic, since it provides a unique methodology for dose measurements in the volume irradiated using complex conformal delivery techniques such as IMRT and VMAT. To date true 3D dosimetry is still not widely practiced in the community; it has been confined to centres of specialized expertise especially for quality assurance or commissioning roles where other dosimetry techniques are difficult to implement. The potential for improved clinical applicability has been advanced considerably in the last decade by the development of improved 3D dosimeters (e.g., radiochromic plastics, radiochromic gel dosimeters and normoxic polymer gel systems) and by improved readout protocols using optical computed tomography or magnetic resonance imaging. In this session, established users of some current 3D chemical dosimeters will briefly review the current status of 3D dosimetry, describe several dosimeters and their appropriate imaging for dose readout, present workflow procedures required for good dosimetry, and analyze some limitations for applications in select settings. We will review the application of 3D dosimetry to various clinical situations describing how 3D approaches can complement other dose delivery validation approaches already available in the clinic. The applications presented will be selected to inform attendees of the unique features provided by full 3D techniques. Learning Objectives: L. John Schreiner: Background and Motivation Understand recent developments enabling clinically practical 3D dosimetry, Appreciate 3D dosimetry workflow and dosimetry procedures, and Observe select examples from the clinic. Sofie Ceberg: Application to dynamic radiotherapy Observe full dosimetry under dynamic radiotherapy during respiratory motion, and Understand how the measurement of high resolution dose data in an

Dosimetry and shielding

International Nuclear Information System (INIS)

Farinelli, U.

1977-01-01

Today, reactor dosimetry and shielding have wide areas of overlap as concerns both problems and methods. Increased interchange of results and know-how would benefit both. The areas of common interest include calculational methods, sensitivity studies, theoretical and experimental benchmarks, cross sections and other nuclear data, multigroup libraries and procedures for their adjustment, experimental techniques and damage functions. This paper reviews the state-of-the-art and the latest development in each of these areas as far as shielding is concerned, and suggests a number of interactions that could be profitable for reactor dosimetry. Among them, re-evaluation of the potentialities of calculational methods (in view of the recent developments) in predicting radiation environments of interest; the application of sensitivity analysis to dosimetry problems; a common effort in the field of theoretical benchmarks; the use of the shielding one-material propagation experiments as reference spectra for detector cross sections; common standardization of the detector nuclear data used in both fields; the setting up of a common (or compatible) multigroup structure and library applicable to shielding, dosimetry and core physics; the exchange of information and experience in the fields of cross section errors, correlations and adjustment; and the intercomparison of experimental techniques

The Latin American Biological Dosimetry Network (LBDNet)

International Nuclear Information System (INIS)

Garcia, O.; Lamadrid, A.I.; Gonzalez, J.E.; Romero, I.; Mandina, T.; Di Giorgio, M.; Radl, A.; Taja, M.R.; Sapienza, C.E.; Deminge, M.M.; Fernandez Rearte, J.; Stuck Oliveira, M.; Valdivia, P.; Guerrero-Carbajal, C.; Arceo Maldonado, C.; Cortina Ramirez, G.E.; Espinoza, M.; Martinez-Lopez, W.; Di Tomasso, M.

2016-01-01

Biological Dosimetry is a necessary support for national radiation protection programmes and emergency response schemes. The Latin American Biological Dosimetry Network (LBDNet) was formally founded in 2007 to provide early biological dosimetry assistance in case of radiation emergencies in the Latin American Region. Here are presented the main topics considered in the foundational document of the network, which comprise: mission, partners, concept of operation, including the mechanism to request support for biological dosimetry assistance in the region, and the network capabilities. The process for network activation and the role of the coordinating laboratory during biological dosimetry emergency response is also presented. This information is preceded by historical remarks on biological dosimetry cooperation in Latin America. A summary of the main experimental and practical results already obtained by the LBDNet is also included. (authors)

Measurement of activation reaction rate distributions in a lead assembly bombarded with 500-MeV protons

CERN Document Server

Takada, H; Sasa, T; Tsujimoto, K; Yasuda, H

2000-01-01

Reaction rate distributions of various activation detectors such as the /sup nat/Ni(n, x)/sup 58/Co, /sup 197/Au(n,2n)/sup 196/Au, and /sup 197/Au(n,4n)/sup 194/Au reactions were measured to study the production and the transport of spallation neutrons in a lead assembly bombarded with protons of 500 MeV. The measured data were analyzed with the nucleon-meson transport code NMTC/JAERI combined with the MCNP4A code using the nuclide production cross sections based on the JENDL Dosimetry File and those calculated with the ALICE-F code. It was found that the NMTC/JAERI-MCNP4A calculations agreed well with the experiments for the low-energy-threshold reaction of /sup nat/Ni(n, x)/sup 58/Co. With the increase of threshold energy, however, the calculation underestimated the experiments, especially above 20 MeV. The reason for the disagreement can be attributed to the underestimation of the neutron yield in the tens of mega-electron-volt regions by the NMTC/JAERI code. (32 refs).

SU-E-T-473: A Patient-Specific QC Paradigm Based On Trajectory Log Files and DICOM Plan Files

International Nuclear Information System (INIS)

DeMarco, J; McCloskey, S; Low, D; Moran, J

2014-01-01

Purpose: To evaluate a remote QC tool for monitoring treatment machine parameters and treatment workflow. Methods: The Varian TrueBeamTM linear accelerator is a digital machine that records machine axis parameters and MLC leaf positions as a function of delivered monitor unit or control point. This information is saved to a binary trajectory log file for every treatment or imaging field in the patient treatment session. A MATLAB analysis routine was developed to parse the trajectory log files for a given patient, compare the expected versus actual machine and MLC positions as well as perform a cross-comparison with the DICOM-RT plan file exported from the treatment planning system. The parsing routine sorts the trajectory log files based on the time and date stamp and generates a sequential report file listing treatment parameters and provides a match relative to the DICOM-RT plan file. Results: The trajectory log parsing-routine was compared against a standard record and verify listing for patients undergoing initial IMRT dosimetry verification and weekly and final chart QC. The complete treatment course was independently verified for 10 patients of varying treatment site and a total of 1267 treatment fields were evaluated including pre-treatment imaging fields where applicable. In the context of IMRT plan verification, eight prostate SBRT plans with 4-arcs per plan were evaluated based on expected versus actual machine axis parameters. The average value for the maximum RMS MLC error was 0.067±0.001mm and 0.066±0.002mm for leaf bank A and B respectively. Conclusion: A real-time QC analysis program was tested using trajectory log files and DICOM-RT plan files. The parsing routine is efficient and able to evaluate all relevant machine axis parameters during a patient treatment course including MLC leaf positions and table positions at time of image acquisition and during treatment

Internal dosimetry, past and future

International Nuclear Information System (INIS)

Johnson, J.R.

1989-03-01

This paper is a review of the progress in the dosimetry of internally deposited radionuclides (internal dosimetry) since World War II. Previous to that, only naturally occurring radionuclides were available and only a limited number of studies of biokinetics and dosimetry were done. The main radionuclides studied were 226 Ra, 228 Ra, and 224 Ra but natural uranium was also studied mainly because of its toxic effect as a heavy metal, and not because it was radioactive. The effects of 226 Ra in bone, mainly from the radium dial painters, also formed the only bases for the radiotoxicity of radionuclides in bone for many years, and it is still, along with 224 Ra, the main source of information on the effects of alpha emitters in bone. The publications of the International Commission on Radiological Protection that have an impact on internal dosimetry are used as mileposts for this review. These series of publications, more than any other, represent a broad consensus of opinion within the radiation protection community at the time of their publication, and have formed the bases for radiation protection practice throughout the world. This review is not meant to be exhaustive; it is meant to be a personnel view of the evolution of internal dosimetry, and to present the author's opinion of what the future directions in internal dosimetry will be. 39 refs., 2 tabs

Individual neutron dosimetry

International Nuclear Information System (INIS)

Mauricio, C.L.P.

1987-01-01

The most important concepts and development in individual neutron dosimetry are presented, especially the dosimetric properties of the albedo technique. The main problem in albedo dosimetry is to calibrate the dosemeter in the environs of each neutron source. Some of the most used calibration techniques are discussed. The IRD albedo dosemeter used in the routine neutron individual monitoring is described in detail. Its dosimetric properties and calibration methods are discussed. (Author) [pt

Advantages and disadvantages of luminescence dosimetry

Energy Technology Data Exchange (ETDEWEB)

Olko, Pawel, E-mail: Pawel.Olko@ifj.edu.p [Institute of Nuclear Physics Polish Academy of Science (IFJ PAN), Krakow (Poland)

2010-03-15

Owing to their excellent dosimetric properties, luminescence detectors of ionizing radiation are now extensively applied in individual dosimetry services. The most frequently used personal dosemeters are based on Optically Stimulated Luminescence (OSL), radiophotoluminescence (RPL) or thermoluminescence (TL). Luminescence detectors have also found several applications in clinical dosimetry, especially around new radiation modalities in radiotherapy, such as Intensity Modulated Radiotherapy (IMRT) or ion beam radiotherapy. Requirements of luminescence detectors applied in individual and clinical dosimetry and some recent developments in luminescence of detectors and techniques leading to significant improvements of the functionality and accuracy of dosimetry systems are reviewed and discussed.

Accidental and retrospective dosimetry using TL method

International Nuclear Information System (INIS)

Mesterházy, D.; Osvay, M.; Kovács, A.; Kelemen, A.

2012-01-01

Retrospective dosimetry is one of the most important tools of accidental dosimetry for dose estimation when dose measurement was not planned. In the affected area many objects can be applied as natural dosimeters. The paper discusses our recent investigations on various electronic components and common salt (NaCl) having useful thermoluminescence (TL) properties. Among materials investigated the electronic components of cell phones seem promising for retrospective dosimetry purposes, having high TL responses, proper glow curve peaks and the intensity of TL peaks vs. gamma dose received provided nearly linear response in the dose range of 10 mGy–1.5 Gy. - Highlights: ► Electronic components and common salt were investigated for accidental and retrospective dosimetry. ► SMD resistors seem promising for retrospective dosimetry purposes. ► Table salt can be used effectively for accidental dosimetry purposes, as well.

Dosimetry and Calibration Section

International Nuclear Information System (INIS)

Otto, T.

1998-01-01

The two tasks of the Dosimetry and Calibration Section at CERN are the Individual Dosimetry Service which assures the personal monitoring of about 5000 persons potentially exposed to ionizing radiation at CERN, and the Calibration Laboratory which verifies all the instruments and monitors. This equipment is used by the sections of the RP Group for assuring radiation protection around CERN's accelerators, and by the Environmental Section of TISTE. In addition, nearly 250 electronic and 300 quartz fibre dosimeters, employed in operational dosimetry, are calibrated at least once a year. The Individual Dosimetry Service uses an extended database (INDOS) which contains information about all the individual doses ever received at CERN. For most of 1997 it was operated without the support of a database administrator as the technician who had assured this work retired. The Software Support Section of TIS-TE took over the technical responsibility of the database, but in view of the many other tasks of this Section and the lack of personnel, only a few interventions for solving immediate problems were possible

Personal dosimetry service of VF, a.s. company

International Nuclear Information System (INIS)

Prasek, P.

2009-01-01

The VF, a.s. Company will extend its services in the area of personal dosimetry at the end of 2008, which is fully in compliance with the requirements of the Atomic Act, section 9 paragraph (1) letter r) and Decree on Radiation Protection, section 59 paragraph (1) letter a). Optically stimulated luminescence was selected in VF .a.s. as the most advantageous and the most advanced technology for the integral personal dosimetry. Optically stimulated luminescence (OSL) has been using in dosimetry for more than ten years. Although it is relatively new technology , its indisputable advantages predetermine that technology has significantly benefited in personal dosimetry services within a short time all over the advanced world. The VF, a.s. personal dosimetry service is based on the licensed products of LANDAUER, the US company, which is the world leader in OSL dosimetry. Crystalline Al 2 O 3 :C was selected as the detection material. All equipment of personal dosimetry service is installed in the VF Centre of Technology in Cerna Hora. The personal dosimetry service is incorporated in the International LANDAUER Dosimetry Service Network, and in the European Union, it is directly linked to the LANDAUER European Headquarters with its office in Paris. As a part of the OSL technology licence, the VF personal dosimetry service was included in the inter-laboratory comparison programme of the LANDAUER syndicate. (author)

Personal dosimetry service of VF, a.s. company

International Nuclear Information System (INIS)

Prasek, P.

2008-01-01

The VF, a.s. Company will extend its services in the area of personal dosimetry at the end of 2008, which is fully in compliance with the requirements of the Atomic Act, section 9 paragraph (1) letter r) and Decree on Radiation Protection, section 59 paragraph (1) letter a). Optically stimulated luminescence was selected in VF .a.s. as the most advantageous and the most advanced technology for the integral personal dosimetry . Optically stimulated luminescence (OSL) has been using in dosimetry for more than ten years. Although it is relatively new technology , its indisputable advantages predetermine that technology has significantly benefited in personal dosimetry services within a short time all over the advanced world. The VF, a.s. personal dosimetry service is based on the licensed products of LANDAUER, the US company, which is the world leader in OSL dosimetry. Crystalline Al 2 O 3 :C was selected as the detection material. All equipment of personal dosimetry service is installed in the VF Centre of Technology in Cerna Hora. The personal dosimetry service is incorporated in the International LANDAUER Dosimetry Service Network, and in the European Union, it is directly linked to the LANDAUER European Headquarters with its office in Paris. As a part of the OSL technology licence, the VF personal dosimetry service was included in the inter-laboratory comparison programme of the LANDAUER syndicate. (author)

Evaluation of cross sections for 14 important neutron-dosimetry reactions

International Nuclear Information System (INIS)

Wagner, M.; Vonach, H.; Pavlik, A.; Strohmaier, B.; Tagesen, S.; Martinez-Rico, J.

1990-01-01

The evaluation of the cross sections for the neutron dosimetry reactions 24 Mg(n,p) 24 Na, 27 Al(n,α) 24 Na, 58 Ni(n,2n) 57 Ni, 64 Zn(n,p) 64 Cu, 90 Zr(n,2n) 89 Zr and 93 Nb(n,n') 93m Nb carried out at the IRK about ten years ago were updated taking into account recent experimental results. Besides, new evaluations were performed for four additional dosimetry reactions, namely 52 Cr(n,2n) 51 Cr, 59 Co(n,2n) 58 Co, 93 Nb(n,2n) 92m Nb and 197 Au(n,2n) 196 Au. The deadlines for the retrieval of data for the different reactions lay between March 1989 and February 1990. The evaluations comprise the neutron energy range from threshold to 20 MeV, in a few cases this range is extended up to 21 MeV or 30 MeV. Cross sections and their uncertainties were evaluated in energy groups with widths of 0.1 MeV to 2.0 MeV, and relative correlation matrices of the evaluated cross sections at different energies were derived. The results of the evaluations are compared to the previous ones and to other recent evaluations reported in the literature. The main results of our previous evaluations for the reactiosn 19 F(n,2n) 18 F, 31 P(n,p) 31 Si, 63 Cu(n,2n) 62 Cu and 103 Rh(n,n') 103m Rh which remain unchanged are also given for completeness. The evaluations reported in this work will be included in the new version of the IRDF (International Reactor Dosimetry File) of the IAEA in ENDF/B-VI format. (orig.)

Neutron personal dosimetry: state-of-art

International Nuclear Information System (INIS)

Spurný, František

2005-03-01

State-of-art of the personal neutron dosimetry is presented, analysed and discussed. Particular attention is devoted to the problems of this type of the dosimetry of external exposure for radiation fields at nuclear power plants. A review of general problems of neutron dosimetry is given and the active individual dosimetry methods available and/or in the stage of development are briefly reviewed. Main attention is devoted to the analysis of the methods available for passive individual neutron dosimetry. The characteristics of these dosemeters were studied and are compared: their energy response functions, detection thresholds and the highest detection limits, the linearity of response, the influence of environmental factors, etc. Particular attention is devoted to their behavior in reactor neutron fields. It is concluded that the choice of the neutron personal dosemeter depends largely on the conditions in which the instrument should be used (neutron spectrum, the level of exposure and the exposure rate, etc.). The results obtained with some of these dosemeters during international intercomparisons are also presented. Particular attention is paid to the personal neutron dosimeter developed and routinely used by National Personal Dosimetry Service Ltd. in the Czech Republic. (author)

Internal dosimetry technical basis manual

International Nuclear Information System (INIS)

1990-01-01

The internal dosimetry program at the Savannah River Site (SRS) consists of radiation protection programs and activities used to detect and evaluate intakes of radioactive material by radiation workers. Examples of such programs are: air monitoring; surface contamination monitoring; personal contamination surveys; radiobioassay; and dose assessment. The objectives of the internal dosimetry program are to demonstrate that the workplace is under control and that workers are not being exposed to radioactive material, and to detect and assess inadvertent intakes in the workplace. The Savannah River Site Internal Dosimetry Technical Basis Manual (TBM) is intended to provide a technical and philosophical discussion of the radiobioassay and dose assessment aspects of the internal dosimetry program. Detailed information on air, surface, and personal contamination surveillance programs is not given in this manual except for how these programs interface with routine and special bioassay programs

Internal dosimetry technical basis manual

Energy Technology Data Exchange (ETDEWEB)

1990-12-20

The internal dosimetry program at the Savannah River Site (SRS) consists of radiation protection programs and activities used to detect and evaluate intakes of radioactive material by radiation workers. Examples of such programs are: air monitoring; surface contamination monitoring; personal contamination surveys; radiobioassay; and dose assessment. The objectives of the internal dosimetry program are to demonstrate that the workplace is under control and that workers are not being exposed to radioactive material, and to detect and assess inadvertent intakes in the workplace. The Savannah River Site Internal Dosimetry Technical Basis Manual (TBM) is intended to provide a technical and philosophical discussion of the radiobioassay and dose assessment aspects of the internal dosimetry program. Detailed information on air, surface, and personal contamination surveillance programs is not given in this manual except for how these programs interface with routine and special bioassay programs.

Dosimetry systems for radiation processing

International Nuclear Information System (INIS)

McLaughlin, W.L.; Desrosiers, M.F.

1995-01-01

Dosimetry serves important functions in radiation processing, where large absorbed doses and dose rates from photon and electron sources have to be measured with reasonable accuracy. Proven dosimetry systems are widely used to perform radiation measurements in development of new processes, validation, qualification and verification (quality control) of established processes and archival documentation of day-to-day and plant-to-plant processing uniformity. Proper calibration and traceability of routine dosimetry systems to standards are crucial to the success of many large-volume radiation processes. Recent innovations and advances in performance of systems that enhance radiation measurement assurance and process diagnostics include dose-mapping media (new radiochromic film and solutions), optical waveguide systems for food irradiation, solid-state devices for real-time and passive dosimetry over wide dose-rate and dose ranges, and improved analytical instruments and data acquisition. (author)

Individual dosimetry of workers and patients: implementation and perspectives; La dosimetrie individuelle des travailleurs et de patients: mise en oeuvre et perspectives

Energy Technology Data Exchange (ETDEWEB)

Rannou, A.; Aubert, B.; Lahaye, Th.; Scaff, P.; Casanova, Ph.; Van Bladel, L.; Queinnec, F.; Valendru, N.; Jehanno, J.; Grude, E.; Berard, Ph.; Desbree, A.; Kafrouni, H.; Paquet, F.; Vanhavere, F.; Bridier, A.; Ginestet, Ch.; Magne, S.; Donadille, L.; Bordy, J.M.; Bottollier-Depois, J.F.; Barrere, J.L.; Ferragut, A.; Metivier, H.; Gaillard-Lecanu, E

2008-07-01

These days organised by the section of the technical protection of the S.F.R.P. review the different techniques of dosimetry used in France and Europe, and present the future orientations.The different interventions are as follow: Individual exposures of the workers: historic assessment and perspectives; medical exposure: where are the doses; legal obligations in individual dosimetry: which are the objective and the need on the subject; the dosimetry follow-up of workers by the S.I.S.E.R.I. system: assessment and perspectives; impact of the norm ISO 20553 on the follow-up of internal exposure; the implementation of the patient dose measurement in Belgium; techniques of passive dosimetry used in Europe; Supervision radiation protection at EDF: long term and short term approach; Comparison active and passive dosimetry at Melox; methodology for the choice of new neutron dosemeters; the working group M.E.D.O.R.: guide of internal dosimetry for the use of practitioners; O.E.D.I.P.E.: tool of modeling for the personalized internal dosimetry; the use of the Monte-Carlo method for the planning of the cancer treatment by radiotherapy becomes a reality; the works of the committee 2 of the ICRP; passive dosimetry versus operational dosimetry: situation in Europe; Implementation of the in vivo dosimetry in a radiotherapy department: experience of the Gustave Roussy institute; experience feedback on the in vivo measures in radiotherapy, based on the use of O.S.L. pellets; multi points O.S.L. instrumentation for the radiation dose monitoring in radiotherapy; dosimetry for extremities for medical applications: principle results of the European contract C.O.N.R.A.D.; references and perspectives in dosimetry; what perspectives for numerical dosimetry, an example: Sievert; system of dose management: how to answer to needs; the last technical evolutions in terms of electronic dosimetry in nuclear power plant; the fourth generation type reactors: what dosimetry. (N.C.)

Information from the Dosimetry Service

CERN Multimedia

2006-01-01

Please note the following opening hours of the Service: From 31st July onwards: Every morning from 8:30 to 12:00 The Service is closed in the afternoons. We should like to remind you that dosimeters cannot be sent to customers by internal mail. Short-term dosimeters (VCTs) must always be returned to the Service after use and must not be left on the racks in the experimental areas or in the secretariats. Dosimetry Service Tel 72155 Bldg. 24 E 011 Dosimetry.service@cern.ch http://cern.ch/rp-dosimetry

Benchmark experiment on molybdenum with graphite by using DT neutrons at JAEA/FNS

Energy Technology Data Exchange (ETDEWEB)

Ohta, Masayuki, E-mail: ohta.masayuki@qst.go.jp [National Institutes for Quantum and Radiological Science and Technology, 2-166 Oaza-Obuchi-Aza-Omotedate, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Kwon, Saerom; Sato, Satoshi [National Institutes for Quantum and Radiological Science and Technology, 2-166 Oaza-Obuchi-Aza-Omotedate, Rokkasho-mura, Kamikita-gun, Aomori (Japan); Konno, Chikara [Japan Atomic Energy Agency, 2-4 Shirakata-Shirane, Tokai-mura, Naka-gun, Ibaraki (Japan); Ochiai, Kentaro [National Institutes for Quantum and Radiological Science and Technology, 2-166 Oaza-Obuchi-Aza-Omotedate, Rokkasho-mura, Kamikita-gun, Aomori (Japan)

2017-01-15

Highlights: • A new benchmark experiment on molybdenum was conducted with DT neutron at JAEA/FNS. • Dosimetry reaction and fission rates were measured in the molybdenum assembly. • Calculated results with MCNP5 code were compared with the measured ones. • A problem on the capture cross section data of molybdenum was pointed out. - Abstract: In our previous benchmark experiment on Mo at JAEA/FNS, we found problems of the (n,2n) and (n,γ) reaction cross sections of Mo in JENDL-4.0 above a few hundred eV. We perform a new benchmark experiment on Mo with a Mo assembly covered with graphite and Li{sub 2}O blocks in order to validate the nuclear data of Mo in lower energy region than in the previous experiment. Several dosimetry reaction and fission rates are measured and compared with calculated ones with the MCNP5-1.40 code and the recent nuclear data libraries, ENDF/B-VII.1, JEFF-3.2, and JENDL-4.0. It is suggested that the (n,γ) reaction cross section of {sup 95}Mo should be larger in the tail region below the large resonance of 45 eV in these nuclear data libraries.

In vivo dosimetry in radiation therapy in Sweden; In vivo-dosimetri inom straalbehandling i Sverige

Energy Technology Data Exchange (ETDEWEB)

Eriksson, Jacob; Blomquist, Michael (Norrlands universitetssjukhus, Umeaa (Sweden))

2010-07-15

A prerequisite for achieving high radiation safety for patients receiving external beam radiation therapy is that the hospitals have a quality assurance program. The program should include include monitoring of the radiation dose given to the patient. Control measurements are performed both at the system level and at the individual level. Control measurement is normally performed using in vivo dosimetry, e.g. a method to measure the radiation dose at the individual level during the actual radiation treatment time. In vivo dosimetry has proven to be an important tool to detect and prevent serious errors in patient treatment. The purpose of this research project was to identify the extent to which vivo dosimetry is used and the methods available for this at Swedish radiation therapy clinics. The authority also wanted to get an overall picture of how hospitals manage results of in vivo dosimetry, and how clinics control radiation dose when using modern treatment techniques. The report reflects the situation in Swedish radiotherapy clinics 2007. The report shows that all hospitals use some form of in vivo dosimetry. The instruments used are mainly diodes and termoluminiscence dosimeters

The Latin American Biological Dosimetry Network (LBDNet).

Science.gov (United States)

García, O; Di Giorgio, M; Radl, A; Taja, M R; Sapienza, C E; Deminge, M M; Fernández Rearte, J; Stuck Oliveira, M; Valdivia, P; Lamadrid, A I; González, J E; Romero, I; Mandina, T; Guerrero-Carbajal, C; ArceoMaldonado, C; Cortina Ramírez, G E; Espinoza, M; Martínez-López, W; Di Tomasso, M

2016-09-01

Biological Dosimetry is a necessary support for national radiation protection programmes and emergency response schemes. The Latin American Biological Dosimetry Network (LBDNet) was formally founded in 2007 to provide early biological dosimetry assistance in case of radiation emergencies in the Latin American Region. Here are presented the main topics considered in the foundational document of the network, which comprise: mission, partners, concept of operation, including the mechanism to request support for biological dosimetry assistance in the region, and the network capabilities. The process for network activation and the role of the coordinating laboratory during biological dosimetry emergency response is also presented. This information is preceded by historical remarks on biological dosimetry cooperation in Latin America. A summary of the main experimental and practical results already obtained by the LBDNet is also included. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Research and innovation in radiation dosimetry

International Nuclear Information System (INIS)

Delgado, A.

1999-01-01

In this article some relevant lines of research in radiation dosimetry are presented. In some of them innovative approaches have been recently proposed in recent years. In others innovation is still to come as it is necessary in view of the insufficiency of the actual methods and techniques. mention is made to Thermoluminescence Dosimetry an to the improvement produced by new computational methods for the analysis of the usually complex TL signals. A solid state dosimetric technique recently proposed, Optically Stimulated Luminescence, OSL, is briefly presented. This technique promises advantages over TLD for personal and environmental dosimetry. The necessity of improving the measurement characteristics of neutron personal dosemeters is commented, making reference to some very recent developments. The situation of the dosimetry in connection with radiobiology research is overviewed, commenting the controversy on the adequacy and utility of the quality absorbed dose for these activities. Finally the special problematic of internal dosimetry is discussed. (Author) 25 refs

Dosimetry control for radiation processing - basic requirements and standards

International Nuclear Information System (INIS)

Ivanova, M.; Tsrunchev, Ts.

2004-01-01

A brief review of the basic international codes and standards for dosimetry control for radiation processing (high doses dosimetry), setting up a dosimetry control for radiation processing and metrology control of the dosimetry system is made. The present state of dosimetry control for food processing and the Bulgarian long experience in food irradiation (three irradiation facilities are operational at these moment) are presented. The absence of neither national standard for high doses nor accredited laboratory for calibration and audit of radiation processing dosimetry systems is also discussed

Secondary gamma-ray skyshine from 14 MeV Neutron Source Facility (OKTAVIAN). Comparison of measurement with its simulation

Energy Technology Data Exchange (ETDEWEB)

Morotomi, Ryutaro; Kondo, Tetsuo; Murata, Isao; Yoshida, Shigeo; Takahashi, Akito [Osaka Univ., Department of Nuclear Engineering, Suita, Osaka (Japan); Yamamoto, Takayoshi [Osaka Univ., Radio Isotope Research Center, Suita, Osaka (Japan)

2000-03-01

Measurement of secondary gamma-ray skyshine was performed at the Intense 14 MeV Neutron Source Facility (OKTAVIAN) of Osaka University with NaI and Hp-Ge detectors. From the result of measurements, some mechanism of secondary gamma-ray skyshine from 14 MeV neutron source facility was found out. The analysis of the measured result were carried out with MCNP-4B for four nuclear data files of JENDL-3.2, JENDL-F.F., FENDL-2, and ENDF/B-VI. It was confirmed that all the nuclear data are fairly reliable for calculations of secondary gamma-ray skyshine. (author)

Dosimetry in dentistry.

Science.gov (United States)

Asha, M L; Chatterjee, Ingita; Patil, Preeti; Naveen, S

2015-01-01

The purpose of this paper was to review various dosimeters used in dentistry and the cumulative results of various studies done with various dosimeters. Several relevant PubMed indexed articles from 1999 to 2013 were electronically searched by typing "dosimeters", "dosimeters in dentistry", "properties of dosimeters", "thermoluminescent and optically stimulated dosimeters", "recent advancements in dosimetry in dentistry." The searches were limited to articles in English to prepare a concise review on dental dosimetry. Titles and abstracts were screened, and articles that fulfilled the criteria of use of dosimeters in dental applications were selected for a full-text reading. Article was divided into four groups: (1) Biological effects of radiation, (2) properties of dosimeters, (3) types of dosimeters and (4) results of various studies using different dosimeters. The present review on dosimetry based on various studies done with dosimeters revealed that, with the advent of radiographic technique the effective dose delivered is low. Therefore, selection of radiological technique plays an important role in dental dose delivery.

Results of the dosimetry intercomparison

International Nuclear Information System (INIS)

Dure, Elsa S.

2000-07-01

The appropriate way to verify the accuracy of the results of dose reported by the laboratories that offer lend personal dosimetry service is in the periodic participation of round of intercomparison dosimetry, undertaken by laboratories whose standards are trace (Secondary Laboratory). The Laboratory of External Personal Dosimetry of the CNEA-PY has participated in three rounds of intercomparison. The first two were organized in the framework of the Model Project RLA/9/030 RADIOLOGICAL WASTE SECURITY, and the irradiations were carried out in the Laboratory of Regional Calibration of the Center of Nuclear Technology Development, Belo Horizonte-Brazil (1998) and in the National Laboratory of Metrology of the ionizing radiations of the Institute of Radioprotection and Dosimetry, Rio de Janeiro-Brazil (1999). The third was organized by the IAEA and the irradiations were made in the Physikalisch-Technische Bundesanstalt PTB, Braunschweig - Federal Republic of Germany (1999-2000) [es

Development of the JAERI computational dosimetry system (JCDS) for boron neutron capture therapy. Cooperative research

CERN Document Server

Kumada, H; Matsumura, A; Nakagawa, Y; Nose, T; Torii, Y; Uchiyama, J; Yamamoto, K; Yamamoto, T

2003-01-01

The Neutron Beam Facility at JRR-4 enables us to carry out boron neutron capture therapy with epithermal neutron beam. In order to make treatment plans for performing the epithermal neutron beam BNCT, it is necessary to estimate radiation doses in a patient's head in advance. The JAERI Computational Dosimetry System (JCDS), which can estimate distributions of radiation doses in a patient's head by simulating in order to support the treatment planning for epithermal neutron beam BNCT, was developed. JCDS is a software that creates a 3-dimentional head model of a patient by using CT and MRI images, and that generates a input data file automatically for calculation of neutron flux and gamma-ray dose distributions in the brain with the Monte Carlo code MCNP, and that displays these dose distributions on the head model for dosimetry by using the MCNP calculation results. JCDS has any advantages as follows; By using CT data and MRI data which are medical images, a detail three-dimensional model of patient's head is...

Dosimetry standards for radiation processing

International Nuclear Information System (INIS)

Farrar, H. IV

1999-01-01

For irradiation treatments to be reproducible in the laboratory and then in the commercial environment, and for products to have certified absorbed doses, standardized dosimetry techniques are needed. This need is being satisfied by standards being developed by experts from around the world under the auspices of Subcommittee E10.01 of the American Society for Testing and Materials (ASTM). In the time period since it was formed in 1984, the subcommittee has grown to 150 members from 43 countries, representing a broad cross-section of industry, government and university interests. With cooperation from other international organizations, it has taken the combined part-time effort of all these people more than 13 years to complete 24 dosimetry standards. Four are specifically for food irradiation or agricultural applications, but the majority apply to all forms of gamma, x-ray, Bremsstrahlung and electron beam radiation processing, including dosimetry for sterilization of health care products and the radiation processing of fruits, vegetables, meats, spices, processed foods, plastics, inks, medical wastes and paper. An additional 6 standards are under development. Most of the standards provide exact procedures for using individual dosimetry systems or for characterizing various types of irradiation facilities, but one covers the selection and calibration of dosimetry systems, and another covers the treatment of uncertainties. Together, this set of standards covers essentially all aspects of dosimetry for radiation processing. The first 20 of these standards have been adopted in their present form by the International Organization of Standardization (ISO), and will be published by ISO in 1999. (author)

Individual dosimetry of workers and patients: implementation and perspectives

International Nuclear Information System (INIS)

Rannou, A.; Aubert, B.; Lahaye, Th.; Scaff, P.; Casanova, Ph.; Van Bladel, L.; Queinnec, F.; Valendru, N.; Jehanno, J.; Grude, E.; Berard, Ph.; Desbree, A.; Kafrouni, H.; Paquet, F.; Vanhavere, F.; Bridier, A.; Ginestet, Ch.; Magne, S.; Donadille, L.; Bordy, J.M.; Bottollier-Depois, J.F.; Barrere, J.L.; Ferragut, A.; Metivier, H.; Gaillard-Lecanu, E.

2008-01-01

These days organised by the section of the technical protection of the S.F.R.P. review the different techniques of dosimetry used in France and Europe, and present the future orientations.The different interventions are as follow: Individual exposures of the workers: historic assessment and perspectives; medical exposure: where are the doses; legal obligations in individual dosimetry: which are the objective and the need on the subject; the dosimetry follow-up of workers by the S.I.S.E.R.I. system: assessment and perspectives; impact of the norm ISO 20553 on the follow-up of internal exposure; the implementation of the patient dose measurement in Belgium; techniques of passive dosimetry used in Europe; Supervision radiation protection at EDF: long term and short term approach; Comparison active and passive dosimetry at Melox; methodology for the choice of new neutron dosemeters; the working group M.E.D.O.R.: guide of internal dosimetry for the use of practitioners; O.E.D.I.P.E.: tool of modeling for the personalized internal dosimetry; the use of the Monte-Carlo method for the planning of the cancer treatment by radiotherapy becomes a reality; the works of the committee 2 of the ICRP; passive dosimetry versus operational dosimetry: situation in Europe; Implementation of the in vivo dosimetry in a radiotherapy department: experience of the Gustave Roussy institute; experience feedback on the in vivo measures in radiotherapy, based on the use of O.S.L. pellets; multi points O.S.L. instrumentation for the radiation dose monitoring in radiotherapy; dosimetry for extremities for medical applications: principle results of the European contract C.O.N.R.A.D.; references and perspectives in dosimetry; what perspectives for numerical dosimetry, an example: Sievert; system of dose management: how to answer to needs; the last technical evolutions in terms of electronic dosimetry in nuclear power plant; the fourth generation type reactors: what dosimetry. (N.C.)

Development of improved procedures for evaluation of neutron cross sections for reactor neutron dosimetry

International Nuclear Information System (INIS)

Vonach, H.

1980-06-01

The cross-sections for the four important neutron dosimetry reactions 19 F(n,2n) 18 F, 31 P(n,p) 31 Si, 93 Nb(n,n')sup(93m)Nb and 103 Rh(n,n')sup(103m)Rh were evaluated in the neutron energy range from threshold to 20 MeV. For the 19 F(n,2n) reaction the evaluation could be based entirely on experimental data; for the reactions 31 P(n,p) 31 Si and 103 Rh(n,n')sup(103m)Rh large gaps in the experimental excitation functions and large discrepancies between the existing data made it necessary to supplement the experimental data by cross-section calculations and to give about equal weight to the experimental and calculated cross-sections. For the 93 Nb(n,n')sup(93m)Nb reaction the evaluation had to be based entirely on the theoretically calculated cross-sections. The cross-section calculations were performed using the statistical model of nuclear reactions allowing for precompound processes in the first reaction step and errors of the calculated cross-sections were estimated from their sensitivity to the various input parameters. Cross-section values were evaluated for energy groups between 0.1 MeV and 1 MeV wide, the width depending on both the slope of the excitation functions and the density of the available data. For each evaluated cross-section also an uncertainty (on a 1 sigma confidence level) was derived taking into account the errors given by the experimentalists, the general consistency of the experimental data and the estimated errors of the theoretically calculated cross-sections. In addition relative correlation matrices were derived for each evaluated excitation function describing the correlations between the uncertainties of the cross-sections at different energies. The correlations between the cross-section uncertainties for different reactions were found to be negligible. The results of this evaluation as well as those of Ref. 1 will be combined with the ENDF/B-V dosimetry file into an international neutron dosimetry file by the nuclear data section of

The dosimetry programme of the IAEA

International Nuclear Information System (INIS)

1987-01-01

Describes the activities of the IAEA's Dosimetry Laboratory which provides calibration and comparison services for secondary standard dosimetry laboratories (SSDLs) of Member States. In addition, a joint IAEA/WHO postal dosimetry service has been established for radiotherapy centers. The International Measurement System and the calibration ''chain'' from measurement standard instruments of the International Bureau of Weights and Measurements (BIPM) through the primary and secondary standards to the dosimeters of the users are presented as well

Hematological dosimetry

International Nuclear Information System (INIS)

Fluery-Herard, A.

1991-01-01

The principles of hematological dosimetry after acute or protracted whole-body irradiation are reviewed. In both cases, over-exposure is never homogeneous and the clinical consequences, viz medullary aplasia, are directly associated with the mean absorbed dose and the seriousness and location of the overexposure. The main hematological data required to assess the seriousness of exposure are the following: repeated blood analysis, blood precursor cultures, as indicators of whole-body exposure; bone marrow puncture, medullary precursor cultures and medullary scintigraphy as indicators of the importance of a local over-exposure and capacity for spontaneous repair. These paraclinical investigations, which are essential for diagnosis and dosimetry, are also used for surveillance and for the main therapeutic issues [fr

Double differential charged particle emission cross sections of vanadium for 14.1 MeV incident neutrons

Energy Technology Data Exchange (ETDEWEB)

Kokooo; Murata, Isao; Takahashi, Akito [Osaka Univ., Suita (Japan). Faculty of Engineering

1997-03-01

The double differential cross sections of V(n,xp) and V(n,x{alpha}) reactions have been measured by using the E-TOF spectrometer. The measured data were compared with other experimental data and evaluated nuclear data of JENDL Fusion-File. (author)

Development and current state of dosimetry in Cuba

International Nuclear Information System (INIS)

Prieto Miranda, E.F.; Cuesta Fuente, G.; Chavez Ardanza, A.

1999-01-01

In Cuba, the application of the radiation technologies has been growing in the last years, and at present there are several dosimetry systems with different ranges of absorbed dose. Diverse researches were carried out on high dose dosimetry with the following dosimetry systems: Fricke, ceric-cerous sulfate, ethanol-chlorobenzene, cupric sulfate and Perspex (Red 4034 AE and Clear HX). In this paper the development achieved during the last 15 years in the high dose dosimetry for radiation processing in Cuba is presented, as well as, the current state of different dosimetry systems employed for standardization and for process control. The paper also reports the results of dosimetry intercomparison studies that were performed with the Ezeiza Atomic Center of Argentine and the International Dose Assurance Service (IDAS) of IAEA. (author)

Miniature semiconductor detectors for in vivo dosimetry

International Nuclear Information System (INIS)

Rosenfeld, A. B.; Cutajar, D.; Lerch, M. L. F.; Takacs, G.; Cornelius, I. M.; Yudelev, M.; Zaider, M.

2006-01-01

Silicon mini-semiconductor detectors are found in wide applications for in vivo personal dosimetry and dosimetry and Micro-dosimetry of different radiation oncology modalities. These applications are based on integral and spectroscopy modes of metal oxide semiconductor field effect transistor and silicon p-n junction detectors. The advantages and limitations of each are discussed. (authors)

Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

CERN Document Server

Chiba, G

2002-01-01

It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

Chemical dosimetry system for criticality accidents.

Science.gov (United States)

Miljanić, Saveta; Ilijas, Boris

2004-01-01

Ruder Bosković Institute (RBI) criticality dosimetry system consists of a chemical dosimetry system for measuring the total (neutron + gamma) dose, and a thermoluminescent (TL) dosimetry system for a separate determination of the gamma ray component. The use of the chemical dosemeter solution chlorobenzene-ethanol-trimethylpentane (CET) is based on the radiolytic formation of hydrochloric acid, which protonates a pH indicator, thymolsulphonphthalein. The high molar absorptivity of its red form at 552 nm is responsible for a high sensitivity of the system: doses in the range 0.2-15 Gy can be measured. The dosemeter has been designed as a glass ampoule filled with the CET solution and inserted into a pen-shaped plastic holder. For dose determinations, a newly constructed optoelectronic reader has been used. The RBI team took part in the International Intercomparison of Criticality Accident Dosimetry Systems at the SILENE Reactor, Valduc, June 2002, with the CET dosimetry system. For gamma ray dose determination TLD-700 TL detectors were used. The results obtained with CET dosemeter show very good agreement with the reference values.

Recent developments in neutron dosimetry and radiation damage calculations for fusion-materials studies

International Nuclear Information System (INIS)

Greenwood, L.R.

1983-01-01

This paper is intended as an overview of activities designed to characterize neutron irradiation facilities in terms of neutron flux and energy spectrum and to use these data to calculate atomic displacements, gas production, and transmutation during fusion materials irradiations. A new computerized data file, called DOSFILE, has recently been developed to record dosimetry and damage data from a wide variety of materials test facilities. At present data are included from 20 different irradiations at fast and mixed-spectrum reactors, T(d,n) 14 MeV neutron sources, Be(d,n) broad-spectrum sources, and spallation neutron sources. Each file entry includes activation data, adjusted neutron flux and spectral data, and calculated atomic displacements and gas production. Such data will be used by materials experimenters to determine the exposure of their samples during specific irradiations. This data base will play an important role in correlating property changes between different facilities and, eventually, in predicting materials performance in fusion reactors. All known uncertainties and covariances are listed for each data record and explicit references are given to nuclear decay data and cross sections

Dosimetry as an integral part of radiation processing

International Nuclear Information System (INIS)

Zagorski, Z.P.

1999-01-01

Different connections between high-dose dosimetry and radiation processing are discussed. Radiation processing cannot be performed without proper dosimetry. Accurate high dose and high dose rate dosimetry exhibits several aspects: first of all it is the preservation of the quality of the product, then fulfillment of legal aspects and last but not the least the safety of processing. Further, seldom discussed topics are as follow: dosimetric problems occurring with double-side EB irradiations, discussed in connection with the deposition of electric charge during electron beam irradiation. Although dosimetry for basic research and for medical purposes are treated here only shortly, some conclusions reached from these fields are considered in dosimetry for radiation processing. High-dose dosimetry of radiation has become a separate field, with many papers published every year, but applied dosimetric projects are usually initiated by a necessity of particular application. (author)

The work programme of EURADOS on internal and external dosimetry.

Science.gov (United States)

Rühm, W; Bottollier-Depois, J F; Gilvin, P; Harrison, R; Knežević, Ž; Lopez, M A; Tanner, R; Vargas, A; Woda, C

2018-01-01

Since the early 1980s, the European Radiation Dosimetry Group (EURADOS) has been maintaining a network of institutions interested in the dosimetry of ionising radiation. As of 2017, this network includes more than 70 institutions (research centres, dosimetry services, university institutes, etc.), and the EURADOS database lists more than 500 scientists who contribute to the EURADOS mission, which is to promote research and technical development in dosimetry and its implementation into practice, and to contribute to harmonisation of dosimetry in Europe and its conformance with international practices. The EURADOS working programme is organised into eight working groups dealing with environmental, computational, internal, and retrospective dosimetry; dosimetry in medical imaging; dosimetry in radiotherapy; dosimetry in high-energy radiation fields; and harmonisation of individual monitoring. Results are published as freely available EURADOS reports and in the peer-reviewed scientific literature. Moreover, EURADOS organises winter schools and training courses on various aspects relevant for radiation dosimetry, and formulates the strategic research needs in dosimetry important for Europe. This paper gives an overview on the most important EURADOS activities. More details can be found at www.eurados.org .

Dosimetry and operation of irradiation facilities

International Nuclear Information System (INIS)

Vidal, P.E.

1985-01-01

The industrial use of ionizing radiation has required, from the very first, the measurement of delivered and absorbed doses; hence the necessity of providing dosimetric systems. Laboratories, scientists, industries and potential equipment manufacturers have all collaborated in this new field of activity. Dosimetric intercomparisons have been made by each industry at their own facilities and in collaboration with specialists, national organizations and the IAEA. Dosimetry has become a way of ensuring that treatment by irradiation has been carried out in accordance with the rules. It has become in effect assurance of quality. Routine dosimetry should determine a maximum and minimum dose. Numerous factors play a part in dosimetry. Industry is currently in possession of routine dosimetric systems that are sufficiently accurate, fairly easy to handle and reasonable in cost, thereby satisfying all the requirements of industry and the need for control. Dosimetry is important in the process of marketing irradiated products. The operator of an industrial irradiation facility bases his dosimetry on comparison with reference systems. Research aimed at simplifying the practice of routine dosimetry should be continued. New physical and chemical techniques will be incorporated into systems already in use. The introduction of microcomputers into the operation of radiation facilities has increased the value of dosimetry and made the conditions of treatment more widespread. Stress should be placed on research in several areas apart from reference systems, for example: dosimetric systems at temperatures from +8 deg. C to -45 deg. C, over the dose range 100 krad to a little more than 1 Mrad, liquids and fluidized solids carried at high speed through ducts, thin-film liquids circulating at a high flow rate, and various other problems. (author)

Foundations of ionizing radiation dosimetry

International Nuclear Information System (INIS)

Denisenko, O.N.; Pereslegin, I.A.

1985-01-01

Foundations of dosimetry in application to radiotherapy are presented. General characteristics of ionizing radiations and main characteristics of ionizing radiation sources, mostly used in radiotherapy, are given. Values and units for measuring ionizing radiation (activity of a radioactive substance, absorbed dose, exposure dose, integral dose and dose equivalent are considered. Different methods and instruments for ionizing radiation dosimetry are discussed. The attention is paid to the foundations of clinical dosimetry (representation of anatomo-topographic information, choice of radiation conditions, realization of radiation methods, corrections for a configuration and inhomogeneity of a patient's body, account of biological factors of radiation effects, instruments of dose field formation, control of irradiation procedure chosen)

Thermoluminescence in medical dosimetry; Termoluminiscencia en dosimetria medica

Energy Technology Data Exchange (ETDEWEB)

Rivera, T., E-mail: trivera@ipn.mx [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Av. Legaria 694, Col. Irrigacion, 11500 Mexico D. F. (Mexico)

2011-10-15

The dosimetry by thermoluminescence (Tl) is applied in the entire world for the dosimetry of ionizing radiations specially to personal and medical dosimetry. This dosimetry method has been very interesting for measures in vivo because the Tl dosimeters have the advantage of being very sensitive in a very small volume and they are also equivalent to tissue and they do not need additional accessories (for example, cable, electrometer, etc.) The main characteristics of the diverse Tl materials to be used in the radiation measures and practical applications are: the Tl curve, the share homogeneity, the signal stability after the irradiation, precision and exactitude, the response in function with the dose and the energy influence. In this work a brief summary of the advances of the radiations dosimetry is presented by means of the thermally stimulated luminescence and its application to the dosimetry in radiotherapy. (Author)

Radiochromic film dosimetry

International Nuclear Information System (INIS)

Soares, Christopher G.

2006-01-01

The object of this paper is to give a new user some practical information on the use of radiochromic films for medical applications. While various aspects of radiochromic film dosimetry for medical applications have been covered in some detail in several other excellent review articles which have appeared in the last few years [Niroomand-Rad, A., Blackwell, C.R., Coursey, B.M., Gall, K.P., McLaughlin, W.L., Meigooni, A.S., Nath, R., Rodgers, J.E., Soares, C.G., 1998. Radiochromic dosimetry: recommendations of the AAPM Radiation Therapy Committee Task Group 55. Med. Phys. 25, 2093-2115; Dempsey, J.F., Low, D.A., Mutic, S., Markman, J., Kirov, A.S., Nussbaum, G.H., Williamson, J.F., 2000. Validation of a precision radiochromic film dosimetry system for quantitative two-dimensional imaging of acute exposure dose distributions. Med. Phys. 27, 2462-2475; Butson, M.J., Yu, P.K.N., Cheung, T., Metcalfe, P., 2003. Radiochromic film for medical radiation dosimetry. Mater. Sci. Eng. R41, 61-120], it is the intent of the present author to present material from a more user-oriented and practical standpoint. That is, how the films work will be stressed much less than how to make the films work well. The strength of radiochromic films is most evident in applications where there is a very high dose gradient and relatively high absorbed dose rates. These conditions are associated with brachytherapy applications, measurement of small fields, and at the edges (penumbra regions) of larger fields

Performing personnel dosimetry investigations and records quality assurance

International Nuclear Information System (INIS)

Perle, S.C.

2002-01-01

Radiation Safety Officers (RSOs) sometimes face situations in which personnel dosimetry estimates are required after dosimeters issued to radiation workers (film or TLD badges, extremity dosimeters, etc.) are lost or damaged before processing. This article was prepared to help those involved with personnel dosimetry investigations became aquatinted with this process. A factor that contributes to the anxiety of those unfamiliar with dosimetry investigations is the lack of published guidance available in this subject. More printed resources are needed to help radiation safety professionals familiarize themselves and understand personnel dosimetry investigations. Topics discussed in this presentation include the justification of performing dosimetry investigations, recommendations on how to perform them and the advantages of performing such investigations

Dosimetry and Shielding of X and Gamma Radiation

International Nuclear Information System (INIS)

Oncescu, M.; Panaitescu, I.

1992-01-01

This book covers the following problems: 1. X and Gamma radiations, 2. Interaction of X-ray and gamma radiations with matter, 3. Interaction of electrons with matter, 4. Principles and basic concepts of dosimetry, 5. Ionization dosimetry, 6. Calorimetric chemical and photographic dosimetry, 7. Solid state dosimetry, 8. Computation of dosimetric quantities, 9. Dosimetry in radiation protection, 10. Shielding of X and gamma radiations. The authors, well-known Romanian experts in Radiation Physics and Engineering, gave an up-dated, complete and readable account of this subject matter. The analyses of physical principles and concepts, of materials and instruments and of computational methods and applications are all well balanced to meat the needs of a broad readership

Radiotherapy Based On α Emitting Radionuclides: Geant4 For Dosimetry And Micro-/Nano-Dosimetry

International Nuclear Information System (INIS)

Guatelli, Susanna

2013-01-01

Possible physics approaches to evaluate the efficacy of TAT are dosimetry, microdosimetry and nanodosimetry. Dosimetry is adequate when mean absorbed dose to a macroscopic target volume is important to understand the biological effect of radiation. General purpose Monte Carlo (MC) codes, based on condensed history approach, are a very useful, cost effective tool to solve dosimetric problems. The condensed history approach is based on the use of multiple scattering theories to calculate the energy losses and angular changes in the direction of the particle. The short α particle range and high LET make the microdosimetric approach more suitable than dosimetry to study TAT from first physics principles, as this approach takes into account the stochastic nature of energy deposition at cellular level

Updating the INDAC computer application of internal dosimetry

International Nuclear Information System (INIS)

Bravo Perez-Tinao, B.; Marchena Gonzalez, P.; Sollet Sanudo, E.; Serrano Calvo, E.

2013-01-01

The initial objective of this project is to expand the application INDAC currently used in internal dosimetry services of the Spanish nuclear power plants and Tecnatom for estimating the effective doses of internal dosimetry of workers in direct action. or in-vivo dosimetry. (Author)

Dosimetry methods

DEFF Research Database (Denmark)

McLaughlin, W.L.; Miller, A.; Kovacs, A.

2003-01-01

Chemical and physical radiation dosimetry methods, used for the measurement of absorbed dose mainly during the practical use of ionizing radiation, are discussed with respect to their characteristics and fields of application....

Radiation dosimetry in nuclear medicine

International Nuclear Information System (INIS)

Stabin, M.G.; Tagesson, M.; Ljungberg, M.; Strand, S.E.; Thomas, S.R.

1999-01-01

Radionuclides are used in nuclear medicine in a variety of diagnostic and therapeutic procedures. A knowledge of the radiation dose received by different organs in the body is essential to an evaluation of the risks and benefits of any procedure. In this paper, current methods for internal dosimetry are reviewed, as they are applied in nuclear medicine. Particularly, the Medical Internal Radiation Dose (MIRD) system for dosimetry is explained, and many of its published resources discussed. Available models representing individuals of different age and gender, including those representing the pregnant woman are described; current trends in establishing models for individual patients are also evaluated. The proper design of kinetic studies for establishing radiation doses for radiopharmaceuticals is discussed. An overview of how to use information obtained in a dosimetry study, including that of the effective dose equivalent (ICRP 30) and effective dose (ICRP 60), is given. Current trends and issues in internal dosimetry, including the calculation of patient-specific doses and in the use of small scale and microdosimetry techniques, are also reviewed

An IAEA Survey of Dosimetry Audit Networks for Radiotherapy

International Nuclear Information System (INIS)

Grochowska, Paulina; Izewska, Joanna

2013-01-01

A Survey: In 2010, the IAEA undertook a task to investigate and review the coverage and operations of national and international dosimetry audit programmes for radiotherapy. The aim was to organize the global database describing the activities of dosimetry audit networks in radiotherapy. A dosimetry audit questionnaire has been designed at an IAEA consultants' meeting held in 2010 for organizations conducting various levels of dosimetry audits for radiotherapy. Using this questionnaire, a survey was conducted for the first time in 2010 and repeated in 2011. Request for information on different aspects of the dosimetry audit was included, such as the audit framework and resources, its coverage and scope, the dosimetry system used and the modes of audit operation, i.e. remotely and through on-site visits. The IAEA questionnaire was sent to over 80 organizations, members of the IAEA/WHO Network of Secondary Standards Dosimetry Laboratories (SSDLs) and other organizations known for having operated dosimetry audits for radiotherapy in their countries or internationally. Survey results and discussion: In response to the IAEA survey, 53 organizations in 45 countries confirmed that they operate dosimetry audit services for radiotherapy. Mostly, audits are conducted nationally, however there are five organizations offering audits abroad, with two of them operating in various parts of the world and three of them at the regional level, auditing radiotherapy centres in neighbouring countries. The distribution of dosimetry audit services in the world is given. (author)

Personnel neutron dosimetry at Department of Energy facilities

International Nuclear Information System (INIS)

Brackenbush, L.W.; Endres, G.W.R.; Selby, J.M.; Vallario, E.J.

1980-08-01

This study assesses the state of personnel neutron dosimetry at DOE facilities. A survey of the personnel dosimetry systems in use at major DOE facilities was conducted, a literature search was made to determine recent advances in neutron dosimetry, and several dosimetry experts were interviewed. It was concluded that personnel neutron dosimeters do not meet current needs and that serious problems exist now and will increase in the future if neutron quality factors are increased and/or dose limits are lowered

Internal dosimetry hazard and risk assessments: methods and applications

International Nuclear Information System (INIS)

Roberts, G.A.

2006-01-01

Routine internal dose exposures are typically (in the UK nuclear industry) less than external dose exposures: however, the costs of internal dosimetry monitoring programmes can be significantly greater than those for external dosimetry. For this reason decisions on when to apply routine monitoring programmes, and the nature of these programmes, can be more critical than for external dosimetry programmes. This paper describes various methods for performing hazard and risk assessments which are being developed by RWE NUKEM Limited Approved Dosimetry Services to provide an indication when routine internal dosimetry monitoring should be considered. (author)

Analysis of OKTAVIAN Shielding Benchmark Experiments by ENDF/B-VII, JEFF-3.1, and JENDL-3.3

International Nuclear Information System (INIS)

Kim, Do-Heon; Gil, Choong-Sup; Lee, Young-Ouk

2007-01-01

International collaborations for the ITER Project and other fusion-related development projects have been conducted to create a reference fusion nuclear data library such as FENDL, which was a collection of the best cross section data from national nuclear data libraries. Recent release of newly evaluated nuclear data libraries requires an extensive and intensive benchmarking of the updated transport libraries to become a candidate for the future collection. In this study, the pulsed sphere experiments for leakage neutron and gamma-ray spectra at the D-T neutron source facility of Osaka University, OKTAVIAN were employed to test the ENDF/B-VII beta 1, JEFF-3.1, and JENDL-3.3 libraries. The continuous energy Monte Carlo transport code MCNPX-2.5 was used along with the ACE format libraries processed by a modified version of the NJOY99.90 code

Radiation protection dosimetry in medicine - Report of the working group n.9 of the European radiation dosimetry group (EURADOS) - coordinated network for radiation dosimetry (CONRAD - contract EC N) fp6-12684; Dosimetrie pour la radioprotection en milieu medical - rapport du groupe de travail n. 9 du European radiation dosimetry group (EURADOS) - coordinated netword for radiation dosimetry (CONRAD - contrat CE fp6-12684)

Energy Technology Data Exchange (ETDEWEB)

NONE

2009-07-01

This report present the results achieved within the frame of the work the WP 7 (Radiation Protection Dosimetry of Medical Staff) of the coordination action CONRAD (Coordinated Network for Radiation Dosimetry) funded through the 6. EU Framework Program. This action was coordinated by EURADOS (European Radiation Dosimetry Group). EURADOS is an organization founded in 1981 to advance the scientific understanding and the technical development of the dosimetry of ionising radiation in the fields of radiation protection, radiobiology, radiation therapy and medical diagnosis by promoting collaboration between European laboratories. WP7 coordinates and promotes European research for the assessment of occupational exposures to staff in therapeutic and diagnostic radiology workplaces. Research is coordinated through sub-groups covering three specific areas: 1. Extremity dosimetry in nuclear medicine and interventional radiology: this sub-group coordinates investigations in the specific fields of the hospitals and studies of doses to different parts of the hands, arms, legs and feet; 2. Practice of double dosimetry: this sub-group reviews and evaluates the different methods and algorithms for the use of dosemeters placed above and below lead aprons in large exposure during interventional radiology procedures, especially to determine effective doses to cardiologists during cardiac catheterization; and 3. Use of electronic personal dosemeters in interventional radiology: this sub-group coordinates investigations in laboratories and hospitals, and intercomparisons with passive dosemeters with the aim to enable the formulation of standards. (authors)

Neutron personnel dosimetry considerations for fusion reactors

International Nuclear Information System (INIS)

Barton, T.P.; Easterly, C.E.

1979-07-01

The increasing development of fusion reactor technology warrants an evaluation of personnel neutron dosimetry systems to aid in the concurrent development of a radiation protection program. For this reason, current state of knowledge neutron dosimeters have been reviewed with emphasis placed on practical utilization and the problems inherent in each type of dosimetry system. Evaluations of salient parameters such as energy response, latent image instability, and minimum detectable dose equivalent are presented for nuclear emulsion films, track etch techniques, albedo and other thermoluminescent dosimetry techniques, electrical conductivity damage effects, lyoluminescence, thermocurrent, and thermally stimulated exoelectron emission. Brief summaries of dosimetry regulatory requirements and intercomparison study results help to establish compliance and recent trends, respectively. Spectrum modeling data generated by the Neutron Physics Division of Oak Ridge National Laboratory for the Princeton Tokamak Fusion Test Reactor (TFTR) Facility have been analyzed by both International Commission on Radiological Protection fluence to dose conversion factors and an adjoint technique of radiation dosimetry, in an attempt to determine the applicability of current neutron dosimetry systems to deuterium and tritium fusion reactor leakage spectra. Based on the modeling data, a wide range of neutron energies will probably be present in the leakage spectra of the TFTR facility, and no appreciable risk of somatic injury to occupationally exposed workers is expected. The relative dose contributions due to high energy and thermal neutrons indicate that neutron dosimetry will probably not be a serious limitation in the development of fusion power

Accuracy Requirements in Medical Radiation Dosimetry

International Nuclear Information System (INIS)

Andreo, P.

2011-01-01

The need for adopting unambiguous terminology on 'accuracy in medical radiation dosimetry' which is consistent with international recommendations for metrology is emphasized. Uncertainties attainable, or the need for improving their estimates, are analysed for the fields of radiotherapy, diagnostic radiology and nuclear medicine dosimetry. This review centres on uncertainties related to the first step of the dosimetry chain in the three fields, which in all cases involves the use of a detector calibrated by a standards laboratory to determine absorbed dose, air kerma or activity under reference conditions in a clinical environment. (author)

Characterization of commercial MOSFETS electron dosimetry

International Nuclear Information System (INIS)

Carvajal, M. A.; Simancas, F.; Guirado, D.; Banqueri, J.; Vilches, M.; Lallena, A. M.; Palma, A. J.

2011-01-01

In recent years there have been commercial dosimetry devices based on transistors Metal-Oxide-Semiconductor (MOSFET) having a number of advantages over traditional systems for dosimetry in medical applications. These include the portability of the sensor element and a reading process quick and relatively simple dose, linearity, and so on. The use of electron beams is important in modern radiotherapy include its use in intra-operative radiotherapy (RIO). This paper presents an initial characterization of different business models MOSFET, not specific for radiation detection, to demonstrate their potential as sensors for electron beam dosimetry. (Author)

Radiographic film orientation in radiotherapy dosimetry

International Nuclear Information System (INIS)

Suchowerska, N.; Davison, A.; Drew, J.; Metcalfe, P.

1996-01-01

Since the discovery of x-rays, film has been used as a detection medium for radiation. More recently radiographic film has become established as a practical tool for the measurement of dose distribution in radiotherapy. The accuracy and reproducibility of film dosimetry depends on photon energy, processing conditions and film plane orientation. The relationship between photon energy, processing conditions and film dosimetry accuracy has been studied. The role of film plane orientation is still controversial. The current work aims to clarify the effects film plane orientation has on film dosimetry. Poster 205. (author)

Sensitivity Study for the Keff of a Lead Based ADS on the Cross Sections Evaluations: ENDFB-6.4 and JENDL-3.2

International Nuclear Information System (INIS)

Embid, M.; Fernandez, R.; Gonzalez, E.

1998-05-01

In any simulation of the neutronic behaviour of a nuclear system, the accuracy of the results is limited by the precision of the nuclear databases used, in particular for the cross sections. In the study of Accelerator Driven System (ADS) for transmutation appear aspects new for the nuclear systems. The best example is the very important role played by unusual isotopes, particularly lead and some transuranics. In the present paper, the results of a comparison of the predicted behaviour of one Th-TRU and lead based ADS according to the ENDFB-6.4 or the JENDL-3.2 databases will be presented

The personal dosimetry in Mexico

International Nuclear Information System (INIS)

Salazar, M.A.

2006-01-01

The Personal Dosimetry in Mexico, has an approximately 30 year-old history; and it had been and it is at the moment, one of the more important resources with which the personnel that works with ionizing radiation sources counts for its protection. The Personal Dosimetry begins with the film dosimetry, technique that even continues being used at the present time by some users, and the main reason of its use is for economic reasons. At the moment this technique, it has been surpassed, by the Thermoluminescent dosimetry, which has taken a lot of peak, mainly by the technological development with which it is counted at the present time; what has given as a result that this technique becomes tip technology; that supported in the characteristic of the used materials, as the handling and processing of the information associated with the new PC, digitizer cards, software etc, what has allowed increases it potential. In this work the current necessities of the market are presented as well as an analysis of the future real necessities in Mexico, at national level, the companies that provide this service and that they spread to satisfy this necessity of the market, including the different used technologies are also mentioned. The application ranges, at the same time, of the advantages and disadvantages of the different systems of Personal Dosimetry in the market. The companies that at the moment provide the service of Personal Dosimetry, its use materials and equipment in indistinct form, for the monitoring of gamma radiation, beta particles, different qualities of x-ray radiation, and sometimes neutrons. The monitoring of the exposed personnel at the diverse sources of ionizing radiation mentioned is carried out in many occasions without having with the materials (detectors), neither the appropriate infrastructure and therefore without the quality control that guarantees a correct evaluation of the dose equivalent, as a result of the exposure to the ionizing radiations; it

Summary Report of the 1. Research Coordination Meeting on Testing and Improving the International Reactor Dosimetry and Fusion File (IRDFF)

International Nuclear Information System (INIS)

Trkov, A.; Greenwood, L.R.; Simakov, S.P.

2013-09-01

In accordance with the recommendations of the International Nuclear Data Committee in May 2012, the Nuclear Data Section of IAEA has initiated a new Coordinated Research Project (CRP number F41031) with the main goal to test, validate and improve the international dosimetry library for fission and fusion (IRDFF). The output of this CRP will be a reference dosimetry database of cross sections and decay data with corresponding documentation. It will serve to the needs of fission, fusion and accelerator applications. The first Research Coordination Meeting (RCM) was held 1 to 5 July 2013 in IAEA. At this meeting, the attendees discussed the objectives of the whole CRP, presented their contributions and elaborated on consolidated recommendations and actions for implementation over the next 1.5 year period. This Summary Report documents the individual contributions and joint decisions made during this meeting. (author)

Production of neutron cross section library based on JENDL-4.0 to continuous-energy Monte Carlo code MVP and its application to criticality analysis of benchmark problems in the ICSBEP handbook

International Nuclear Information System (INIS)

Okumura, Keisuke; Nagaya, Yasunobu

2011-09-01

In May 2010, JENDL-4.0 was released from Japan Atomic Energy Agency as the updated Japanese Nuclear Data Library. It was processed by the nuclear data processing system LICEM and an arbitrary-temperature neutron cross section library MVPlib - nJ40 was produced for the neutron and photon transport calculation code MVP based on the continuous-energy Monte Carlo method. The library contains neutron cross sections for 406 nuclides on the free gas model, thermal scattering cross sections, and cross sections of pseudo fission products for burn-up calculations with MVP. Criticality benchmark calculations were carried out with MVP and MVPlib - nJ40 for about 1,000 cases of critical experiments stored in the hand book of International Criticality Safety Benchmark Evaluation Project (ICSBEP), which covers a wide variety of fuel materials, fuel forms, and neutron spectra. We report all comparison results (C/E values) of effective neutron multiplication factors between calculations and experiments to give a validation data for the prediction accuracy of JENDL-4.0 for criticalities. (author)

Report on external occupational dosimetry in Canada

International Nuclear Information System (INIS)

1995-12-01

In light of the new recommendations of the ICRP in Report 60 on dose quantities and dose limits, this working group was set up to examine the implications for external dosimetry in Canada. The operational quantities proposed by the ICRU are discussed in detail with regard to their applicability in Canada. The current occupational dosimetry services available in Canada are described as well as the several performance intercomparisons that have been carried out within the country as well as internationally. Recommendations are given with respect to standards for dosimetry, including accuracy and precision. More practical advice is given on the choice of dosimeter to use for external dosimetry, frequency of monitoring, and who should be monitored. Specific advice is given on the monitoring of pregnant workers and problem of non-uniform irradiation. Accident and emergency dosimetry are dealt with briefly. Suggestions are given regarding record keeping both for employers and for the national dose registry. 48 refs., 6 tabs., 1 fig

Activity report of Japanese Nuclear Data Committee in period of April 1993 to March 1995

International Nuclear Information System (INIS)

1995-01-01

This is a progress report on the activities of the Japanese Nuclear Data Committee (JNDC) in the period of April 1993 to March 1995. In this period, the latest version of the Japanese Evaluated Nuclear Data Library, JENDL-3.2, has been released for users after benchmark tests of data therein. Preparation of the special purpose nuclear data files has been also progressed. As special topics, the followings are presented here : (1) the completion and benchmark tests of the JENDL-3.2, (2) progress in evaluating special purpose nuclear data, (3) progress in studying high energy nuclear data, and (4) utilization of nuclear data by means of internet. (author)

Development of automatic editing system for MCNP library 'autonj'

International Nuclear Information System (INIS)

Maekawa, Fujio; Sakurai, Kiyoshi; Kume, Etsuo; Nomura, Yasushi; Kosako, Kazuaki; Kawasaki, Nobuo; Naito, Yoshitaka

1999-12-01

As an activity of the MCNP High-Temperature Library Production Working Group under the Nuclear Code Evaluation Special Committee of Nuclear Code Committee, the automatic editing system for MCNP library 'autonj' was developed. The autonj includes the NJOY-97 code as its main body, and is a system that enables us to easily produce cross section libraries for MCNP from evaluated nuclear data files such as JENDL-3.2. A temperature dependent library at six temperature points based on JENDL-3.2 was produced by using autonj. The autonj system and the temperature dependent library were installed on the JAERI AP3000 computer. (author)

ESR/tooth enamel dosimetry application to Chernobyl case: individual retrospective dosimetry of the liquidators and wild animals

International Nuclear Information System (INIS)

Bugai, A.; Baryakchtar, V.G.; Baran, N.

1996-01-01

ESR/tooth enamel dosimetry technique was used for individual retrospective dosimetry of the servicemen who had worked in 1986-1987 at the liquidation of consequences of the Chernobyl accident. For 18 investigated cases, the values varied from 0,10 (sensitivity limit) to 1,75 Gy. The same technique was used for individual dosimetry of wild animals boars, red deers, elks) hunted at contaminated 30-km area around the Chernobyl Power Plant. Measured values varied from 0,20 to 5,0 Gy/year and were compared with calculated for external and internal irradiation

Dosimetry methods for fuels, cladding and structural materials

International Nuclear Information System (INIS)

Roettger, H.

1980-01-01

This volume of the proceedings of the symposium on reactor dosimetry covers the following topics: the metallurgy and dosimetry interface, radiation damage correlations of structural materials and damage analyses techniques, dosimetry for fusion materials, light water reactor pressure vessel surveillance in practice and irradiation experiments, fast reactor and reseach reactor characterization

Glucinium dosimetry in beryl

International Nuclear Information System (INIS)

Kremer, M.

1949-05-01

The application of the method developed by Kolthoff and Sandell (1928) for the dosimetry of glucinium (beryllium) in beryl gives non-reproducible results with up to 20% discrepancies. This method recommends to separate beryllium and aluminium using 8 hydroxyquinoline and then to directly precipitate glucinium in the filtrate using ammonia. One possible reason of the problems generated by this method should be the formation of a volatile complex between beryllium and the oxine. This work shows that when the oxine is eliminated before the precipitation with ammonia the dosimetry of beryllium becomes accurate. The destruction of the oxine requires the dry evaporation of the filtrate, which is a long process. Thus the search for a reagent allowing the quantitative precipitation of beryllium in its solutions and in presence of oxine has been made. It has been verified also that the quantitative precipitation of the double beryllium and ammonium phosphate is not disturbed by the oxine in acetic buffer. This method, which gives good results, has also the advantage to separate beryllium from the alkaline-earth compounds still present in the filtrate. The report details the operation mode of the method: beryllium dosimetry using ammonium phosphate, aluminium-beryllium separation, application to beryl dosimetry (ore processing, insolubilization of silica, precipitation with ammonia, precipitation with oxine, precipitation of PO 4 NH 4 Gl, preciseness). (J.S.)

Proceedings of the 5. symposium on neutron dosimetry. Beam dosimetry

International Nuclear Information System (INIS)

Schraube, H.; Burger, G.; Booz, J.

1985-01-01

Proceedings of the fifth symposium on neutron dosimetry, organized at Neuherberg, 17-21 September 1984, by the Commission of the European Communities and the GSF Neuherberg, with the co-sponsorship of the US Department of Energy, Office of Health and Environmental Research. The proceedings deal with research on concepts, instruments and methods in radiological protection for neutrons and mixed neutron-gamma fields, including the generation, collection and evaluation of new dosimetric data, the derivation of relevant radiation protection quantitites, and the harmonization of experimental methods and instrumentation by intercomparison programmes. Besides radiation protection monitoring, the proceedings also report on the improvement of neutron beam dosimetry in the fields of radiobiology and radiation therapy

Standardized physics-dosimetry for US pressure vessel cavity surveillance programs

International Nuclear Information System (INIS)

Ruddy, F.H.; McElroy, W.N.; Lippincott, E.P.

1984-01-01

This paper summarizes the applications of ASTM standard methods, guides and practices to define the selection and deployment of recommended dosimetry sets, the selection of dosimetry capsules and thermal neutron shields, the placement of dosimetry, the methods of measurement of dosimetry sensor reaction products, data analysis procedures, and uncertainty evaluation procedures. It also describes the validation of these standards both by in-reactor testing of advanced PV cavity surveillance physics-dosimetry and by data development. The use of these standards to guide selection and development of advanced dosimetry sets for commercial reactors is also summarized. (Auth.)

Cross sections required for FMIT dosimetry

International Nuclear Information System (INIS)

Gold, R.; McElroy, W.N.; Lippincott, E.P.; Mann, F.M.; Oberg, D.L.; Roberts, J.H.; Ruddy, F.H.

1980-01-01

The Fusion Materials Irradiation Test (FMIT) facility, currently under construction, is designed to produce a high flux of high energy neutrons for irradiation effects experiments on fusion reactor materials. Characterization of the flux-fluence-spectrum in this rapidly varying neutron field requires adaptation and extension of currently available dosimetry techniques. This characterization will be carried out by a combination of active, passive, and calculational dosimetry. The goal is to provide the experimenter with accurate neutron flux-fluence-spectra at all positions in the test cell. Plans have been completed for a number of experimental dosimetry stations and provision for these facilities has been incorporated into the FMIT design. Overall needs of the FMIT irradiation damage program delineate goal accuracies for dosimetry that, in turn, create new requirements for high energy neutron cross section data. Recommendations based on these needs have been derived for required cross section data and accuracies

Sensitivity analysis of characteristics of spent mixed oxide fuel

International Nuclear Information System (INIS)

Hagura, Naoto; Yoshida, Tadashi

2008-01-01

Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 σ basis. (authors)

Sensitivity analysis of characteristics of spent mixed oxide fuel

Energy Technology Data Exchange (ETDEWEB)

Hagura, Naoto; Yoshida, Tadashi [Musashi Institute of Technology, 1-28-1 Tamazutsumi, Setagaya, Tokyo 158-8557 (Japan)

2008-07-01

Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 sigma basis. (authors)

Special workshop on lung dosimetry

International Nuclear Information System (INIS)

Fisher, D.R.

1983-01-01

A Special Workshop on Lung Dosimetry was convened in Salt Lake City, Utah, on April 21-22, 1982, to stimulate the use of improved radiation dosimetry and to formulate a stronger basis for dose-response relationships for inhaled radionuclides. The two-day workshop was held in conjunction with the 30th Annual Meeting of the Radiation Research Society. Publication is planned

Dosimetry of internal emitters

International Nuclear Information System (INIS)

Anon.

1982-01-01

The Dosimetry of Internal Emitter Program endeavors to refine the correlation between radiation dose and observed biological effects. The program is presently engaged in the development of studies that will demonstrate the applicability of microdosimetry models developed under the Microdosimetry of Internal Sources Program. The program also provides guidance and assistance to Pacific Northwest Laboratory's Biology Department in the dosimetric analysis of internally deposited radionuclides. This report deals with alpha particle dosimetry plutonium 239 inhalation, and in vitro studies of chromosomal observations

Nuclear medicine radiation dosimetry

CERN Document Server

McParland, Brian J

2010-01-01

Complexities of the requirements for accurate radiation dosimetry evaluation in both diagnostic and therapeutic nuclear medicine (including PET) have grown over the past decade. This is due primarily to four factors: growing consideration of accurate patient-specific treatment planning for radionuclide therapy as a means of improving the therapeutic benefit, development of more realistic anthropomorphic phantoms and their use in estimating radiation transport and dosimetry in patients, design and use of advanced Monte Carlo algorithms in calculating the above-mentioned radiation transport and

Information from the Dosimetry Service

CERN Multimedia

2006-01-01

Please note the following opening hours of the Service: In June: Every morning from 8:30 to 12:00 In July: Mondays, Wednesdays and Fridays from 8:30 to 11:30 Closed all day on Tuesdays and Thursdays From 31st July onwards: Every morning from 8:30 to 12:00 The Service is closed in the afternoons. We should like to remind you that dosimeters cannot be sent to customers by internal mail. Short-term dosimeters (VCTs) must always be returned to the Service after use and must not be left on the racks in the experimental areas or in the secretariats. Dosimetry Service Tel 72155 Bldg. 24 E 011 Dosimetry.service@cern.ch http://cern.ch/rp-dosimetry

INDIVIDUAL DOSIMETRY SERVICE

CERN Multimedia

2000-01-01

Personnel in the distribution groups Aleph, Delphi, L3, Opal who also work for other experiments than at LEP, should contact their dispatchers to explain their activities for the future, after LEP dismantling in order to be maintained on the regular distribution list at Individual DosimetryWe inform all staff and users under regular dosimetric control that the dosimeters for the monitoring period MAY/JUNE will be available from their usual dispatchers on Tuesday 2 May.Please have your films changed before the 12 May.The colour of the dosimeter valid in is MAY/JUNE is YELLOW.Individual Dosimetry Service will be closed on Friday 28 April.

Dosimetry system of the RB reactor

International Nuclear Information System (INIS)

Lolic, B.; Vukadin, D.

1962-01-01

Although RB reactor is operated at very low power levels, safety and dosimetry systems have high importance. This paper shows detailed dosimetry system with fundamental typical components. Estimated radiation doses dependent on reactor power are given at some characteristic points in the rooms nearby reactor

Experimental verification of internal dosimetry calculations. Annual progress report

International Nuclear Information System (INIS)

1980-05-01

During the past year a dosimetry research program has been established in the School of Nuclear Engineering at the Georgia Institute of Technology. The major objective of this program has been to provide research results upon which a useful internal dosimetry system could be based. The important application of this dosimetry system will be the experimental verification of internal dosimetry calculations such as those published by the MIRD Committee

Dosimetry: an ARDENT topic

CERN Multimedia

CERN Bulletin

2012-01-01

The first annual ARDENT workshop took place in Vienna from 20 to 23 November. The workshop gathered together the Early-Stage Researchers (ESR) and their supervisors, plus other people involved from all the participating institutions.   “The meeting, which was organised with the local support of the Austrian Institute of Technology, was a nice opportunity for the ESRs to get together, meet each other, and present their research plans and some preliminary results of their work,” says Marco Silari, a member of CERN Radiation Protection Group and the scientist in charge of the programme. Two full days were devoted to a training course on radiation dosimetry, delivered by renowned experts. The workshop closed with a half-day visit to the MedAustron facility in Wiener Neustadt. ARDENT (Advanced Radiation Dosimetry European Network Training) is a Marie Curie ITN project funded under EU FP7 with €4 million. The project focuses on radiation dosimetry exploiting se...

Neutron dosimetry in biology

International Nuclear Information System (INIS)

Sigurbjoernsson, B.; Smith, H.H.; Gustafsson, A.

1965-01-01

To study adequately the biological effects of different energy neutrons it is necessary to have high-intensity sources which are not contaminated by other radiations, the most serious of which are gamma rays. An effective dosimetry must provide an accurate measure of the absorbed dose, in biological materials, of each type of radiation at any reactor facility involved in radiobiological research. A standardized biological dosimetry, in addition to physical and chemical methods, may be desirable. The ideal data needed to achieve a fully documented dosimetry has been compiled by H. Glubrecht: (1) Energy spectrum and intensity of neutrons; (2) Angular distribution of neutrons on the whole surface of the irradiated object; (3) Additional undesired radiation accompanying the neutrons; (4) Physical state and chemical composition of the irradiated object. It is not sufficient to note only an integral dose value (e.g. in 'rad') as the biological effect depends on the above data

Hanford External Dosimetry Technical Basis Manual PNL-MA-842

International Nuclear Information System (INIS)

Rathbone, Bruce A.

2006-01-01

The Hanford External Dosimetry Technical Basis Manual PNL-MA-842 documents the design and implementation of the external dosimetry system used at Hanford. The manual describes the dosimeter design, processing protocols, dose calculation methodology, radiation fields encountered, dosimeter response characteristics, limitations of dosimeter design under field conditions, and makes recommendations for effective use of the dosimeters in the field. The manual describes the technical basis for the dosimetry system in a manner intended to help ensure defensibility of the dose of record at Hanford and to demonstrate compliance with 10 CFR 835, DOELAP, DOE-RL, ORP, PNSO, and Hanford contractor requirements. The dosimetry system is operated by PNNL's Hanford External Dosimetry Program which provides dosimetry services to all Hanford contractors. The primary users of this manual are DOE and DOE contractors at Hanford using the dosimetry services of PNNL. Development and maintenance of this manual is funded directly by DOE and DOE contractors. Its contents have been reviewed and approved by DOE and DOE contractors at Hanford through the Hanford Personnel Dosimetry Advisory Committee which is chartered and chaired by DOE-RL and serves as means of coordinating dosimetry practices across contractors at Hanford. This manual was established in 1996. Since inception, it has been revised many times and maintained by PNNL as a controlled document with controlled distribution. Rev. 0 marks the first revision to be released through PNNL's Electronic Records & Information Capture Architecture (ERICA) database

Hanford External Dosimetry Technical Basis Manual PNL-MA-842

Energy Technology Data Exchange (ETDEWEB)

Rathbone, Bruce A.

2005-02-25

The Hanford External Dosimetry Technical Basis Manual PNL-MA-842 documents the design and implementation of the external dosimetry system used at Hanford. The manual describes the dosimeter design, processing protocols, dose calculation methodology, radiation fields encountered, dosimeter response characteristics, limitations of dosimeter design under field conditions, and makes recommendations for effective use of the dosimeters in the field. The manual describes the technical basis for the dosimetry system in a manner intended to help ensure defensibility of the dose of record at Hanford and to demonstrate compliance with 10 CFR 835, DOELAP, DOE-RL, ORP, PNSO, and Hanford contractor requirements. The dosimetry system is operated by PNNL’s Hanford External Dosimetry Program which provides dosimetry services to all Hanford contractors. The primary users of this manual are DOE and DOE contractors at Hanford using the dosimetry services of PNNL. Development and maintenance of this manual is funded directly by DOE and DOE contractors. Its contents have been reviewed and approved by DOE and DOE contractors at Hanford through the Hanford Personnel Dosimetry Advisory Committee which is chartered and chaired by DOE-RL and serves as means of coordinating dosimetry practices across contractors at Hanford. This manual was established in 1996. Since inception, it has been revised many times and maintained by PNNL as a controlled document with controlled distribution. Rev. 0 marks the first revision to be released through PNNL’s Electronic Records & Information Capture Architecture (ERICA) database.

Dosimetry techniques applied to thermoluminescent age estimation

International Nuclear Information System (INIS)

Erramli, H.

1986-12-01

The reliability and the ease of the field application of the measuring techniques of natural radioactivity dosimetry are studied. The natural radioactivity in minerals in composed of the internal dose deposited by alpha and beta radiations issued from the sample itself and the external dose deposited by gamma and cosmic radiations issued from the surroundings of the sample. Two technics for external dosimetry are examined in details. TL Dosimetry and field gamma dosimetry. Calibration and experimental conditions are presented. A new integrated dosimetric method for internal and external dose measure is proposed: the TL dosimeter is placed in the soil in exactly the same conditions as the sample ones, during a time long enough for the total dose evaluation [fr

Report on high energy neutron dosimetry workshop

International Nuclear Information System (INIS)

Alvar, K.R.; Gavron, A.

1993-01-01

The workshop was called to assess the performance of neutron dosimetry per the responses from ten DOE accelerator facilities to an Office of Energy Research questionnaire regarding implementation of a personnel dosimetry requirement in DRAFT DOE 5480.ACC, ''Safety of Accelerator Facilities''. The goals of the workshop were to assess the state of dosimetry at high energy accelerators and if such dosimetry requires improvement, to reach consensus on how to proceed with such improvements. There were 22 attendees, from DOE Programs and contract facilities, DOE, Office of Energy Research (ER), Office of Environmental Safety and Health (EH), Office of Fusion Energy, and the DOE high energy accelerator facilities. A list of attendees and the meeting agenda are attached. Copies of the presentations are also attached

Differences between cross-section libraries for neutron dosimetry

International Nuclear Information System (INIS)

Tardelli, T.C.; Stecher, L.C.; Coelho, T.S.; Castro, V.A. De; Cavalieri, T.A.; Menzel, F.; Giarola, R.S.; Domingos, D.B.; Yoriyaz, H.

2013-01-01

Absorbed dose calculations depend on a consistent set of nuclear data used in simulations in computer codes. Nuclear data are stored in libraries, however, the information available about the differences in dose caused by different libraries are rare. The libraries are processed by a computer system to be able to be used by a radiation transport code. One of the systems capable of processing nuclear data is the NJOY system. The objective of this study is to evaluate the nuclear data libraries for neutrons available in the literature, and to quantify the differences in absorbed dose obtained using the libraries JENDL 4.0, JEFF 3.3.1 and ENDF/B.VII. The absorbed dose calculation was performed on a simple geometric model, as spheres, and in anthropomorphic model of the human body based on the ICRP-110 for neutron transport simulation using the MCNP5 code. The results were compared with literature data. The results obtained with cross sections from the libraries JEFF and ENDF/B.VII have shown to be identical in most cases, except for one case where the difference has exceeded 10%. The results obtained with JENDL library has shown to be considerably different in most cases comparing to other two libraries. Some differences were over 200%. The dose calculations showed differences between the libraries, which is justified by differences in the cross sections. It has been observed that the cross sections values of certain nuclides assume quite different values in different libraries. These differences in turn cause considerable differences in dose calculations. (author)

Results from 2010 Caliban Criticality Dosimetry Intercomparison

Energy Technology Data Exchange (ETDEWEB)

Veinot, K. G.

2011-10-12

The external dosimetry program participated in a criticality dosimetry intercomparison conducted at the Caliban facility in Valduc, France in 2010. Representatives from the dosimetry and instrumentation groups were present during testing which included irradiations of whole-body beta/gamma (HBGT) and neutron thermoluminescent dosimeters (TLDs), a fixed nuclear accident dosimeter (FNAD), electronic alarming dosimeters, and a humanoid phantom filled with reference man concentrations of sodium. This report reviews the testing procedures, preparations, irradiations, and presents results of the tests.

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

Technical Basis Document for PFP Area Monitoring Dosimetry Program

CERN Document Server

Cooper, J R

2000-01-01

This document describes the phantom dosimetry used for the PFP Area Monitoring program and establishes the basis for the Plutonium Finishing Plant's (PFP) area monitoring dosimetry program in accordance with the following requirements: Title 10, Code of Federal Regulations (CFR), part 835, ''Occupational Radiation Protection'' Part 835.403; Hanford Site Radiological Control Manual (HSRCM-1), Part 514; HNF-PRO-382, Area Dosimetry Program; and PNL-MA-842, Hanford External Dosimetry Technical Basis Manual.

Technical Basis Document for PFP Area Monitoring Dosimetry Program

International Nuclear Information System (INIS)

COOPER, J.R.

2000-01-01

This document describes the phantom dosimetry used for the PFP Area Monitoring program and establishes the basis for the Plutonium Finishing Plant's (PFP) area monitoring dosimetry program in accordance with the following requirements: Title 10, Code of Federal Regulations (CFR), part 835, ''Occupational Radiation Protection'' Part 835.403; Hanford Site Radiological Control Manual (HSRCM-1), Part 514; HNF-PRO-382, Area Dosimetry Program; and PNL-MA-842, Hanford External Dosimetry Technical Basis Manual

Hanford internal dosimetry program manual

International Nuclear Information System (INIS)

Carbaugh, E.H.; Sula, M.J.; Bihl, D.E.; Aldridge, T.L.

1989-10-01

This document describes the Hanford Internal Dosimetry program. Program Services include administrating the bioassay monitoring program, evaluating and documenting assessments of internal exposure and dose, ensuring that analytical laboratories conform to requirements, selecting and applying appropriate models and procedures for evaluating internal radionuclide deposition and the resulting dose, and technically guiding and supporting Hanford contractors in matters regarding internal dosimetry. 13 refs., 16 figs., 42 tabs

Uncertainty analysis of dosimetry spectrum unfolding

International Nuclear Information System (INIS)

Perey, F.G.

1977-01-01

The propagation of uncertainties in the input data is analyzed for the usual dosimetry unfolding solution. A new formulation of the dosimetry unfolding problem is proposed in which the most likely value of the spectrum is obtained. The relationship of this solution to the usual one is discussed

GENII: The Hanford Environmental Radiation Dosimetry Software System: Volume 2, Users' manual: Hanford Environmental Dosimetry Upgrade Project

International Nuclear Information System (INIS)

Napier, B.A.; Peloquin, R.A.; Strenge, D.L.; Ramsdell, J.V.

1988-11-01

The Hanford Environmental Dosimetry Upgrade Project was undertaken to incorporate the internal dosimetry models recommended by the International Commission on Radiological Protection (ICRP) in updated versions of the environmental pathway analysis models used at Hanford. The resulting second generation of Hanford environmental dosimetry computer codes is compiled in the Hanford Environmental Dosimetry System (Generation II, or GENII). The purpose of this coupled system of computer codes is to analyze environmental contamination of, air, water, or soil. This is accomplished by calculating radiation doses to individuals or populations. GENII is described in three volumes of documentation. This second volume is a Users' Manual, providing code structure, users' instructions, required system configurations, and QA-related topics. The first volume describes the theoretical considerations of the system. The third volume is a Code Maintenance Manual for the user who requires knowledge of code detail. It includes logic diagrams, global dictionary, worksheets, example hand calculations, and listings of the code and its associated data libraries. 27 refs., 17 figs., 23 tabs

Technical Basis Document for PFP Area Monitoring Dosimetry Program

Energy Technology Data Exchange (ETDEWEB)

COOPER, J.R.

2000-04-17

This document describes the phantom dosimetry used for the PFP Area Monitoring program and establishes the basis for the Plutonium Finishing Plant's (PFP) area monitoring dosimetry program in accordance with the following requirements: Title 10, Code of Federal Regulations (CFR), part 835, ''Occupational Radiation Protection'' Part 835.403; Hanford Site Radiological Control Manual (HSRCM-1), Part 514; HNF-PRO-382, Area Dosimetry Program; and PNL-MA-842, Hanford External Dosimetry Technical Basis Manual.

Introduction to radiological physics and radiation dosimetry

CERN Document Server

Attix, Frank Herbert

2004-01-01

A straightforward presentation of the broad concepts underlying radiological physics and radiation dosimetry for the graduate-level student. Covers photon and neutron attenuation, radiation and charged particle equilibrium, interactions of photons and charged particles with matter, radiotherapy dosimetry, as well as photographic, calorimetric, chemical, and thermoluminescence dosimetry. Includes many new derivations, such as Kramers X-ray spectrum, as well as topics that have not been thoroughly analyzed in other texts, such as broad-beam attenuation and geometrics, and the reciprocity theorem

Sixth international radiopharmaceutical dosimetry symposium: Proceedings. Volume 2

Energy Technology Data Exchange (ETDEWEB)

S.-Stelson, A.T. [ed.] [comp.; Stabin, M.G.; Sparks, R.B. [eds.; Smith, F.B. [comp.

1999-01-01

This conference was held May 7--10 in Gatlinburg, Tennessee. The purpose of this conference was to provide a multidisciplinary forum for exchange of state-of-the-art information on radiopharmaceutical dosimetry. Attention is focused on the following: quantitative analysis and treatment planning; cellular and small-scale dosimetry; dosimetric models; radiopharmaceutical kinetics and dosimetry; and animal models, extrapolation, and uncertainty.

Reactor dosimetry integral reaction rate data in LMFBR Benchmark and standard neutron fields: status, accuracy and implications

International Nuclear Information System (INIS)

Fabry, A.; Ceulemans, H.; Vandeplas, P.; McElroy, W.N.; Lippincott, E.P.

1977-01-01

This paper provides conclusions that may be drawn regarding the consistency and accuracy of dosimetry cross-section files on the basis of integral reaction rate data measured in U.S. and European benchmark and standard neutron fields. In a discussion of the major experimental facilities CFRMF (Idaho Falls), BIGTEN (Los Alamos), ΣΣ (Mol, Bucharest), NISUS (London), TAPIRO (Roma), FISSION SPECTRA (NBS, Mol, PTB), attention is paid to quantifying the sensitivity of computed integral data relative to the presently evaluated accuracy of the various neutron spectral distributions. The status of available integral data is reviewed and the assigned uncertainties are appraised, including experience gained by interlaboratory comparisons. For all reactions studied and for the various neutron fields, the measured integral data are compared to the ones computed from the ENDF/B-IV and the SAND-II dosimetry cross-section libraries as well as to some other differential data in relevant cases. This comparison, together with the proposed sensitivity and accuracy assessments, is used, whenever possible, to establish how well the best cross-sections evaluated on the basis of differential measurements (category I dosimetry reactions) are reliable in terms of integral reaction rates prediction and, for those reactions for which discrepancies are indicated, in which energy range it is presumed that additional differential measurements might help. For the other reactions (category II), the inconsistencies and trends are examined. The need for further integral measurements and interlaboratory comparisons is also considered

Proceedings of the recent developments in radiation dosimetry

International Nuclear Information System (INIS)

Bhat, Nagesh; Palani Selvan, T.

2016-01-01

Whilst 'Dosimetry' in its original sense deals with methods for a quantitative determination of energy deposited in a given medium by directly or indirectly ionizing radiations, the term is better known as a scientific sub-specialty in the fields of health physics and medical physics, where it is the calculation and assessment of the radiation dose received by the human body. Dosimetry is used extensively for radiation protection and is routinely applied to ensure radiological safety of occupational radiation workers. Internal dosimetry due to the ingestion or inhalation of radioactive materials relies on a variety of physiological or imaging techniques. External dosimetry, due to irradiation from an external source is based on measurements with a dosimeter, or inferred from other radiological protection instruments. Radiation dosimetry is one of the important research areas of Department of Atomic Energy (DAE). This research work is centered on the facilities such as nuclear reactors, reprocessing plants, high energy accelerators (research/industry/medical), radiation standards, food processing, radiation technology development, etc. In each of these facilities, radiation field environment is different and the associated dosimetry concepts are different. Papers relevant to INIS are indexed separately

Quality control through dosimetry at a contract radiation processing facility

International Nuclear Information System (INIS)

Du Plessis, T.A.; Roediger, A.H.A.

1985-01-01

Reliable dosimetry procedures constitute a very important part of process control and quality assurance at a contract gamma radiation processing facility that caters for a large variety of different radiation applications. The choice, calibration and routine intercalibration of the dosimetry systems employed form the basis of a sound dosimetry policy in radiation processing. With the dosimetric procedures established, detailed dosimetric mapping of the irradiator upon commissioning (and whenever source modifications take place) is carried out to determine the radiation processing characteristics and peformance of the plant. Having established the irradiator parameters, routine dosimetry procedures, being part of the overall quality control measures, are employed. In addition to routine dosimetry, independent monitoring of routine dosimetry is performed on a bi-monthly basis and the results indicate a variation of better than 3%. On an annaul basis the dosimetry systems are intercalibrated through at least one primary standard dosimetry laboratory and to date a variation of better than 5% has been experienced. The company also participates in the Pilot Dose Assurance Service of the International Atomic Energy Agency, using the alanine/ESR dosimetry system. Routine calibration of the instrumentation employed is carried out on a regular basis. Detailed permanent records are compiled on all dosimetric and instrumentation calibrations, and the routine dosimetry employed at the plant. Certificates indicating the measured absorbed radiation doses are issued on request and in many cases are used for the dosimetric release of sterilized medical and pharmaceutical products. These procedures, used by Iso-Ster at its industrial gamma radiation facility, as well as the experience built up over a number of years using radiation dosimetry for process control and quality assurance are discussed. (author)

User's manual of a supporting system for treatment planning in boron neutron capture therapy. JAERI computational dosimetry system

CERN Document Server

Kumada, H

2002-01-01

A boron neutron capture therapy (BNCT) with epithermal neutron beam is expected to treat effectively for malignant tumor that is located deeply in the brain. It is indispensable to estimate preliminarily the irradiation dose in the brain of a patient in order to perform the epithermal neutron beam BNCT. Thus, the JAERI Computational Dosimetry System (JCDS), which can calculate the dose distributions in the brain, has been developed. JCDS is a software that creates a 3-dimensional head model of a patient by using CT and MRI images and that generates a input data file automatically for calculation neutron flux and gamma-ray dose distribution in the brain by the Monte Carlo code: MCNP, and that displays the dose distribution on the head model for dosimetry by using the MCNP calculation results. JCDS has any advantages as follows; By treating CT data and MRI data which are medical images, a detail three-dimensional model of patient's head is able to be made easily. The three-dimensional head image is editable to ...

Glass badge dosimetry system for large scale personal monitoring

International Nuclear Information System (INIS)

Norimichi Juto

2002-01-01

Glass Badge using silver activated phosphate glass dosemeter was specially developed for large scale personal monitoring. And dosimetry systems such as an automatic leader and a dose equipment calculation algorithm were developed at once to achieve reasonable personal monitoring. In large scale personal monitoring, both of precision for dosimetry and confidence for lot of personal data handling become very important. The silver activated phosphate glass dosemeter has basically excellent characteristics for dosimetry such as homogeneous and stable sensitivity, negligible fading and so on. Glass Badge was designed to measure 10 keV - 10 MeV range of photon. 300 keV - 3 MeV range of beta, and 0.025 eV - 15 MeV range of neutron by included SSNTD. And developed Glass Badge dosimetry system has not only these basic characteristics but also lot of features to keep good precision for dosimetry and data handling. In this presentation, features of Glass Badge dosimetry systems and examples for practical personal monitoring systems will be presented. (Author)

A-bomb survivor dosimetry update

International Nuclear Information System (INIS)

Loewe, W.E.

1982-06-01

A-bomb survivor data have been generally accepted as applicable. Also, the initial radiations have tended to be accepted as the dominant radiation source for all survivors. There was general acceptance of the essential reliability of both the biological effects data and the causative radiation dose values. There are considerations casting doubt on these acceptances, but very little quantification of th implied uncertainties has been attempted. The exception was A-bomb survivor dosimetry, where free-field kerma values for initial radiations were thought to be accurate to about 30%, and doses to individual survivors were treated as effectively error-free. In 1980, a major challenge to the accepted A-bomb survivor dosimetry was announced, and was quickly followed by a succession of explanations and displays showing the soundness of that challenge. In fact, a complete replacement set of free-field kerma values was provided which was suitable for use in constructing an entire new dosimetry for Hiroshima and Nagasaki. The new values showed many changes greater than the accepted 30% uncertainty. An approximate new dosimetry was indeed constructed, and used to convert existing leukemia cause-and-effect data from the old to the new dose values, by way of assessing the impact

Standardized physics-dosimetry for US pressure vessel cavity surveillance programs

International Nuclear Information System (INIS)

Ruddy, F.H.; McElroy, W.N.; Lippincott, E.P.

1984-01-01

Standardized Physics-Dosimetry procedures and data are being developed and tested for monitoring the neutron doses accumulated by reactor pressure vessels (PV) and their support structures. These procedures and data are governed by a set of 21 ASTM standard practices, guides, and methods for the prediction of neutron-induced changes in light water reactor (LWR) PVs and support structure steels throughout the service life of the PV. This paper summarizes the applications of these standards to define the selection and deployment of recommended dosimetry sets, the selection of dosimetry capsules and thermal neutron shields, the placement of dosimetry, the methods of measurement of dosimetry sensor reaction products, data analysis procedures, and uncertainty evaluation procedures. It also describes the validation of these standards both by in-reactor testing of advanced PV cavity surveillance physics-dosimetry and by data development. The use of these standards to guide selection and deployment of advanced dosimetry sets for commercial reactors is also summarized

Cellular dosimetry in nuclear medicine imaging: training

International Nuclear Information System (INIS)

Gardin, I.; Faraggi, M.; Stievenart, J.L.; Le Guludec, D.; Bok, B.

1998-01-01

The radionuclides used in nuclear medicine imaging emit not only diagnostically useful photons, but also energy electron emissions, responsible for dose heterogeneity at the cellular level. The mean dose delivered to the cell nucleus by electron emissions of 99m Tc, 123 I, 111 In, 67 Ga, and 201 Tl, has been calculated, for the cell nucleus, a cytoplasmic and a cell membrane distribution of radioactivity. This model takes into account both the self-dose which results from the radionuclide located in the target cell, and the cross-dose, which comes from the surrounding cells. The results obtained by cellular dosimetry (D cel ) have been compared with those obtained with conventional dosimetry (D conv ), by assuming the same amount of radioactivity per cell. Cellular dosimetry shows, for a cytoplasmic and a cell membrane distributions of radioactivity, that the main contribution to the dose to the cell nucleus, comes from the surrounding cells. On the other hand, for a cell nucleus distribution of radioactivity, the self-dose is not negligible and may be the main contribution. The comparison between cellular and conventional dosimetry shows that D cel /D conv ratio ranges from 0.61 and O.89, in case of a cytoplasmic and a cell membrane distributions of radioactivity, depending on the radionuclide and cell dimensions. Thus, conventional dosimetry slightly overestimates the mean dose to the cell nucleus. On the other hand, D cel /D conv ranges from 1.1 to 75, in case of a cell nucleus distribution of radioactivity. Conventional dosimetry may strongly underestimates the absorbed dose to the nucleus, when radioactivity is located in the nucleus. The study indicates that in nuclear medicine imaging, cellular dosimetry may lead to a better understanding of biological effects of radiopharmaceuticals. (authors)

Dosimetry in life sciences

International Nuclear Information System (INIS)

1975-01-01

The uses of radiation in medicine and biology have grown in scope and diversity to make the Radiological Sciences a significant factor in both research and medical practice. Of critical importance in the applications and development of biomedical and radiological techniques is the precision with which the dose may be determined at all points of interest in the absorbing medium. This has developed as a result of efficacy of investigations in clinical radiation therapy, concern for patient safety and diagnostic accuracy in diagnostic radiology and the advent of clinical trials and research into the use of heavily ionizing radiations in biology and medicine. Since the last IAEA Symposium on Dosimetry Techniques applied to Agriculture, Industry, Biology and Medicine, held in Vienna in 1972, it has become increasingly clear that advances in the techniques and hardware of biomedical dosimetry have been rapid. It is for these reasons that this symposium was organized in a concerted effort to focus on the problems, developments and areas of further research in dosimetry in the Life Sciences. (author)

Dosimetry in life sciences

Energy Technology Data Exchange (ETDEWEB)

NONE

1975-06-15

The uses of radiation in medicine and biology have grown in scope and diversity to make the Radiological Sciences a significant factor in both research and medical practice. Of critical importance in the applications and development of biomedical and radiological techniques is the precision with which the dose may be determined at all points of interest in the absorbing medium. This has developed as a result of efficacy of investigations in clinical radiation therapy, concern for patient safety and diagnostic accuracy in diagnostic radiology and the advent of clinical trials and research into the use of heavily ionizing radiations in biology and medicine. Since the last IAEA Symposium on Dosimetry Techniques applied to Agriculture, Industry, Biology and Medicine, held in Vienna in 1972, it has become increasingly clear that advances in the techniques and hardware of biomedical dosimetry have been rapid. It is for these reasons that this symposium was organized in a concerted effort to focus on the problems, developments and areas of further research in dosimetry in the Life Sciences. (author)

3D dosimetry by optical-CT scanning

Science.gov (United States)

Oldham, Mark

2006-12-01

The need for an accurate, practical, low-cost 3D dosimetry system is becoming ever more critical as modern dose delivery techniques increase in complexity and sophistication. A recent report from the Radiological Physics Center (RPC) (1), revealed that 38% of institutions failed the head-and-neck IMRT phantom credentialing test at the first attempt. This was despite generous passing criteria (within 7% dose-difference or 4mm distance-to-agreement) evaluated at a half-dozen points and a single axial plane. The question that arises from this disturbing finding is - what percentage of institutions would have failed if a comprehensive 3D measurement had been feasible, rather than measurements restricted to the central film-plane and TLD points? This question can only be adequately answered by a comprehensive 3D-dosimetry system, which presents a compelling argument for its development as a clinically viable low cost dosimetry solution. Optical-CT dosimetry is perhaps the closest system to providing such a comprehensive solution. In this article, we review the origins and recent developments of optical-CT dosimetry systems. The principle focus is on first generation systems known to have highest accuracy but longer scan times.

Thermoluminescent measurement in space radiation dosimetry

International Nuclear Information System (INIS)

Chen Mei; Qi Zhangnian; Li Xianggao; Huang Zengxin; Jia Xianghong; Wang Genliang

1999-01-01

The author introduced the space radiation environment and the application of thermoluminescent measurement in space radiation dosimetry. Space ionization radiation is charged particles radiation. Space radiation dosimetry was developed for protecting astronauts against space radiation. Thermoluminescent measurement is an excellent method used in the spaceship cabin. Also the authors mentioned the recent works here

Fifth international radiopharmaceutical dosimetry symposium

International Nuclear Information System (INIS)

Watson, E.E.; Schlafke-Stelson, A.T.

1992-05-01

This meeting was held to exchange information on how to get better estimates of the radiation absorbed dose. There seems to be a high interest of late in patient dosimetry; discussions were held in the light of revised risk estimates for radiation. Topics included: Strategies of Dose Assessment; Dose Estimation for Radioimmunotherapy; Dose Calculation Techniques and Models; Dose Estimation for Positron Emission Tomography (PET); Kinetics for Dose Estimation; and Small Scale Dosimetry and Microdosimetry. (VC)

Dosimetry on the radiological risks prevention in radiotherapy

International Nuclear Information System (INIS)

Fornet R, O. M.; Perez G, F.

2014-08-01

Dosimetry in its various forms plays a determining role on the radiological risks prevention in radiotherapy. To prove this in this paper is shown an analysis based on the risk matrix method, how the dosimetry can influence in each stages of a radiotherapy service; installation and acceptance, operation, maintenance and calibration. For each one of these stages the role that can play is analyzed as either the initiating event of a radiological accident or limiting barrier of these events of the dosimetric processes used for the individual dosimetry, the area monitoring, fixed or portable, for radiation beam dosimetry and of the patients for a radiotherapy service with cobalt-therapy equipment. The result of the study shows that the application of a prospective approach in the role evaluation of dosimetry in the prevention and mitigation of the consequences of a radiological accident in radiotherapy is crucial and should be subject to permanent evaluation at each development stage of these services. (author)

Dosimetry in support of wholesomeness studies

International Nuclear Information System (INIS)

Jarrett, R.D.; Halliday, J.W.

1979-01-01

Interest in dosimetry procedures in the context of a large-scale processing situation exceeds the purely documentary aspects of this report. The numerous combinations afforded by the various types, strengths and configurations of irradiation sources and the possibilities for various conveyors and other facility design factors impacting on irradiation logistics render a completely general treatment of dosimetry procedures in such instances almost impossible. While the exact combination of these various factors represented by the irradiation facilities at NARADCOM may be duplicated nowhere else, the dosimetry procedures documented in this report offer both experience and solutions that might be more generally useful. Therefore, this report complements and supplements more general discussions found in the literature and cited in the text

Radiation dosimetry activities in the Netherlands

International Nuclear Information System (INIS)

Broerse, J.J.; Mijnheer, B.J.

1986-07-01

The Netherlands Commission for Radiation Dosimetry (NCS) was officially established on 3 September 1982 with the aim of promoting the appropriate use of dosimetry of ionizing radiation both for scientific research and practical applications. The present report provides a compilation of the dosimetry acitivities and expertise available in the Netherlands, based on the replies to a questionnaire mailed under the auspices of the NCS and might suffer from some incompleteness in specific details. The addresses of the Dutch groups with the names of the scientists are given. Individual scientists, not connected with a scientific group, hospital or organization have not been included in this list. Also the names of commercial firms producing dosimetric systems have been omitted. (Auth.)

Mister Voxel: 3D internal dosimetry software for nuclear medicine

International Nuclear Information System (INIS)

McKay, E.

1998-01-01

Full text: Calculation of individual internal dosimetry in nuclear medicine is a complex, multi-stage process. Most often, calculations are biased on the MIRD methodology, which assumes uniform distribution of cumulated activity inside a set of mathematically described internal organs. The MIRD 'reference man' geometry is highly simplified and the dosimetry estimates generated by this method were originally only intended to predict the average dose expected in an exposed population. We have developed a software package for the Macintosh computer ('Mister Voxel') that uses a fast Fourier transform to calculate the 3D distribution of absorbed dose by convolving a 3D dose kernel with a 3D distribution of cumulated activity. This makes it possible to generate dose volume histograms and isodose contours for organs or tumours treated with radiopharmaceuticals, a task not possible using the MIRD technique. In addition to providing 3D convolution, Mister Voxel performs basic image processing functions (image math, filters, cut and paste) and provides a collection of painting tools and simple morphological operators to facilitate the delineation of regions of interest (ROIs) along anatomical boundaries. The package also includes an image registration module with tools for automated or manual registration of 3D data sets. The structure of the package allows ROIs drawn on CT or MRI data to be easily transferred to registered SPECT data. Dose kernels are implemented by plug-in code modules, allowing the user to extend the system's capabilities if required. File import and export capabilities are also extensible

Gamma-ray emission spectra from spheres with 14 MeV neutron source

International Nuclear Information System (INIS)

Yamamoto, Junji; Kanaoka, Takeshi; Murata, Isao; Takahashi, Akito; Sumita, Kenji

1989-01-01

Energy spectra of neutron-induced gamma-rays emitted from spherical samples were measured using a 14 MeV neutron source. The samples in use were LiF, Teflon:(CF 2 ) n , Si, Cr, Mn, Co, Cu, Nb, Mo, W and Pb. A diameter of the sphere was either 40 or 60 cm. The gamma-ray energy in the emission spectra covered the range from 500 keV to 10 MeV. Measured spectra were compared with transport calculations using the nuclear data files of JENDL-3T and ENDF/B-IV. The agreements between the measurements and the JENDL-3T calculations were good in the emission spectra for the low energy gamma-rays from inelastic scattering. (author)

FRENDY: A new nuclear data processing system being developed at JAEA

Directory of Open Access Journals (Sweden)

Tada Kenichi

2017-01-01

Full Text Available JAEA has provided an evaluated nuclear data library JENDL and nuclear application codes such as MARBLE, SRAC, MVP and PHITS. These domestic codes have been widely used in many universities and industrial companies in Japan. However, we sometimes find problems in imported processing systems and need to revise them when the new JENDL is released. To overcome such problems and immediately process the nuclear data when it is released, JAEA started developing a new nuclear data processing system, FRENDY in 2013. This paper describes the outline of the development of FRENDY and both its capabilities and performances by the analyses of criticality experiments. The verification results indicate that FRENDY properly generates ACE files.

FRENDY: A new nuclear data processing system being developed at JAEA

Science.gov (United States)

Tada, Kenichi; Nagaya, Yasunobu; Kunieda, Satoshi; Suyama, Kenya; Fukahori, Tokio

2017-09-01

JAEA has provided an evaluated nuclear data library JENDL and nuclear application codes such as MARBLE, SRAC, MVP and PHITS. These domestic codes have been widely used in many universities and industrial companies in Japan. However, we sometimes find problems in imported processing systems and need to revise them when the new JENDL is released. To overcome such problems and immediately process the nuclear data when it is released, JAEA started developing a new nuclear data processing system, FRENDY in 2013. This paper describes the outline of the development of FRENDY and both its capabilities and performances by the analyses of criticality experiments. The verification results indicate that FRENDY properly generates ACE files.

Fast neutron dosimetry and spectrometry using radioactivation (1963); Dosimetrie et spectrometrie des neutrons rapides par radioactivation (1963)

Energy Technology Data Exchange (ETDEWEB)

Lamberieux, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

The author first recalls rapidly a few generalities concerning induced radioactivity detectors and gives, in an appendix, tables summarizing the properties of detector elements which may be used in radioprotection. The excitation functions found in the literature and also given. The technological characteristics of the detectors used are given, together with the counting methods. The many advantages of activation dosimetry for strong or periodic neutron fluxes and for those in the presence of {gamma}-radiation are stressed. The main problem in activation dosimetry is, however, the calculation of the dose absorbed from the results of the measurements. It is shown how the dose is expressed, fairly accurately, as a function of the radioactivities induced in a series of detectors. As an example, the spectrometry and the dosimetry of the neutron flux emitted by a Po-Be source are presented. (author) [French] L'auteur fait d'abord un bref rappel des generalites sur les detecteurs a radioactivite induite, accompagne, en annexe, des tableaux resumant les proprietes d'elements detecteurs utilisables en radioprotection. Les fonctions d'excitation trouvees dans la litterature y sont egalement annexees. On donne ensuite les caracteristiques technologiques des detecteurs employes ainsi que les methodes de comptage utilisees. On souligne les nombreux avantages de la dosimetrie par activation dans les flux de neutrons intenses ou periodiques et en presence de rayonnement {gamma}. Il reste que le probleme central de la dosimetrie par activation est le calcul de la dose absorbee a partir des resultats de mesure. On montre comment la dose s'exprime, de maniere approchee, en fonction des radioactivites induites dans une serie de detecteurs. A titre d'exemple, la spectrometrie et la dosimetrie du flux de neutrons emis par une source de Po-Be sont presentees. (auteur)

Radiation protection dosimetry in medicine - Report of the working group n.9 of the European radiation dosimetry group (EURADOS) - coordinated network for radiation dosimetry (CONRAD - contract EC N) fp6-12684

International Nuclear Information System (INIS)

2009-01-01

This report present the results achieved within the frame of the work the WP 7 (Radiation Protection Dosimetry of Medical Staff) of the coordination action CONRAD (Coordinated Network for Radiation Dosimetry) funded through the 6. EU Framework Program. This action was coordinated by EURADOS (European Radiation Dosimetry Group). EURADOS is an organization founded in 1981 to advance the scientific understanding and the technical development of the dosimetry of ionising radiation in the fields of radiation protection, radiobiology, radiation therapy and medical diagnosis by promoting collaboration between European laboratories. WP7 coordinates and promotes European research for the assessment of occupational exposures to staff in therapeutic and diagnostic radiology workplaces. Research is coordinated through sub-groups covering three specific areas: 1. Extremity dosimetry in nuclear medicine and interventional radiology: this sub-group coordinates investigations in the specific fields of the hospitals and studies of doses to different parts of the hands, arms, legs and feet; 2. Practice of double dosimetry: this sub-group reviews and evaluates the different methods and algorithms for the use of dosemeters placed above and below lead aprons in large exposure during interventional radiology procedures, especially to determine effective doses to cardiologists during cardiac catheterization; and 3. Use of electronic personal dosemeters in interventional radiology: this sub-group coordinates investigations in laboratories and hospitals, and intercomparisons with passive dosemeters with the aim to enable the formulation of standards. (authors)

Dosimetry of high energy radiation

CERN Document Server

Sahare, P D

2018-01-01

High energy radiation is hazardous to living beings and a threat to mankind. The correct estimation of the high energy radiation is a must and a single technique may not be very successful. The process of estimating the dose (the absorbed energy that could cause damages) is called dosimetry. This book covers the basic technical knowledge in the field of radiation dosimetry. It also makes readers aware of the dangers and hazards of high energy radiation.

Hanford External Dosimetry Technical Basis Manual PNL-MA-842

Energy Technology Data Exchange (ETDEWEB)

Rathbone, Bruce A.

2009-08-28

The Hanford External Dosimetry Technical Basis Manual PNL-MA-842 documents the design and implementation of the external dosimetry system used at Hanford. The manual describes the dosimeter design, processing protocols, dose calculation methodology, radiation fields encountered, dosimeter response characteristics, limitations of dosimeter design under field conditions, and makes recommendations for effective use of the dosimeters in the field. The manual describes the technical basis for the dosimetry system in a manner intended to help ensure defensibility of the dose of record at Hanford and to demonstrate compliance with 10 CFR 835, DOELAP, DOE-RL, ORP, PNSO, and Hanford contractor requirements. The dosimetry system is operated by PNNL’s Hanford External Dosimetry Program (HEDP) which provides dosimetry services to all Hanford contractors. The primary users of this manual are DOE and DOE contractors at Hanford using the dosimetry services of PNNL. Development and maintenance of this manual is funded directly by DOE and DOE contractors. Its contents have been reviewed and approved by DOE and DOE contractors at Hanford through the Hanford Personnel Dosimetry Advisory Committee (HPDAC) which is chartered and chaired by DOE-RL and serves as means of coordinating dosimetry practices across contractors at Hanford. This manual was established in 1996. Since inception, it has been revised many times and maintained by PNNL as a controlled document with controlled distribution. The first revision to be released through PNNL’s Electronic Records & Information Capture Architecture (ERICA) database was designated Revision 0. Revision numbers that are whole numbers reflect major revisions typically involving changes to all chapters in the document. Revision numbers that include a decimal fraction reflect minor revisions, usually restricted to selected chapters or selected pages in the document.

External audit in radiotherapy dosimetry

International Nuclear Information System (INIS)

Thwaites, D.I.; Western General Hospital, Edinburgh

1996-01-01

Quality audit forms an essential part of any comprehensive quality assurance programme. This is true in radiotherapy generally and in specific areas such as radiotherapy dosimetry. Quality audit can independently test the effectiveness of the quality system and in so doing can identify problem areas and minimize their possible consequences. Some general points concerning quality audit applied to radiotherapy are followed by specific discussion of its practical role in radiotherapy dosimetry, following its evolution from dosimetric intercomparison exercises to routine measurement-based on-going audit in the various developing audit networks both in the UK and internationally. Specific examples of methods and results are given from some of these, including the Scottish+ audit group. Quality audit in radiotherapy dosimetry is now well proven and participation by individual centres is strongly recommended. Similar audit approaches are to be encouraged in other areas of the radiotherapy process. (author)

User's manual of a supporting system for treatment planning in boron neutron capture therapy. JAERI computational dosimetry system

International Nuclear Information System (INIS)

Kumada, Hiroaki; Torii, Yoshiya

2002-09-01

A boron neutron capture therapy (BNCT) with epithermal neutron beam is expected to treat effectively for malignant tumor that is located deeply in the brain. It is indispensable to estimate preliminarily the irradiation dose in the brain of a patient in order to perform the epithermal neutron beam BNCT. Thus, the JAERI Computational Dosimetry System (JCDS), which can calculate the dose distributions in the brain, has been developed. JCDS is a software that creates a 3-dimensional head model of a patient by using CT and MRI images and that generates a input data file automatically for calculation neutron flux and gamma-ray dose distribution in the brain by the Monte Carlo code: MCNP, and that displays the dose distribution on the head model for dosimetry by using the MCNP calculation results. JCDS has any advantages as follows; By treating CT data and MRI data which are medical images, a detail three-dimensional model of patient's head is able to be made easily. The three-dimensional head image is editable to simulate the state of a head after its surgical processes such as skin flap opening and bone removal for the BNCT with craniotomy that are being performed in Japan. JCDS can provide information for the Patient Setting System to set the patient in an actual irradiation position swiftly and accurately. This report describes basic design and procedure of dosimetry, operation manual, data and library structure for JCDS (ver.1.0). (author)

Radioactive decay data tables: a handbook of decay data for application to radiation dosimetry and radiological assessments

International Nuclear Information System (INIS)

Kocher, D.C.

1982-01-01

Reviews compendium containing recommended decay data for approx. 500 radionuclides of interest in nuclear medicine and fusion reactor technology or of potential importance in routine or accidental releases from the nuclear fuel cycle. Primary source of the decay data presented in this handbook is the Evaluated Nuclear Structure Data File (ENSDF), developed and maintained by the US Nuclear Data Network. Topics covered include various radioactive decay processes; evaluation process and standards of ENSDF; tables and computer code MEDLIST used to produce ENSDF tables; radiation dosimetry and radiological assessments; parent-daughter activity ratios wherever the adopted decay data may contain significant uncertainties or errors due to the lack of appropriate experimental data

Recent progress in application of JAERI alanine/ESR dosimetry system

International Nuclear Information System (INIS)

Kojima, T.

1995-01-01

Feasibility studies of application of JAERI alanine/ESR dosimetry system were performed on radiotherapy level dosimetry, low dose-rate dosimetry for residual life estimation of cable insulators used in nuclear power facilities, and dose monitoring for electron processing. (author)

10 CFR 835.1304 - Nuclear accident dosimetry.

Science.gov (United States)

2010-01-01

...) Methods and equipment for analysis of biological materials; (3) A system of fixed nuclear accident... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... 10 Energy 4 2010-01-01 2010-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304...

Mixed field dosimetry with paired ionization chambers

International Nuclear Information System (INIS)

Coppola, M.; Porro, F.

1977-01-01

This report describes the results of neutron and gamma mixed-field dosimetry obtained by the Ispra Group in the framework of the European Neutron Dosimetry intercomparison Project (ENDIP). The experimental method and the formulation employed for the derivation of Kerma results are also present

Treating voxel geometries in radiation protection dosimetry with a patched version of the Monte Carlo codes MCNP and MCNPX.

Science.gov (United States)

Burn, K W; Daffara, C; Gualdrini, G; Pierantoni, M; Ferrari, P

2007-01-01

The question of Monte Carlo simulation of radiation transport in voxel geometries is addressed. Patched versions of the MCNP and MCNPX codes are developed aimed at transporting radiation both in the standard geometry mode and in the voxel geometry treatment. The patched code reads an unformatted FORTRAN file derived from DICOM format data and uses special subroutines to handle voxel-to-voxel radiation transport. The various phases of the development of the methodology are discussed together with the new input options. Examples are given of employment of the code in internal and external dosimetry and comparisons with results from other groups are reported.

Dosimetry. Standard practice for dosimetry in gamma irradiation facilities for food and non-food processing

International Nuclear Information System (INIS)

2008-01-01

This Ghana Standard outlines the installation qualification program for an irradiator and the dosimetry procedures to be followed during operational qualification, performance qualification and routine processing in facilities that process food and non-food with gamma rays. This is to ensure that the product has been treated with predetermined range of absorbed dose. It is not intended for use in X-ray and electron beam facilities and therefore dosimetry systems in such facilities are not covered

Alanine-ESR dosimetry for radiotherapy IAEA experience

International Nuclear Information System (INIS)

Mehta, K.; Girzikowsky, R.; )

1997-01-01

At present, the most commonly used transfer dosimeters for radiotherapy applications are TL dosemeters. They are being used for intercomparison between SSDLs (about 70) and the IAEA dosimetry laboratory. However, there are some undesirable characteristics of this dosimetry system. We have a study in progress at the IAEA to evaluate the alanine-ESR systems as an alternative to TLDs. There are several desirable qualities which make alanine an attractive dosemeter. Preliminary data suggest that the alanine-ESR dosimetry system has the potential to replace TLDs for intercomparison amongst SSDLs in the therapy-level dose regions. (Author)

Review of the correlation between results of cytogenetic dosimetry from blood lymphocytes and EPR dosimetry from tooth enamel for victims of radiation accidents

International Nuclear Information System (INIS)

Khvostunov, I.K.; Ivannikov, A.I.; Skvortsov, V.G.; Golub, E.V.; Nugis, V. Yu.

2015-01-01

The goal of this study was to compare dose estimates from electron paramagnetic resonance (EPR) dosimetry with teeth and cytogenetic dosimetry with blood lymphocytes for 30 victims of radiation accidents. The whole-body exposures estimated by tooth enamel EPR dosimetry were ranging from 0.01 to 9.3 Gy. Study group comprised victims exposed to acute and prolonged irradiation at high and low dose rate in different accidents. Blood samples were taken from each of them for cytogenetic analysis. Aberrations were scored and analysed according to International Atomic Energy Agency (IAEA) guidelines for conventional and FISH analysis. Tooth samples were collected in dental clinics after they had been extracted during ordinary practice. EPR dosimetry was performed according to the IAEA protocol. EPR dosimetry showed good correlation with dosimetry based on chromosomal analysis. All estimations of cytogenetic dose below detection limit coincide with EPR dose estimates within the ranges of uncertainty. The differences between cytogenetic and EPR assays may occur in a case of previous unaccounted exposure, non-homogeneous irradiation and due to contribution to absorbed dose from neutron irradiation. (authors)

Image in nuclear dosimetry using thermoluminescent dosimetry

International Nuclear Information System (INIS)

Guinsburg, G.; Matsuoka, M.; Watanabe, S.

1987-01-01

A low cost methodology to produce images of internal sick organs by radioisotopic intake, is presented. Dosimetries of thermoluminescent material and Teflon (ratio:50%) in bidimensional matrix shape are used with a Pb collimator. This collimator-bidimensional matrix system was tested ''in vivo'' and in thyroid phantoms using 99m Tc. A comparative evaluation between this method and the scintigraphy one is presented. (M.A.C.) [pt

The physics of small megavoltage photon beam dosimetry.

Science.gov (United States)

Andreo, Pedro

2018-02-01

The increased interest during recent years in the use of small megavoltage photon beams in advanced radiotherapy techniques has led to the development of dosimetry recommendations by different national and international organizations. Their requirement of data suitable for the different clinical options available, regarding treatment units and dosimetry equipment, has generated a considerable amount of research by the scientific community during the last decade. The multiple publications in the field have led not only to the availability of new invaluable data, but have also contributed substantially to an improved understanding of the physics of their dosimetry. This work provides an overview of the most important aspects that govern the physics of small megavoltage photon beam dosimetry. Copyright © 2017 Elsevier B.V. All rights reserved.

Neutron cross sections of 28 fission product nuclides adopted in JENDL-1

International Nuclear Information System (INIS)

Kikuchi, Yasuyuki; Nakagawa, Tsuneo; Igarasi, Sin-iti; Matsunobu, Hiroyuki; Kawai, Masayoshi; Iijima, Shungo.

1981-02-01

This is the final report concerning the evaluated neutron cross sections of 28 fission product nuclides adopted in the first version of Japanese Evaluated Nuclear Data Library (JENDL-1). These 28 nuclides were selected as being most important for fast reactor calculations, and are 90 Sr, 93 Zr, 95 Mo, 97 Mo, 99 Tc, 101 Ru, 102 Ru, 103 Rh, 104 Ru, 105 Pd, 106 Ru, 107 Pd, 109 Ag, 129 I, 131 Xe, 133 Cs, 135 Cs, 137 Cs, 143 Nd, 144 Ce, 144 Nd, 145 Nd, 147 Pm, 147 Sm, 149 Sm, 151 Sm, 153 Eu and 155 Eu. The status of the experimental data was reviewed over the whole energy range. The present evaluation was performed on the basis of the measured data with the aid of theoretical calculations. The optical and statical models were used for evaluation of the smooth cross sections. An improved method was developed in treating the multilevel Breit-Wigner formula for the resonance region. Various physical parameters and the level schemes, adopted in the present work are discussed by comparing with those used in the other evaluations such as ENDF/B-IV, CEA, CNEN-2 and RCN-2. Furthermore, the evaluation method and results are described in detail for each nuclide. The evaluated total, capture and inelastic scattering cross sections are compared with the other evaluated data and some recent measured data. Some problems of the present work are pointed out and ways of their improvement are suggested. (author)

Reassessment of the atomic bomb radiation dosimetry for Hiroshima and Nagasaki. Dosimetry system 2002. DS02. Volume 1

International Nuclear Information System (INIS)

Young, Robert W.; Kerr, George D.

2005-01-01

The extensive efforts to review the dosimetry of the atomic-bomb survivors and formulate the new dosimetry system DS02 have been greatly welcomed by the Radiation Effects Research Foundation (RERF). This accomplishment is a fine tribute to the importance of the epidemiological studies being conducted at RERF. No other study is so informative of the effects of radiation on human health. The gracious participation in the RERF program by the atomic-bomb survivors allows us to contribute to the well being of these individuals, and the high quality of the data obtained allows the RERF results to feature so prominently in the formulation of international guidelines for radiation protection. Such a great effort to improve and substantiate the dosimetry would not otherwise have been justified. RERF greatly appreciates the independent work of the U.S. and Japanese Working Groups on the atomic-bomb dosimetry and the review by the Joint Senior Review Group of this overall effort. We are assured that unbiased development of the new dosimetry system will reflect well in its application in the RERF epidemiology study. The documentation included in this report will serve as reference for the many deliberations concluded. The title publications are divided into 2 volumes. This is the first volume. The 8 of the reports in each chapter are indexed individually. (J.P.N.)

Reassessment of the atomic bomb radiation dosimetry for Hiroshima and Nagasaki. Dosimetry system 2002. DS02. Volume 2

International Nuclear Information System (INIS)

Young, Robert W.; Kerr, George D.

2005-01-01

The extensive efforts to review the dosimetry of the atomic-bomb survivors and formulate the new dosimetry system DS02 have been greatly welcomed by the Radiation Effects Research Foundation (RERF). This accomplishment is a fine tribute to the importance of the epidemiological studies being conducted at RERF. No other study is so informative of the effects of radiation on human health. The gracious participation in the RERF program by the atomic-bomb survivors allows us to contribute to the well being of these individuals, and the high quality of the data obtained allows the RERF results to feature so prominently in the formulation of international guidelines for radiation protection. Such a great effort to improve and substantiate the dosimetry would not otherwise have been justified. RERF greatly appreciates the independent work of the U.S. and Japanese Working Groups on the atomic-bomb dosimetry and the review by the Joint Senior Review Group of this overall effort. We are assured that unbiased development of the new dosimetry system will reflect well in its application in the RERF epidemiology study. The documentation included in this report will serve as reference for the many deliberations concluded. The title publications are divided into 2 volumes. This is the second volume. The 29 of the reports in each chapter are indexed individually. (J.P.N.)

Advances in biomedical dosimetry

International Nuclear Information System (INIS)

1981-01-01

Full text: Radiation dosimetry, the accurate determination of the absorbed dose within an irradiated body or a piece of material, is a prerequisite for all applications of ionizing radiation. This has been known since the very first radiation applications in medicine and biology, and increasing efforts are being made by radiation researchers to develop more reliable, effective and safe instruments, and to further improve dosimetric accuracy for all types of radiation used. Development of new techniques and instrumentation was particularly fast in the field of both medical diagnostic and therapeutic radiology. Thus, in Paris in October the IAEA held the latest symposium in its continuing series on dosimetry in medicine and biology. The last one was held in Vienna in 1975. High-quality dosimetry is obviously of great importance for human health, whether the objectives lie in the prevention and control of risks associated with the nuclear industry, in medical uses of radioactive substances or X-ray beams for diagnostic purposes, or in the application of photon, electron or neutron beams in radiotherapy. The symposium dealt with the following subjects: General aspects of dosimetry; Special physical and biomedical aspects; Determination of absorbed dose; Standardization and calibration of dosimetric systems; and Development of dosimetric systems. The forty or so papers presented and the discussions that followed them brought out a certain number of dominant themes, among which three deserve particular mention. - The recent generalization of the International System of Units having prompted a fundamental reassessment of the dosimetric quantities to be considered in calibrating measuring instruments, various proposals were advanced by the representatives of national metrology laboratories to replace the quantity 'exposure' (SI unit = coulomb/kg) by 'Kerma' or 'absorbed dose' (unit joule/kg, the special name of which is 'gray'), this latter being closer to the practical

Software for evaluation of EPR-dosimetry performance

International Nuclear Information System (INIS)

Shishkina, E.A.; Timofeev, Yu.S.; Ivanov, D.V.

2014-01-01

Electron paramagnetic resonance (EPR) with tooth enamel is a method extensively used for retrospective external dosimetry. Different research groups apply different equipment, sample preparation procedures and spectrum processing algorithms for EPR dosimetry. A uniform algorithm for description and comparison of performances was designed and implemented in a new computer code. The aim of the paper is to introduce the new software 'EPR-dosimetry performance'. The computer code is a user-friendly tool for providing a full description of method-specific capabilities of EPR tooth dosimetry, from metrological characteristics to practical limitations in applications. The software designed for scientists and engineers has several applications, including support of method calibration by evaluation of calibration parameters, evaluation of critical value and detection limit for registration of radiation-induced signal amplitude, estimation of critical value and detection limit for dose evaluation, estimation of minimal detectable value for anthropogenic dose assessment and description of method uncertainty. (authors)

Dosimetry-based treatment planning for molecular radiotherapy: a summary of the 2017 report from the Internal Dosimetry Task Force

Directory of Open Access Journals (Sweden)

Caroline Stokke

2017-11-01

Full Text Available Abstract Background The European directive on basic safety standards (Council directive 2013/59 Euratom mandates dosimetry-based treatment planning for radiopharmaceutical therapies. The directive comes into operation February 2018, and the aim of a report produced by the Internal Dosimetry Task Force of the European Association of Nuclear Medicine is to address this aspect of the directive. A summary of the report is presented. Results A brief review of five of the most common therapy procedures is included in the current text, focused on the potential to perform patient-specific dosimetry. In the full report, 11 different therapeutic procedures are included, allowing additional considerations of effectiveness, references to specific literature on quantitative imaging and dosimetry, and existing evidence for absorbed dose-effect correlations for each treatment. Individualized treatment planning with tracer diagnostics and verification of the absorbed doses delivered following therapy is found to be scientifically feasible for almost all procedures investigated, using quantitative imaging and/or external monitoring. Translation of this directive into clinical practice will have significant implications for resource requirements. Conclusions Molecular radiotherapy is undergoing a significant expansion, and the groundwork for dosimetry-based treatment planning is already in place. The mandated individualization is likely to improve the effectiveness of the treatments, although must be adequately resourced.

Role of secondary standard dosimetry laboratory in radiation protection program

International Nuclear Information System (INIS)

Rahman, Sohaila; Ali, Noriah Mohd.

2008-01-01

Full text: The radiation dosimetry program is an important element of operational radiation protection. Dosimetry data enable workers and radiation protection professionals to evaluate and control work practices to eliminate unnecessary exposure to ionizing radiation. The usefulness of the data produced however depends on its quality and traceability. The emphasis of the global dosimetry program is focused through the IAEA/WHO network of secondary standard dosimetry laboratories (SSDLs), which aims for the determination of SI quantities through proper traceable calibration of radiation protection equipment. The responsibility of SSDL-NUCLEAR MALAYSIA to guarantee a reliable dosimetry service, which is traceable to international standards, is elucidated. It acts as the basis for harmonized occupational radiation monitoring in Malaysia.

EPR Dosimetry - Present and Future

Energy Technology Data Exchange (ETDEWEB)

Regulla, D.F. [GSF - National Research Centre for Environment and Health, Institute of Radiation Protection, 85764 Neuherberg (Germany)

1999-07-01

In the past, IAEA has played a central role in stipulating research and development in EPR high-dose standardisation as well as in coordinating and organising international dose intercomparison programs, within the Member States of the United Nations from the mid-seventies till today. The future tasks of EPR dosimetry seem to tend towards different subjects such as bio markers, biological radiation effects, post-accident dose reconstruction in the environment, and retrospective human dosimetry. The latter may be considered a promising tool for epidemiology on the way to re-define radiation risk of man for chronicle radiation exposures, based on e.g. South Ural civil population and radiation workers. There are on-going international activities in the field of standardising high-level dosimetry by the American Standards on Testing and Materials (Astm), and by the International Organisation of Standards (ISO). The International Commission on Radiation Units and Measurements (ICRU) is considering the establishment of relevant recommendations concerning industrial radiation processing, but also human dose reconstruction. (Author)

Fundamentals of Dosimetry. Chapter 3

Energy Technology Data Exchange (ETDEWEB)

Yoshimura, E. M. [Universidade de São Paulo, São Paulo (Brazil)

2014-09-15

Determination of the energy imparted to matter by radiation is the subject of dosimetry. The energy deposited as radiation interacts with atoms of the material, as seen in the previous chapter. The imparted energy is responsible for the effects that radiation causes in matter, for instance, a rise in temperature, or chemical or physical changes in the material properties. Several of the changes produced in matter by radiation are proportional to the absorbed dose, giving rise to the possibility of using the material as the sensitive part of a dosimeter. Also, the biological effects of radiation depend on the absorbed dose. A set of quantities related to the radiation field is also defined within the scope of dosimetry. It will be shown in this chapter that, under special conditions, there are simple relations between dosimetric and field description quantities. Thus, the framework of dosimetry is the set of physical and operational quantities that are studied in this chapter.

EPR Dosimetry - Present and Future

International Nuclear Information System (INIS)

Regulla, D.F.

1999-01-01

In the past, IAEA has played a central role in stipulating research and development in EPR high-dose standardisation as well as in coordinating and organising international dose intercomparison programs, within the Member States of the United Nations from the mid-seventies till today. The future tasks of EPR dosimetry seem to tend towards different subjects such as bio markers, biological radiation effects, post-accident dose reconstruction in the environment, and retrospective human dosimetry. The latter may be considered a promising tool for epidemiology on the way to re-define radiation risk of man for chronicle radiation exposures, based on e.g. South Ural civil population and radiation workers. There are on-going international activities in the field of standardising high-level dosimetry by the American Standards on Testing and Materials (Astm), and by the International Organisation of Standards (ISO). The International Commission on Radiation Units and Measurements (ICRU) is considering the establishment of relevant recommendations concerning industrial radiation processing, but also human dose reconstruction. (Author)

I-124 Imaging and Dosimetry

Directory of Open Access Journals (Sweden)

Russ Kuker

2017-02-01

Full Text Available Although radioactive iodine imaging and therapy are one of the earliest applications of theranostics, there still remain a number of unresolved clinical questions as to the optimization of diagnostic techniques and dosimetry protocols. I-124 as a positron emission tomography (PET radiotracer has the potential to improve the current clinical practice in the diagnosis and treatment of differentiated thyroid cancer. The higher sensitivity and spatial resolution of PET/computed tomography (CT compared to standard gamma scintigraphy can aid in the detection of recurrent or metastatic disease and provide more accurate measurements of metabolic tumor volumes. However the complex decay schema of I-124 poses challenges to quantitative PET imaging. More prospective studies are needed to define optimal dosimetry protocols and to improve patient-specific treatment planning strategies, taking into account not only the absorbed dose to tumors but also methods to avoid toxicity to normal organs. A historical perspective of I-124 imaging and dosimetry as well as future concepts are discussed.

Use of the GATE Monte Carlo package for dosimetry applications

Energy Technology Data Exchange (ETDEWEB)

Visvikis, D. [INSERM U650, LaTIM, University Hospital Medical School, F 29609 Brest (France)]. E-mail: Visvikis.Dimitris@univ-brest.fr; Bardies, M. [INSERM U601, CHU Nantes, F 44093 Nantes (France); Chiavassa, S. [INSERM U601, CHU Nantes, F 44093 Nantes (France); Danford, C. [Department of Medical Physics, MSKCC, New York (United States); Kirov, A. [Department of Medical Physics, MSKCC, New York (United States); Lamare, F. [INSERM U650, LaTIM, University Hospital Medical School, F 29609 Brest (France); Maigne, L. [Departement de Curietherapie-Radiotherapie, Centre Jean Perrin, F 63000 Clemont-Ferrand (France); Staelens, S. [UGent-ELIS, St-Pietersnieuwstraat, 41, B 9000 Gent (Belgium); Taschereau, R. [CRUMP Institute for Molecular Imaging, UCLA, Los Angeles (United States)

2006-12-20

One of the roles for Monte Carlo (MC) simulation studies is in the area of dosimetry. A number of different codes dedicated to dosimetry applications are available and widely used today, such as MCNP, EGSnrc and PTRAN. However, such codes do not easily facilitate the description of complicated 3D sources or emission tomography systems and associated data flow, which may be useful in different dosimetry application domains. Such problems can be overcome by the use of specific MC codes such as GATE (GEANT4 Application to Tomographic Emission), which is based on Geant4 libraries, providing a scripting interface with a number of advantages for the simulation of SPECT and PET systems. Despite this potential, its major disadvantage is in terms of efficiency involving long execution times for applications such as dosimetry. The strong points and disadvantages of GATE in comparison to other dosimetry specific codes are discussed and illustrated in terms of accuracy, efficiency and flexibility. A number of features, such as the use of voxelised and moving sources, as well as developments such as advanced visualization tools and the development of dose estimation maps allowing GATE to be used for dosimetry applications are presented. In addition, different examples from dosimetry applications with GATE are given. Finally, future directions with respect to the use of GATE for dosimetry applications are outlined.

Database to manage personal dosimetry Hospital Universitario de La Ribera

International Nuclear Information System (INIS)

Melchor, M.; Martinez, D.; Asensio, M.; Candela, F.; Camara, A.

2011-01-01

For the management of professionally exposed personnel dosimetry, da La are required for the use and return of dosimeters. in the Department of Radio Physics and Radiation Protection have designed and implemented a database management staff dosimetry Hospital and Area Health Centers. The specific objectives were easily import data from the National Center dosimetric dosimetry, consulting records in a simple dosimetry, dosimeters allow rotary handle, and also get reports from different periods of time to know the return data for users, services, etc.

Activity Of EURADOS In Environmental Solid State Dosimetry

International Nuclear Information System (INIS)

Ranogajec-Komor, M.; Duch, M. A.; Haninger, T.

2015-01-01

Working Group 3 (WG3) of the European Radiation Dosimetry Group (EURADOS) carries out research projects and coordinated activities to advance the scientific understanding of environmental dosimetry and especially to promote the technical development of new methods in environmental monitoring. In this field of dosimetry, the measurement of small additional doses caused by artificial radiation on top of the natural environmental radiation is a challenge. Further, WG3 stimulates the organisation of intercomparison programmes and the definition of standards and recommendations in the field of environmental radiation monitoring (ERM). WG3 has played a significant role in the harmonisation of early warning dosimetry network stations in Europe and has organised 6 EURADOS intercomparison exercises; in which 42 institutions from 19 countries have participated. Today, about 5000 stations provide real-time dose rate data to a database run by the European Commission. Within WG3 a subgroup (S1) on spectrometry system was formed in 2013. Since then, WG3 has been involved in the field of spectrometry systems used both for dosimetric and spectrometric monitoring in the environment. A remarkable result of the WG3 - S1 is that many members contributed to the new European Joint Research Project 'Metrology for radiological early warning networks in Europe' which started in 2014. A second subgroup WG3 - S2 on passive dosimetry in ERM was inaugurated in 2014. To gain an overview of the passive dosimetry practice in ERM, WG3 - S2 decided to collect information by means of a questionnaire which has been send to European dosimetry services. One of the results was the identification of some open questions, problems in ERM (for example terminology, protocol of routine dosimetry, uncertainty assessment) which require clarification for harmonisation of ERM using passive dosimeters. Another result was that there exists a need for intercomparisons. The first intercomparison for passive

Accuracy and efficiency of published film dosimetry techniques using a flat-bed scanner and EBT3 film.

Science.gov (United States)

Spelleken, E; Crowe, S B; Sutherland, B; Challens, C; Kairn, T

2018-03-01

Gafchromic EBT3 film is widely used for patient specific quality assurance of complex treatment plans. Film dosimetry techniques commonly involve the use of transmission scanning to produce TIFF files, which are analysed using a non-linear calibration relationship between the dose and red channel net optical density (netOD). Numerous film calibration techniques featured in the literature have not been independently verified or evaluated. A range of previously published film dosimetry techniques were re-evaluated, to identify whether these methods produce better results than the commonly-used non-linear, netOD method. EBT3 film was irradiated at calibration doses between 0 and 4000 cGy and 25 pieces of film were irradiated at 200 cGy to evaluate uniformity. The film was scanned using two different scanners: The Epson Perfection V800 and the Epson Expression 10000XL. Calibration curves, uncertainty in the fit of the curve, overall uncertainty and uniformity were calculated following the methods described by the different calibration techniques. It was found that protocols based on a conventional film dosimetry technique produced results that were accurate and uniform to within 1%, while some of the unconventional techniques produced much higher uncertainties (> 25% for some techniques). Some of the uncommon methods produced reliable results when irradiated to the standard treatment doses (film analysis technique which uses transmission scanning, red colour channel analysis, netOD and a non-linear calibration curve for measuring doses up to 4000 cGy when using EBT3 film.

Medical reference dosimetry using EPR measurements of alanine

DEFF Research Database (Denmark)

Helt-Hansen, Jakob; Rosendal, F.; Kofoed, I.M.

2009-01-01

Background. Electron spin resonance (EPR) is used to determine the absorbed dose of alanine dosimeters exposed to clinical photon beams in a solid-water phantom. Alanine is potentially suitable for medical reference dosimetry, because of its near water equivalence over a wide energy spectrum, low...... methods the proposed algorithm can be applied without normalisation of phase shifts caused by changes in the g-value of the cavity. The study shows that alanine dosimetry is a suitable candidate for medical reference dosimetry especially for quality control applications.......Background. Electron spin resonance (EPR) is used to determine the absorbed dose of alanine dosimeters exposed to clinical photon beams in a solid-water phantom. Alanine is potentially suitable for medical reference dosimetry, because of its near water equivalence over a wide energy spectrum, low...

Review of recent benchmark experiments on integral test for high energy nuclear data evaluation

Energy Technology Data Exchange (ETDEWEB)

Nakashima, Hiroshi; Tanaka, Susumu; Konno, Chikara; Fukahori, Tokio; Hayashi, Katsumi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-11-01

A survey work of recent benchmark experiments on an integral test for high energy nuclear data evaluation was carried out as one of the work of the Task Force on JENDL High Energy File Integral Evaluation (JHEFIE). In this paper the results are compiled and the status of recent benchmark experiments is described. (author)

The IAEA/WHO thermoluminescent dosimetry intercomparison used for the improvement of clinical dosimetry

International Nuclear Information System (INIS)

Racoveanu, N.T.

1981-01-01

Results of thermoluminescent dosimetry collected over 5 years in the Eastern Mediterranean region of WHO were analyzed in an attempt to improve clinical dosimetry. Data for 16 radiotherapy departments showed considerable inconsistencies. It was found that the clinical dosemeters used by 3 of the departments were not working properly. The remainder of the departments had one or more dosemeters in perfect working order but the procedure for measuring machine output was inadequate or the correction factors (pressure, temperature) were wrongly applied due to lack of reliable instruments for such measurements. Problems encountered in the sending and returning of TLD dosemeters for assessment are discussed

Hanford External Dosimetry Technical Basis Manual PNL-MA-842

Energy Technology Data Exchange (ETDEWEB)

Rathbone, Bruce A.

2010-04-01

The Hanford External Dosimetry Technical Basis Manual PNL-MA-842 documents the design and implementation of the external dosimetry system used at the U.S. Department of Energy (DOE) Hanford site. The manual describes the dosimeter design, processing protocols, dose calculation methodology, radiation fields encountered, dosimeter response characteristics, limitations of dosimeter design under field conditions, and makes recommendations for effective use of the dosimeters in the field. The manual describes the technical basis for the dosimetry system in a manner intended to help ensure defensibility of the dose of record at Hanford and to demonstrate compliance with requirements of 10 CFR 835, the DOE Laboratory Accreditation Program, the DOE Richland Operations Office, DOE Office of River Protection, DOE Pacific Northwest Office of Science, and Hanford’s DOE contractors. The dosimetry system is operated by the Pacific Northwest National Laboratory (PNNL) Hanford External Dosimetry Program which provides dosimetry services to PNNL and all Hanford contractors. The primary users of this manual are DOE and DOE contractors at Hanford using the dosimetry services of PNNL. Development and maintenance of this manual is funded directly by DOE and DOE contractors. Its contents have been reviewed and approved by DOE and DOE contractors at Hanford through the Hanford Personnel Dosimetry Advisory Committee which is chartered and chaired by DOE-RL and serves as means of coordinating dosimetry practices across contractors at Hanford. This manual was established in 1996. Since its inception, it has been revised many times and maintained by PNNL as a controlled document with controlled distribution. The first revision to be released through PNNL’s Electronic Records & Information Capture Architecture database was designated Revision 0. Revision numbers that are whole numbers reflect major revisions typically involving significant changes to all chapters in the document. Revision

Hanford External Dosimetry Technical Basis Manual PNL-MA-842

Energy Technology Data Exchange (ETDEWEB)

Rathbone, Bruce A.

2011-04-04

The Hanford External Dosimetry Technical Basis Manual PNL-MA-842 documents the design and implementation of the external dosimetry system used at the U.S. Department of Energy (DOE) Hanford site. The manual describes the dosimeter design, processing protocols, dose calculation methodology, radiation fields encountered, dosimeter response characteristics, limitations of dosimeter design under field conditions, and makes recommendations for effective use of the dosimeters in the field. The manual describes the technical basis for the dosimetry system in a manner intended to help ensure defensibility of the dose of record at Hanford and to demonstrate compliance with requirements of 10 CFR 835, the DOE Laboratory Accreditation Program, the DOE Richland Operations Office, DOE Office of River Protection, DOE Pacific Northwest Office of Science, and Hanford’s DOE contractors. The dosimetry system is operated by the Pacific Northwest National Laboratory (PNNL) Hanford External Dosimetry Program which provides dosimetry services to PNNL and all Hanford contractors. The primary users of this manual are DOE and DOE contractors at Hanford using the dosimetry services of PNNL. Development and maintenance of this manual is funded directly by DOE and DOE contractors. Its contents have been reviewed and approved by DOE and DOE contractors at Hanford through the Hanford Personnel Dosimetry Advisory Committee which is chartered and chaired by DOE-RL and serves as means of coordinating dosimetry practices across contractors at Hanford. This manual was established in 1996. Since its inception, it has been revised many times and maintained by PNNL as a controlled document with controlled distribution. The first revision to be released through PNNL’s Electronic Records & Information Capture Architecture database was designated Revision 0. Revision numbers that are whole numbers reflect major revisions typically involving significant changes to all chapters in the document. Revision

Nuclear Data Processing for Reactor Physics Calculation

International Nuclear Information System (INIS)

Suwoto; Zuhair; Pandiangan, Tumpal

2003-01-01

Nuclear data processing for reactor physics calculation has been done. Raw nuclear data cross-sections on file ENDF should be prepared and processed before it used in neutronic calculation. The processing code system such as NJOY-PC code has been used from linearization of nuclear cross-sections data and background contribution of resonance parameter (MF2) using RECONR module (0K) with energy range from 10 -5 to 10 7 eV. Afterward, the neutron cross-sections data should be processed and broadened to desire temperature (i.e. 293K) by using BROADR module. The Grouper and Therma modules will be applied for multi-groups calculation which suitable for WIMS/D4 (69 groups) and thermalization of nuclear constants. The final stage of processing nuclear cross-sections is updating WIMS/D4 library. The WIMSR module in NJOY-PC and WILLIE code will be applied in this stage. The evaluated nuclear data file, especially for 1 H 1 isotope, was taken from JENDL-3.2 and ENDF/B-VI for preliminary study. The results of nuclear data processing 1 H 1 shows that the old-WIMS (WIMS-lama) library have much discrepancies comparing with JENDL-3.2 or ENDF/B-VI files, especially in energy around 5 keV

Development of covariance date for fast reactor cores. 3

International Nuclear Information System (INIS)

Shibata, Keiichi; Hasegawa, Akira

1999-03-01

Covariances have been estimated for nuclear data contained in JENDL-3.2. As for Cr and Ni, the physical quantities for which covariances are deduced are cross sections and the first order Legendre-polynomial coefficient for the angular distribution of elastically scattered neutrons. The covariances were estimated by using the same methodology that had been used in the JENDL-3.2 evaluation in order to keep a consistency between mean values and their covariances. In a case where evaluated data were based on experimental data, the covariances were estimated from the same experimental data. For cross section that had been evaluated by nuclear model calculations, the same model was applied to generate the covariances. The covariances obtained were compiled into ENDF-6 format files. The covariances, which had been prepared by the previous fiscal year, were re-examined, and some improvements were performed. Parts of Fe and 235 U covariances were updated. Covariances of nu-p and nu-d for 241 Pu and of fission neutron spectra for 233,235,238 U and 239,240 Pu were newly added to data files. (author)

Computerized dosimetry management systems within EDF

International Nuclear Information System (INIS)

Daubert, G.

1996-01-01

EDF, using the ALARA approach, has embarked an ambitious project of optimising the doses received in its power plants. In directing its choice of actions and the effectiveness of such actions, the French operator is using a computerized personal and collective dosimetry management system. This system provides for ongoing monitoring of dosimetry at personal, site and unit level or indeed for the entire population of EDF nuclear power plants. (author)

In vivo dosimetry in radiation therapy in Sweden

International Nuclear Information System (INIS)

Eriksson, Jacob; Blomquist, Michael

2010-07-01

A prerequisite for achieving high radiation safety for patients receiving external beam radiation therapy is that the hospitals have a quality assurance program. The program should include include monitoring of the radiation dose given to the patient. Control measurements are performed both at the system level and at the individual level. Control measurement is normally performed using in vivo dosimetry, e.g. a method to measure the radiation dose at the individual level during the actual radiation treatment time. In vivo dosimetry has proven to be an important tool to detect and prevent serious errors in patient treatment. The purpose of this research project was to identify the extent to which vivo dosimetry is used and the methods available for this at Swedish radiation therapy clinics. The authority also wanted to get an overall picture of how hospitals manage results of in vivo dosimetry, and how clinics control radiation dose when using modern treatment techniques. The report reflects the situation in Swedish radiotherapy clinics 2007. The report shows that all hospitals use some form of in vivo dosimetry. The instruments used are mainly diodes and termoluminiscence dosimeters

Trends of personal dosimetry at atomic power plants

International Nuclear Information System (INIS)

Yamamura, Seini

1998-01-01

The individual dosimetry at the atomic power station is sorted for monthly dosimetry, daily dosimetry and special job dosimetry in high dose circumstance. Film badge (passive dosimeter) can measure gamma dose, beta dose and neutron dose respectively lower than about 0.1 mSv. While workers are in the radiation controlled area, they have to wear the dosimeters and the individual dose is accumulated for every one month. Recently the Silicon semiconductors detecting beta ray and neutron have been developed. With microcircuit technology and these new sensors, new multiple function dosimeter of the card size had been put to practical use. The result of dose measurement obtained by the electronic dosimeter is consistent well with the measurement of usual film badge and new dosimeter can determine the dose as low as 0.01 mSv. The result is stored in the non-volatile memory in the electronic personal dosimeter and held for more than one year without the power supply. The function to read data directly from the memory improves the reliability of the data protection. The realization of the unified radiation control system that uses the electronic personal dosimeter for monthly dosimetry is expected. (J.P.N.)

Information from the Dosimetry Service

CERN Multimedia

2006-01-01

CERN Staff and Users can now consult their dose records for an individual or an organizational unit with HRT. Please see more information on our web page http://cern.ch/rp-dosimetry. The Dosimetry Service is open every morning from 8.30 - 12.00, and closed in the afternoons. We would like to remind you that dosimeters cannot be sent to customers by internal mail. Short-term dosimeters (VCT's) must always be returned to the Service after use and must not be left on the racks in the experimental areas or in the secretariats.

European protocol for neutron dosimetry for external beam therapy

International Nuclear Information System (INIS)

Broerse, J.J.; Mijnheer, B.J.; Williams, J.R.

1981-01-01

The paper attempts to serve the needs of European centres participating in the High LET Therapy Project Group set up under the sponsorship of The European Organization for Research on Treatment of Cancer, to promote cooperation between physicists involved in fast neutron therapy and establish a common basis for neutron dosimetry. Differences in dosimetry procedures between European and American Groups are indicated if relevant. The subject is dealt with under the following main headings: principles of dosimetry of neutron fields, dosimetric methods, physical parameters, determination of absorbed dose at a reference point, determination of absorbed dose at any point, check of absorbed dose given to a patient, dosimetry intercomparisons between institutes. There is an ample bibliography. (U.K.)

Neutron beam measurement dosimetry

International Nuclear Information System (INIS)

Amaro, C.R.

1995-01-01

This report describes animal dosimetry studies and phantom measurements. During 1994, 12 dogs were irradiated at BMRR as part of a 4 fraction dose tolerance study. The animals were first infused with BSH and irradiated daily for 4 consecutive days. BNL irradiated 2 beagles as part of their dose tolerance study using BPA fructose. In addition, a dog at WSU was irradiated at BMRR after an infusion of BPA fructose. During 1994, the INEL BNCT dosimetry team measured neutron flux and gamma dose profiles in two phantoms exposed to the epithermal neutron beam at the BMRR. These measurements were performed as a preparatory step to the commencement of human clinical trials in progress at the BMRR

WIPP radiation dosimetry program

International Nuclear Information System (INIS)

Wu, C.F.

1991-01-01

Radiation dosimetry is the process by which various measurement results and procedures are applied to quantify the radiation exposure of an individual. Accurate and precise determination of radiation dose is a key factor to the success of a radiation protection program. The Waste Isolation Pilot Plant (WIPP), a Department of Energy (DOE) facility designed for permanent repository of transuranic wastes in a 2000-foot-thick salt bed 2150 feet underground, has established a dosimetry program developed to meet the requirements of DOE Order 5480.11, ''Radiation Protection for Occupational Workers''; ANSI/ASME NQA-1, ''Quality Assurance Program Requirements for Nuclear Facilities''; DOE Order 5484.1, ''Environmental Protection, Safety, and Health Protection Information Reporting Requirements''; and other applicable regulations

High-accuracy dosimetry study for intensity-modulated radiation therapy(IMRT) commissioning

International Nuclear Information System (INIS)

Jeong, Hae Sun

2010-02-01

.5 cm 2 ). In addition, a method using a pixel-based unfolding curve was developed and applied to correct the non-uniform response of flat-bed type scanners for a radiochromic film. Also, the accuracy of the method was finally evaluated by comparing the results with those of an ion chamber, Monte Carlo simulation, and CF-based conventional method. For individual dose, the dosimetric error of using conventional method and using the pixel-based unfolding curve was reduced to less than 3%, and 1%, respectively. In case of step-wise doses, the average difference of 16% with MC calculation was reduced up to 1% by using the correction method in this study. Consequently, the accuracy of dose computation algorithms in TPS can be evaluated by the developed LEGO-type solid phantom, small filed dosimetry, the correction method for non-uniform response of scanners. It is also recognized that the developed hardware and software which are possible to be used for QA procedure are very reliable and they could be used for reference study of other radiation therapies

NMLIB

International Nuclear Information System (INIS)

Konno, Chikara; Ikeda, Yujiro

2003-01-01

A new general purpose neutron incident nuclear data library, NMLIB, for neutron energy up to 3 GeV is generated by using NMTC/JAM for neutron cross sections above 20 MeV and JENDL-3.2 or -3.3 for neutron cross sections below 20 MeV. MATXS files (NMLIB/MG) and ACE files (NMLIB/MC) of NMLIB are also produced from HMLIB with the NJOY code. A simple test calculation suggests that NMLIB, NMLIB/MG and NMLIB/MC have no problem except for self-shielding. (author)

Database for 238U inelastic scattering cross section evaluation

International Nuclear Information System (INIS)

Kanda, Yukinori; Fujikawa, Noboru; Kawano, Toshihiko

1993-10-01

There are discrepancies among evaluated neutron inelastic scattering cross sections for 238 U in the evaluated nuclear data files, JENDL-3, ENDF/B-VI, JEF-2, BROND-2 and CENDL-2. Re-evaluating them is internationally being discussed to obtain the best outcome which can be accepted in common at the present by experts in the world. This report has been compiled to review the discrepancies among the evaluations in the present data files and to provide a common database for the re-evaluation work (author)

Development and implementation of own software for dosimetry multichannel film

International Nuclear Information System (INIS)

Jimenez Feltstrom, D.; Reyes Garcia, R.; Luis Simon, F. J.; Carrasco Herrera, M.; Sanchez Carmona, G.; Herrador Cordoba, M.

2013-01-01

The objective of this work is to develop its own software for multichannel film dosimetry Radiochromic EBT2. Compare the results obtained with its use in multichannel and single-channel dosimetry. Check that the multi-channel dosimetry eliminates much of the artifacts caused by dirt, fingerprints, scratches, etc. Radiochromic in film and scanner devices. (Author)

Report of a consultants meeting on dosimetry in diagnostic radiology

International Nuclear Information System (INIS)

Pernicka, F.

1999-01-01

During its biennial meeting in 1996, the Standing Advisory Committee 'SSDL Scientific Committee', recommended extending the long experience of the Agency in the field of standardization and monitoring dosimetry calibrations at radiotherapy and radiation protection level for the Secondary Standard Dosimetry Laboratory (SSDL) Network, to the field of diagnostic x-ray dosimetry. It was emphasized that 'Measurements on diagnostic x-ray machines have become increasingly important and some SSDLs are involved in such measurements. The Agency's dosimetry laboratory should, therefore, have proper radiation sources available to provide traceable calibrations to the SSDLs'. The purpose of the consultants' meeting was to advise the Agency on dosimetry in diagnostic radiology. They were specifically requested to overview scientific achievements in the field and to give advice to the Agency on the need for further developments. The purpose of the consultants' meeting was to advise the Agency on dosimetry in diagnostic radiology. They were specifically requested to overview scientific achievements in the field and to give advice to the Agency on the need for further developments

Biological dosimetry in cases gives occupational high exposition to ionizing radiations

International Nuclear Information System (INIS)

Ramalho, Adriana T.; Costa, Maria Lucia P.; Oliveira, Monica S.; Silva, Francisco Cesar A.

1998-01-01

From 1983 the cytogenetics dosimetry method it has been used as routine in the IRD laboratory in the period 1983 at 1997 but a high exposition occupational case the physical dosimeters happened in Brazil they were investigated through the cytogenetics dosimetry technique. This technique is employ when the dosimetry personal marks a high dose to 100 mSv (0,1 Gy) that is the cut-off minimum detected in the dosimetry cytogenetics

Benchmark calculations for critical experiments at FKBN-M facility with uranium-plutonium-polyethylene systems using JENDL-3.2 and MVP Monte-Carlo code

International Nuclear Information System (INIS)

Obara, Toru; Morozov, A.G.; Kevrolev, V.V.; Kuznetsov, V.V.; Treschalin, S.A.; Lukin, A.V.; Terekhin, V.A.; Sokolov, Yu.A.; Kravchenko, V.G.

2000-01-01

Benchmark calculations were performed for critical experiments at FKBN-M facility in RFNC-VNIITF, Russia using JENDL-3.2 nuclear data library and continuous energy Monte-Carlo code MVP. The fissile materials were high-enriched uranium and plutonium. Polyethylene was used as moderator. The neutron spectrum was changed by changing the geometry. Calculation results by MVP showed some errors. Discussion was made by reaction rates and η values obtained by MVP. It showed the possibility that cross sections of U-235 had different trend of error in fast and thermal energy region respectively. It also showed the possibility of some error of cross section of Pu-239 in high energy region. (author)

The creation of voxel phantoms for the purpose of environmental dosimetry

Energy Technology Data Exchange (ETDEWEB)

Caffrey, E.; Higley, K. [Oregon State University (United States)

2014-07-01

Basic geometric shapes have long been used as the standard for calculating radiation dose rates in non-human biota (NHB). Regulation standards have seen a shift recently, towards protection of NHB as its own endpoint. As such, there has been a growing interest in improving the calculations for NHB dose rates. To address calls for additional data, the development of voxelized models for the International Commission on Radiological Protection's (ICRP) twelve reference animal and plants (RAP) has been undertaken. Voxel models of a crab (Metacarcinus magister), flatfish (Pleuronectiformes), trout (Oncorhynchus mykiss), worm (Lumbricina), honey bee (Apis), frog (Anura), and rat, (Rattus) have been created to date. The purpose of this submission is to describe the processes of creating these voxel phantoms from radiological imaging data (i.e., Computed Tomography (CT), Magnetic Resonance Imaging (MRI), etc.). CT/MRI images of the organism are obtained and uploaded into a software package capable of segmenting the images (3D Doctor was used for the crab, flatfish, trout, worm, and honey bee). On each image slice, individual organs and other relevant anatomical features (e.g. bones or other structural tissues) are identified and segmented. Once segmentation is complete, a boundary file that describes the positioning of the organs and tissues in lattice geometry format is exported into software called Voxelizer, created by the Human Monitoring Laboratory of Canada. This software writes the boundary file geometry into an input file for Monte Carlo N-Particle (MCNP) based simulations. The user can then add appropriate materials, densities, and a desired source term. These simulations yield absorbed fraction (AF) values that are used in subsequent dose calculations with environmental concentration data. AFs are now available for the crab, flatfish, trout, worm, and honey bee at twelve photon and nine electron energies, consistent with ICRP AFs for human dosimetry

EPR dosimetry - present and future

International Nuclear Information System (INIS)

Regulla, D.F.

1999-01-01

In the past, IAEA has played a central role in stipulating research and development in EPR high-dose standardisation as well as co-ordinating and organising international dose intercomparison programs, within the Member States of the United Nations from the mid-seventies till today. The future tasks of EPR dosimetry seem to tend towards different subjects such as biomarkers, biological radiation effects, post-accident dose reconstruction in the environment, and retrospective human dosimetry. The latter may be considered a promising tool for epidemiology on the way to re-define radiation risk of man for chronicle radiation exposures, based on e.g. South Ural civil population and radiation workers. There are on-going international activities in the field of standardising high-level dosimetry by the American Standards on Testing and Materials (ASTM), and the International Organisation of Standards (ISO) as well as those of the International Commission on Radiation Units and Measurements (ICRU) considering the establishment of relevant recommendations concerning industrial radiation processing, but also human dose reconstruction. (author)

User's manual of a supporting system for treatment planning in boron neutron capture therapy. JAERI computational dosimetry system

Energy Technology Data Exchange (ETDEWEB)

Kumada, Hiroaki; Torii, Yoshiya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2002-09-01

A boron neutron capture therapy (BNCT) with epithermal neutron beam is expected to treat effectively for malignant tumor that is located deeply in the brain. It is indispensable to estimate preliminarily the irradiation dose in the brain of a patient in order to perform the epithermal neutron beam BNCT. Thus, the JAERI Computational Dosimetry System (JCDS), which can calculate the dose distributions in the brain, has been developed. JCDS is a software that creates a 3-dimensional head model of a patient by using CT and MRI images and that generates a input data file automatically for calculation neutron flux and gamma-ray dose distribution in the brain by the Monte Carlo code: MCNP, and that displays the dose distribution on the head model for dosimetry by using the MCNP calculation results. JCDS has any advantages as follows; By treating CT data and MRI data which are medical images, a detail three-dimensional model of patient's head is able to be made easily. The three-dimensional head image is editable to simulate the state of a head after its surgical processes such as skin flap opening and bone removal for the BNCT with craniotomy that are being performed in Japan. JCDS can provide information for the Patient Setting System to set the patient in an actual irradiation position swiftly and accurately. This report describes basic design and procedure of dosimetry, operation manual, data and library structure for JCDS (ver.1.0). (author)

Phantom positioning variation in the Gamma Knife® Perfexion dosimetry

Energy Technology Data Exchange (ETDEWEB)

Costa, Nathalia Almeida; Potiens, Maria da Penha Albuquerque [Instituto de Pesquisas Energeticas e Nucleres (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Saraiva, Crystian [Hospital do Coracao, Sao Paulo, SP (Brazil)

2015-07-01

The use of small volume ionization chamber has become required for the dosimetry of equipment that use small radiation fields. A pinpoint ionization chamber is ideal for the dosimetry of a Gamma Knife® Perfexion (GKP) unit. In this work, this chamber was inserted into the phantom, and measurements were performed with the phantom in different positions, in order to verify if the change in the phantom positioning affects the dosimetry of the GKP. Three different phantom positions were performed. The variation in the result is within the range allowed for the dosimetry of a GKP equipment. (author)

Cytogenetic dosimetry in suspected cases of ionizing radiation occupational exposure

International Nuclear Information System (INIS)

Ramalho, Adriana T.; Costa, Maria Lucia P.; Oliveira, Monica S.; Silva, Francisco Cesar A. da

2001-01-01

Cytogenetic dosimetry is very useful in routine as well as in serious accident situations in which exposed individuals do not wear physical dosimeters. Since 1984, the technique of cytogenetic dosimetry has been used as a routine in our laboratory at IRD/CNEN to complement the data of physical dosimetry. In the period from 1984 to 2000, 138 cases of occupational overexposure of individual dosimeters were investigated by us. In total, only in 36 of the 138 cases investigated the overexposure was confirmed by cytogenetic dosimetry. The data indicates a total confirmation index of just 26% of the suspected cases.(author)

Hanford External Dosimetry Technical Basis Manual PNL-MA-842

International Nuclear Information System (INIS)

Rathbone, Bruce A.

2007-01-01

The Hanford External Dosimetry Technical Basis Manual PNL-MA-842 documents the design and implementation of the external dosimetry system used at Hanford. The manual describes the dosimeter design, processing protocols, dose calculation methodology, radiation fields encountered, dosimeter response characteristics, limitations of dosimeter design under field conditions, and makes recommendations for effective use of the dosimeters in the field. The manual describes the technical basis for the dosimetry system in a manner intended to help ensure defensibility of the dose of record at Hanford and to demonstrate compliance with 10 CFR 835, DOELAP, DOE-RL, ORP, PNSO, and Hanford contractor requirements. The dosimetry system is operated by PNNL's Hanford External Dosimetry Program (HEDP) which provides dosimetry services to all Hanford contractors. The primary users of this manual are DOE and DOE contractors at Hanford using the dosimetry services of PNNL. Development and maintenance of this manual is funded directly by DOE and DOE contractors. Its contents have been reviewed and approved by DOE and DOE contractors at Hanford through the Hanford Personnel Dosimetry Advisory Committee (HPDAC) which is chartered and chaired by DOE-RL and serves as means of coordinating dosimetry practices across contractors at Hanford. This manual was established in 1996. Since inception, it has been revised many times and maintained by PNNL as a controlled document with controlled distribution. Rev. 0 marks the first revision to be released through PNNL's Electronic Records and Information Capture Architecture (ERICA) database. Revision numbers that are whole numbers reflect major revisions typically involving changes to all chapters in the document. Revision numbers that include a decimal fraction reflect minor revisions, usually restricted to selected chapters or selected pages in the document. Revision Log: Rev. 0 (2/25/2005) Major revision and expansion. Rev. 0.1 (3/12/2007) Minor

Hanford External Dosimetry Technical Basis Manual PNL-MA-842

Energy Technology Data Exchange (ETDEWEB)

Rathbone, Bruce A.

2007-03-12

The Hanford External Dosimetry Technical Basis Manual PNL-MA-842 documents the design and implementation of the external dosimetry system used at Hanford. The manual describes the dosimeter design, processing protocols, dose calculation methodology, radiation fields encountered, dosimeter response characteristics, limitations of dosimeter design under field conditions, and makes recommendations for effective use of the dosimeters in the field. The manual describes the technical basis for the dosimetry system in a manner intended to help ensure defensibility of the dose of record at Hanford and to demonstrate compliance with 10 CFR 835, DOELAP, DOE-RL, ORP, PNSO, and Hanford contractor requirements. The dosimetry system is operated by PNNL’s Hanford External Dosimetry Program (HEDP) which provides dosimetry services to all Hanford contractors. The primary users of this manual are DOE and DOE contractors at Hanford using the dosimetry services of PNNL. Development and maintenance of this manual is funded directly by DOE and DOE contractors. Its contents have been reviewed and approved by DOE and DOE contractors at Hanford through the Hanford Personnel Dosimetry Advisory Committee (HPDAC) which is chartered and chaired by DOE-RL and serves as means of coordinating dosimetry practices across contractors at Hanford. This manual was established in 1996. Since inception, it has been revised many times and maintained by PNNL as a controlled document with controlled distribution. Rev. 0 marks the first revision to be released through PNNL’s Electronic Records & Information Capture Architecture (ERICA) database. Revision numbers that are whole numbers reflect major revisions typically involving changes to all chapters in the document. Revision numbers that include a decimal fraction reflect minor revisions, usually restricted to selected chapters or selected pages in the document. Revision Log: Rev. 0 (2/25/2005) Major revision and expansion. Rev. 0.1 (3/12/2007) Minor

Nuclear decay data for dosimetry calculation. Revised data of ICRP Publication 38

International Nuclear Information System (INIS)

Endo, Akira; Yamaguchi, Yasuhiro

2005-02-01

New nuclear decay data used for dose calculation have been compiled for 1034 radionuclides, which are significant in medical, environmental and occupational exposures. The decay data were assembled from decay data sets of the Evaluated Nuclear Structure Data File (ENSDF), the latest version as of 2003. Basic nuclear properties in the ENSDF that are particularly important for calculating energies and intensities of radiations were examined and updated by referring to UNBASE2003/AME2003, the database for nuclear and decay properties of nuclides. In addition, modification of incomplete ENSDF was done for their format errors, level schemes, normalization records, and so on. The energies and intensities of emitted radiations by the nuclear decay and the subsequent atomic process were computed from the ENSDF using the computer code EDISTR04. EDISTR04 is an enhanced version of EDISTR used for assembling ICRP Publication 38 (ICRP38), and incorporates updates of atomic data and computation methods for calculating atomic radiations and spontaneous fission radiations. Quality assurance of the compiled data has been made by comparisons with various experimental data and decay databases prepared from different computer codes and data libraries. A package of the data files, called DECDC2 (Nuclear DECay Data for Dosimetry Calculation, Version 2), will succeed ICRP38 that has been used extensively in dose calculation and will be utilized in various fields. (author)

Medical radiation dosimetry with radiochromic film

International Nuclear Information System (INIS)

Butson, M.J.; Cancer Services, NSW; Cheung, T.; Yu, P.K.N.; Metcalfe, P.

2004-01-01

Full text: Photon, electron and proton radiation are used extensively for medical purposes in diagnostic and therapeutic procedures. Dosimetry of these radiation sources can be performed with radiochromic films, devices that have the ability to produce a permanent visible colour change upon irradiation. Within the last ten years, the use of radiochromic films has expanded rapidly in the medical world due to commercial products becoming more readily available, higher sensitivity films and technology advances in imaging which have allowed scientists to use two-dimensional dosimetry more accurately and inexpensively. Radiochromic film dosimeters are now available in formats, which have accurate dose measurement ranges from less than 1 Gy up to many kGy. A relatively energy independent dose response combined with automatic development of radiochromic film products has made these detectors most useful in medical radiation dosimetry. Copyright (2004) Australasian College of Physical Scientists and Engineers in Medicine

Personal dosimetry and information platforms

International Nuclear Information System (INIS)

Sanchez Hidalgo, M.; Galan Montenegro, P.; Bodineau Gil, C.; Hernandez Rodriguez, R.; Jimenez Nartin, A.; Cano Sanchez, J. J.

2011-01-01

One question often raised by the hospital personnel dosimetry is the high incidence in the no monthly turnover of dosimeters, which is currently a high number of administrative dose assignments. The high number of workers with personal dosimetry and in many cases, the dispersion of workplaces makes it impossible to personalized management. To make a more direct and personal, and transmit information quickly and with guaranteed reception, has developed and implemented a system of personalized dosimetric information through messaging Short Message Service (SMS) and access to the history of dosimetric dosimetric and management through web space Service Hospital Radio physics.

Radiation dosimetry and radiation biophysics

International Nuclear Information System (INIS)

Anon.

1981-01-01

Radiation dosimetry and radiation biophysics are two closely integrated programs whose joint purpose is to explore the connections between the primary physical events produced by radiation and their biological consequences in cellular systems. The radiation dosimetry program includes the theoretical description of primary events and their connection with the observable biological effects. This program also is concerned with the design and measurement of physical parameters used in theory or to support biological experiments. The radiation biophysics program tests and uses the theoretical developments for experimental design, and provides information for further theoretical development through experiments on cellular systems

Radiation dosimetry and radiation biophysics

International Nuclear Information System (INIS)

Anon.

1979-01-01

Radiation dosimetry and radiation biophysics are two closely integrated programs whose joint purpose is to explore the connections between the primary physical events produced by radiation and their biological consequences in cellular systems. The radiation dosimetry program includes the theoretical description of primary events and their connection with the observable biological effects. This program also is concerned with design and measurement of those physical parameters used in the theory or to support biological experiments. The radiation biophysics program tests and makes use of the theoretical developments for experimental design. Also, this program provides information for further theoretical development through experiments on cellular systems

Information from the Dosimetry Service

CERN Multimedia

2006-01-01

CERN Staff and Users can now consult their dose records for an individual or an organizational unit with HRT. Please see more information on our web page: http://cern.ch/rp-dosimetry. The Dosimetry Service is open every morning from 8.30 to 12.00 and is closed in the afternoons. We would like to remind you that dosimeters cannot be sent to customers by internal mail. Short-term dosimeters (VCT's) must always be returned to the Service after use and must not be left on the racks in the experimental areas or in the secretariats.

Sizewell B Power Station control dosimetry system

International Nuclear Information System (INIS)

Renn, G.

1995-01-01

Sizewell B Power Station is the first Pressurized Water Reactor (PWR) built in the UK for commercial electricity production. An effective control dosimetry system is a crucial tool, in allowing the station to assess its radiological performance against targets. This paper gives an overview of the control dosimetry system at Sizewell B and describes early operating experience with the system. (UK)

Retrospective dosimetry (or self dosimetry): Application to French Nuclear Power Plants

International Nuclear Information System (INIS)

Lloret, R.

1993-01-01

In this text we give the dosimetry principle on irradiated materials such baffle screw, pressure vessel and control element cans. This measure, made by gammametry, is based on the steel activation and comparison with calculated measures by Actige code. 4 figs., 6 refs

Advances on radiation protection dosimetry research, development and services at AEOI

International Nuclear Information System (INIS)

Sohrabi, M.

1993-01-01

Radiation dosimetry is the main counterpart of an effective national radiation protection program to protect radiation workers, public and the environment against harmful effects of radiation. Research and development on radiation dosimetry are of vital needs to support national dosimetry services. The National Radiation Protection Department (NRPD) of the Atomic Energy Organization of Iran (AEOI) being a National Authority on radiation protection is also responsible for radiation dosimetry research, development and services. Some highlights of such activities at NRPD are reviewed and discussed

Legal aspects of dosimetry

International Nuclear Information System (INIS)

Pomarola, H.

1976-01-01

The use of ionizing radiations is regulated in France in all fields of application. The main principles governing inspection activities in the food industry are outlined. Conventional preservation methods are mentioned, after which a discussion is devoted to the preservation of food products by irradiation treatment and the increasing importance given to this technique. Consumer protection automatically implies the obligatory use of dosimetry by inspection organisms if the irradiated merchandise is likely to serve for human or animal consumption. Irradiation treatment permits are granted in a context of specific statutory texts mentioned here. Supervision is constant, but always both realistic and flexible. Each aspect of this treatment is discussed in maximum detail if not quite exhaustively, with special emphasis on dosimetry as an indispensable safety factor [fr

Is intraoperative real-time dosimetry in prostate seed brachytherapy predictive of biochemical outcome?

Directory of Open Access Journals (Sweden)

Daniel Taussky

2017-06-01

Full Text Available Purpose : To analyze intraoperative (IO dosimetry using transrectal ultrasound (TRUS, performed before and after prostate low-dose-rate brachytherapy (LDR-BT, and compare it to dosimetry performed 30 days following the LDR-BT implant (Day 30. Material and methods : A total of 236 patients underwent prostate LDR-BT using 125 I that was performed with a three-dimensional TRUS-guided interactive inverse preplanning system (preimplant dosimetry. After the implant procedure, the TRUS was repeated in the operating room, and the dosimetry was recalculated (postimplant dosimetry and compared to dosimetry on Day 30 computed tomography (CT scans. Area under curve (AUC statistics was used for models predictive of dosimetric parameters at Day 30. Results : The median follow-up for patients without BF was 96 months, the 5-year and 8-year biochemical recurrence (BR-free rate was 96% and 90%, respectively. The postimplant median D 90 was 3.8 Gy lower (interquartile range [IQR], 12.4-0.9, and the V 100 only 1% less (IQR, 2.9-0.2% than the preimplant dosimetry. When comparing the postimplant and the Day 30 dosimetries, the postimplant median D 90 was 9.6 Gy higher (IQR [–] 9.5-30.3 Gy, and the V 100 was 3.2% greater (0.2-8.9% than Day 30 postimplant dosimetry. The variables that best predicted the D 90 of Day 30 was the postimplant D 90 (AUC = 0.62, p = 0.038. None of the analyzed values for IO or Day 30 dosimetry showed any predictive value for BR. Conclusions : Although improving the IO preimplant and postimplant dosimetry improved dosimetry on Day 30, the BR-free rate was not dependent on any dosimetric parameter. Unpredictable factors such as intraprostatic seed migration and IO factors, prevented the accurate prediction of Day 30 dosimetry.

Dosimetry services for internal and external radiation sources

International Nuclear Information System (INIS)

1988-01-01

The Canadian Atomic Energy Control Board (AECB) sets radiation dose limits for the operation of nuclear facilities and the possession of prescribed substances within Canada. To administer these regulations the AECB must be satisfied that the dosimetry services used by a licensee meet adequate standards. Licensees are required to use the Occupational Dosimetry Service operated by the Bureau of Radiation and Medical Devices, Department of National Health and Welfare (BRMD) to determine doses from external sources of radiation, except where a detailed rationale is given for using another service. No national dosimetry service exists for internal sources of radiation. Licensees who operate or use a dosimetry service other than the BRMD must provide the AECB with evidence of the competence of the staff and adequacy of the equipment, techniques and procedures; provide the AECB with evidence that a quality assurance program has been implemented; and send individual dose or exposure data to the National Dose Registry. (L.L.)

High Energy Electron Dosimetry by Alanine/ESR Spectroscopy

International Nuclear Information System (INIS)

Chu, Sung Sil

1989-01-01

Dosimetry based on electron spin resonance(ESR) analysis of radiation induced free radicals in amino acids is relevant to biological dosimetry applications. Alanine detectors are without walls and are tissue equivalent. Therefore, alanine ESR dosimetry looks promising for use in the therapy level. The dose range of the alanine/ESR dosimetry system can be extended down to l Gy. In a water phantom the absorbed dose of electrons generated by a medical linear accelerator of different initial energies (6-21 MeV) and therapeutic dose levels(1-60 Gy) was measured. Furthermore, depth dose measurements carried out with alanine dosimeters were compared with ionization chamber measurements. As the results, the measured absorbed doses for shallow depth of initial electron energies above 15 MeV were higher by 2-5% than those calculated by nominal energy CE factors. This seems to be caused by low energy scattered beams generated from the scattering foil and electron cones of beam projecting device in medical linear accelerator

Dosimetry applied to radiology and radiotherapy

International Nuclear Information System (INIS)

Yoshimura, Elisabeth Mateus

2010-01-01

Full text. The uses of ionizing radiation in medicine are increasing worldwide, and the population doses increase as well. The actual radiation protection philosophy is based on the balance of risks and benefits related to the practices, and patient dosimetry has an important role in the implementation of this point of view. In radiology the goal is to obtain an image with diagnostic quality with the minimum patient dose. In modern Radiotherapy the cure indexes are higher, giving rise to longer survival times to the patients. Dosimetry in radiotherapy helps the treatment planning systems to get a better protection to critical organs, with higher doses to the tumor, with a guarantee of better life quality to the patient. We will talk about the new trends in dosimetry of medical procedures, including experimental techniques and calculation tools developed to increase reliability and precision of dose determination. In radiology the main concerns of dosimetry are: the transition from film- radiography to digital image, the pediatric patient doses, and the choice of dosimetric quantities to quantify fluoroscopy and tomography patient doses. As far as Radiotherapy is concerned, there is a search for good experimental techniques to quantify doses to tissues adjacent to the target volumes in patients treated with new radiotherapy techniques, as IMRT and heavy particle therapy. (author)

Guide for selection and calibration of dosimetry systems for radiation processing

International Nuclear Information System (INIS)

2002-01-01

This guide covers the basis for selecting and calibrating dosimetry systems used to measure absorbed dose in gamma ray or X-ray fields and in electron beams used for radiation processing. It discusses the types of dosimetry systems that may be employed during calibration or on a routine basis as part of quality assurance in commercial radiation processing of products. This guide also discusses interpretation of absorbed dose and briefly outlines measurements of the uncertainties associated with the dosimetry. The details of the calibration of the analytical instrumentation are addressed in individual dosimetry system standard practices. The absorbed-dose range covered is up to 1 MGy (100 Mrad). Source energies covered are from 0.1 to 50 MeV photons and electrons. This guide should be used along with standard practices and guides for specific dosimetry systems and applications covered in other standards. Dosimetry for radiation processing with neutrons or heavy charged particles is not covered in this guide

Current internal-dosimetry practices at US Department of Energy facilities

International Nuclear Information System (INIS)

Traub, R.J.; Murphy, B.L.; Selby, J.M.; Vallario, E.J.

1985-04-01

The internal dosimetry practice at DOE facilities were characterized. The purpose was to determine the size of the facilities' internal dosimetry programs, the uniformity of the programs among the facilities, and the areas of greatest concern to health physicists in providing and reporting accurate estimates of internal radiation dose and in meeting proposed changes in internal dosimetry. The differences among the internal-dosimetry programs are related to the radioelements in use at each facility and, to some extent, the number of workers at each facility. The differences include different frequencies in the use of quality control samples, different minimum detection levels, different methods of recording radionuclides, different amounts of data recorded in the permanent record, and apparent differences in modeling the metabolism of radionuclides within the body. Recommendations for improving internal-dosimetry practices include studying the relationship between air-monitoring/survey readings and bioassay data, establishing uniform methods for recording bioassay results, developing more sensitive direct-bioassay procedures, establishing a mechanism for sharing information on internal-dosimetry procedures among DOE facilities, and developing mathematical models and interactive computer codes that can help quantify the uptake of radioactive materials and predict their distribution in the body. 19 refs., 8 tabs

Dosimetry in Interventional Radiology - Effective Dose Estimation

International Nuclear Information System (INIS)

Miljanic, S.; Buls, N.; Clerinx, P.; Jarvinen, H.; Nikodemova, D.; Ranogajec-Komor, M; D'Errico, F.

2008-01-01

Interventional radiological procedures can lead to significant radiation doses to patients and to staff members. In order to evaluate the personal doses with respect to the regulatory dose limits, doses measured by dosimeters have to be converted to effective doses (E). Measurement of personal dose equivalent Hp(10) using a single unshielded dosimeter above the lead apron can lead to significant overestimation of the effective dose, while the measurement with dosimeter under the apron can lead to underestimation. To improve the accuracy, measurements with two dosimeters, one above and the other under the apron have been suggested ( d ouble dosimetry ) . The ICRP has recommended that interventional radiology departments develop a policy that staff should wear two dosimeters. The aim of this study was to review the double dosimetry algorithms for the calculation of effective dose in high dose interventional radiology procedures. The results will be used to develop general guidelines for personal dosimetry in interventional radiology procedures. This work has been carried out by Working Group 9 (Radiation protection dosimetry of medical staff) of the CONRAD project, which is a Coordination Action supported by the European Commission within its 6th Framework Program.(author)

Advances in kilovoltage x-ray beam dosimetry

Science.gov (United States)

Hill, Robin; Healy, Brendan; Holloway, Lois; Kuncic, Zdenka; Thwaites, David; Baldock, Clive

2014-03-01

This topical review provides an up-to-date overview of the theoretical and practical aspects of therapeutic kilovoltage x-ray beam dosimetry. Kilovoltage x-ray beams have the property that the maximum dose occurs very close to the surface and thus, they are predominantly used in the treatment of skin cancers but also have applications for the treatment of other cancers. In addition, kilovoltage x-ray beams are used in intra operative units, within animal irradiators and in on-board imagers on linear accelerators and kilovoltage dosimetry is important in these applications as well. This review covers both reference and relative dosimetry of kilovoltage x-ray beams and provides recommendations for clinical measurements based on the literature to date. In particular, practical aspects for the selection of dosimeter and phantom material are reviewed to provide suitable advice for medical physicists. An overview is also presented of dosimeters other than ionization chambers which can be used for both relative and in vivo dosimetry. Finally, issues related to the treatment planning and the use of Monte Carlo codes for solving radiation transport problems in kilovoltage x-ray beams are presented.