present status and future developments in nuclear data for shielding ...

PRESENT STATUS AND FUTURE DEVELOPMENTS IN NUCLEAR DATA FOR SHIELDING APPLICATIONS WITH SPECIAL REFERENCE TO THE SCENE IN INDIA S. Ganesan Neutron Physics Division, Purnima Laboratories, Bhabha Atomic Research Centre, Trombay, Mumbai - 400 085

1.

INTRODUCTION:

This paper attempts to look at some aspects of the current international status of nuclear data for shielding applications ~ith special reference to the scene in India. The uses of radiation in energy or non-energy applications require appropriate shielding. India has her original three stage Atomic Energy Programme involving Pressurized Heavy Water Reactors, Fast Reactors and utilization of thorium in advanced thermal and fast reactors covering both the front and back end of the nuclear fuel cycle. The generation of databases and their proper use in design calculations forms an essential and an integral part of any scientifically founded design activity. In each of the design process, the extent of agreement between experiment and calculation for an integral parameter is influenced not only by the quality of the nuclear data input in -the design calculatioqs but also by the combined effect of the physical and numerical approximations in modeling the system and tolerances in fabrication of the engineering components. The basic requirements of infrastructure to perform accurate radiation shielding design studies, in any given radiation environment, therefore, naturally include the use of the most modern nuclear data in a detailed manner to describe the basic neutron interactions, photon production and photon-atom interaction cross sections for all the elements/isotopes present in the system. It is our desire that our ability to perform radiation shielding studies should be on par with the advanced countries. While the Indian capabilities are presently very high, the subject of modern radiation shielding should be purtured and supported. This paper attempts to put the matter into proper perspective taking our vision for the next century into. account. InVernationally, until the early sixties, the shield design wethods were based on simple empirical procedures by which performance data for a known reactor and shield were used to predict the results for a new configuration (Ref.l). The modern shielding methods were under continuous development during the last three decades aided by the improvements in basic nuclear data, development of more accurate algorithms, and advances in computer software and hardware. These methods use rigorous nu~erical methods-as rigorous as permissible by th~ available computers- to solve the coupled neutron-photon transport equation using basic differential nucl.ear data as accurately as possibl~.

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In the opinion of the author, up-grading our software capabilities includes upgrading the quality in the associated microscopic nuclear data which in turn will help us to leap-frog in innovative studies with a firm scientific foundation. 2.

PHILOSOPHY AND EVOLUTION OF 'THE RADIATION SHIELDING STUDIES

The advanced shielding methodology is essentially a fundamental approach without depending much on pure empiricism. It originated largely as a result of demanding shielding requirements with aircraft nuclear propulsion systems (ANP) , the space nuclear programs (SNAP) and liquid metal fast reactor program (FBR). The current Lrrt.e.rnat.Lona I trend is to follow the advanced shielding methodology also in other areas such as nuclear medicine and health, fusion R'& 0, conceptual studies of advanced emerging nuclear energy systems etc. Internationally, the evolution of the sUbject of radiation shielding has inclUded five basic technologies as pointed out 24 years ago in Ref'. 1: 1. 2. 3. 4. 5.

Cross section measurements Cross-section evaluations Cross section processing Integral experiments Radiation transport calculations.

All these five basic technologies are well known and one or more of these steps are always encountered in the literature (Refs. 1 40). We will make some remarks on these five basic technologies to put the matter into proper perspective with respect to our own strength, limitations and approach. The process of getting the working libraries for design calculations requires an iterative sequence of events to yield a quality assured transport cross section library. See Figs. 1-3 for some illustrations. 3.

REMARKS ON PRODUCTION OF CALCULATIONS

WORKING LIBRARIES FOR

APPLICATION

The international nuclear data community has, over several decades, efficiently provided state-of-the-art experimental and theoretical data which were needed in applications (See Refs. 2-6). Several evaluated nuclear data files providing cross sections for neutron induced reactions and photon productions in various important isotopes/elements in the periodic table are available (Refs 2-6). The first two steps, "cross section measurements" and "cross-section evaluations" lead to the creation of the basic evaluated nuclear .data files in the well known ENDF/B format. Today, for a scientist looking freshly at the status of nuclear data, it'is seen that, taking any nuclide as example, say Fe-56, one can get the neutron interaction cross section data from various countries in ENDF/B format, for example: E N 0 F / B-VI.2 (USA), JEF-2.2 (Europe), BROND2.2 (Russia), JENOL-3.2( Japan), CENDL-2 (China) etc. These files have been compiled by

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international nuclear reaction data centres; and have recently become available (Refs. 2-6). These recent basic evaluated data files can be down-loaded over the internet from the Nuclear Data section of the IAEA. Presently, with so many data files for the same isotope for the same physical quantity being available from various countries, it ~3 a complex task for the user team to arrive at a decision as to which data file is better for a specific application. The author sincerely wishes that over the next few decades, many data files existing to provide data for the same physical quantity will converge and we will hopefully have one unique numerical data file along with uncertainties internationally accepted for use in applications. We stress that the recently released basic evaluated nuclear data files such as ENDF/B-VI, JENDL-3.2, BROND-2.2 and CENDL-2.2 etc., are not directly used as input to neutronics or other applied calculations but are first converted to pre-processed files which are post-processed into multigroup files which are then cast into specially formatted working libraries that are compatible with neutronic codes. Both the developed and developing countries are finding it difficult to sustain and fund adequately the nuclear data processing (Refs. 7-12) activities. The processing tasks demand specialized expertise using computer resources committed over a long period of time (several years) to execute the tasks. See the flowcharts in Fig. 1 and Fig. 3. There has been no active common forum to bring the scientists, in ·particular in developing countries working in this area together. The field of study of nuclear data processing and preparation of processed nuclear data libraries form the connecting bridge between basic evaluated nuclear data files and application calculations. The sUbject involves development and validation of computer software using knowledge of ENDF/B formats and conventions, numerical reconstruction of resonance line shapes, calculations of Doppler broadening, thermalization effects, self-shielding factors in resolved and unresolved resonance regions, transfer matrices for various Legendre orders etc. The nuclear data processing requirement must satisfy the very important requirement that errors due to processing of the basic data does not introduce unacceptable errors in the processed data Which are the results of processing (Ref.ll). The accuracy of the processed data should correctly reflect the quality of the basic evaluated data. See for example Ref. 40. The task of updating the mUltigroup library for shielding applications involves the tasks of processing the basic neutron cross section files ENDF/B-VI.x, JENDL-3.x, BROND-2.x and CENDL-2.x etc. ,(where x is the revision/update number) using the nuclear data processing code system such the NJOY code system (Ref. 7), taking care to ensure that the quality of the working library correctly reflects the quality of the basic evaluated data file from which the working library is derived, keeping the conventions and definitions of the group constants as. expected by the shielding design system, without introducing unacceptable errors in processing. . There is a

basic need to bring

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out a handbook which

will

provide comparison of evaluated data from various sources (examples: JENDL3.x, BROND-2.x, ENDF/B-VI.x, JEF-2.x, CENDL-2.x etc.) with experimental data from EXFOR to highlight existing convergence or non-convergence of basic evaluated data and spread of available experimental data with uncertainties for all isotopes and elements for which evaluations are available. Such a handbook will help every scientist and engineer to appreciate the existing gaps and spread in measurements and in the data of evaluations made by different countries in relation to the experimental uncertainties for each reaction cross section for each isotope/element. In the resolved resonance region such a comparison is presently of limited scope in most of the cases of isotopes and elements as point experimental cross section in the resolved resonance region have not been traditionally compiled into EXFOR thus far except in the case of some isotopes like Fe-56. The compiled data in the resonance region, traditionally into EXFOR have been in the form of resonance parameters to save disk space thus forcing the present, day user to assume that the resolved resonance parameters represent exactly the experimental data in the resolved resonance region. 4.e

REMARKS ON THE NUCLEAR DATA PROCESSING ACTIVITY

It should be noted that the size of the databases have become very large and the representation of the data in terms of physics has become increasingly complex leading to the requirement of specialized skills in electronic access, storage, retrieval, management and processing. The nuclear data programmes in IGC~~ (Ref. 8 and Ref. 9) and in BARC (Ref. 9) are oriented towards developing a comprehensive cap~bility to make use of computerized nuclear data files already freely available from international sources such as the IAEA. The third step, "cross section processing" mentioned in Section 2 is needed as an interface between the ENDF/B formatted files and the application codes as the ENDF/B data files which are freely available are not usable directly as input to core design and shielding or other applied calculations. In this connection some remarks on the use of the NJOY code system (Ref.7) are appropriate: The NJOY code system has been widely used around the world in many leading laboratories and has served successfully as a general- purpose link between ENDF-formatted evaluated nuclear data files and important applications such as shield design, thermal reactor core design, fusion reactor blanket design, radiotherapy facility design and many others. The NJOY system has been under continuous development in Los Alamos National Laboratory 'since 1974. During the last 22 years it has gone through a number of revisions, in response to the development of the new ENDF formats f in order to add new processing capabilities, in response to changing computer systems and in order to fix errors. It should be noted that the NJOY code system is, by necessity, complex and large with over 80000 lines of Fortran instructions. It has ~ comprehensive approach to interfacing ENDF/B formatted data to application codes. The task of int~rfacing of electronic databases to application software to produce user-friendly application dependent libraries has assumed strategic importance. Considering that the NJOY code is not available freely to India,

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efforts that are being made in developing i~digenous processing codes should be nurtured, supported and conti'nued. 5.

INFORMATION ON POST-PROCESSING OF MATXS FILES:

In Fig. 2 presented are the flowcharts illustrating the various steps in nuclear data processing. One can also use the AMPX route (Ref. 12) starting from the GENDF file to produce ANISN compatible working libraries. Some statements on the code, TRANSX-2 (Ref. 10), are made below as the NJOY code system (Ref. 7) has not been designed to produce the final transport cross section tables for use by the transport codes such as ANISN, DOT etc., for use in application calculations. The MATXS files are produced by MATXSR module of NJOY. The MATXSR module of NJOY reformats mutigroup constants from the GENDF tape (i.e the output file of GROUPR for neutron interaction cross sections and photon production cross sections or the output file of GAMINR for photon atomic interaction cross sections) into the MATXS interface format. The MATXS file has a very general organization to hold arbitrary vectors and matrices. The file is first-divided into "data types" such as neutron scattering, photon prod.!1ction, gamma scattering, and neutron thermal data. Each data type is assigned a name (NSCAT, NG, GSCAT, NTHERM).Datatypes are distinguished by the choice of incident and secondary group structures. Each data type is divided into materials specified by nuclide, temperature, and background cross section. Each material is further subdivided into "vector partials" and "matrix partials". These reaction partials are labeled with Hollerith names so there is no limit on the quantities that can be stored. MATXSR reads cross sections from the GENDF tape, ass1gns the Hollerith names, and packs the cross sections into MATXS format. The code TRANSX 2.0 (Ref. 10), for instance, serves to interface MATXS cross section libraries to nuclear transport codes such as.ANISN. TRANSX reads nuclear data from a library in MATXS format and produces transport tables compatible with many discreteordin~tes (SN) and diffusion codes. The FENDr. mUltigroup library (Ref. 23) may be post- processed using TRANSX to produce tables for neutron, photon or coupled transport for specific application calculations. It has been noted (Ref. 11 and Ref. 12) that the process of preparing the transport cross section tables in a format compatible with Monte Carlo or discrete ordinates codes is a difficult technical task involving an iterative sequence of events to yield a Quality Assured transport cross-section library. Some illustration is provided in Fig. 3 following J. E. White (Ref. 12) 6.

ON COVARIANCE ERROR

MP~TRIX:

Some remarks on uncertainty information is needed; the basic nuclear dat~ have their associated unc~rtainties which can be expressed in the form of covariance error matrices. However, the ENDF/B files, except for the dosimetry reactions do not have the covariance data. This sUbject is still evolving (Refs. 13-16). Reactor physicists,· for example, the French (Ref.15) have

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elaborated covariance data on the basis of personal jUdgement and systematic studies to define the conditions for a statistical adjustment procedure to obtain adjusted multigroup cross section sets thus making. maximum practical use of integral data and differential data. The interested reader is referred to papers in Ref.15 for more details. 70

INFORMATION ON PHOTON-ATOM INTERACTION COUPLED PHOTON-ELECTRON CALCULATIONS:

DATA AND

DATA FOR

A brief description on the availability of ENDL Family of Data (Refs. 17-19) for Coupled Photon-Electron Calculations follows: A package of three data libraries for interaction of atoms,Evaluated Atomic Data Library (EADL) (Ref.17), Evaluated Electron Data Library (EEDL) (Ref.18) and Evaluated Photon Data Library (EPDL) (Ref.19) has been announced recently (Report UCRL-50400 series). The files totalling 44 MB in size are available from the IAEA Nuclear Data Section. Used in combination, these three data libraries can be used to completely describe coupled photon-electron transport. All photon induced electron production through pair, incoherent scattering, primary electrons from photo ionization, as well as secondary electrons from atomic relaxation can be accounted for. Similarly all electron induced photon production through positron annihilation, bremsstrahlung, as well as secondary photons from atomic relaxation can be accounted for. Following the literature (Refo 19), reproduced below are a few remarks on the ENDF/B-VI photon-atom interaction data to place before the user the improvements in the new version as these data are used in the generation of modern coupled neutron-photon cross section library. The Livermore Evaluated Photon Data Library (EPDL), also adopted for ENDF/B-VI contains photo-atomic interaction data library consisting of pair production cros~-sections, photoelectric cross-sections, coherent scattering cross sections, atomic form factors and other data for all elements from Z=l· to 100. The new ENDF/B-VI photon-atom interaction library covers an extended element and energy range and is the first major extension of the ENDF/B photon interaction library since the inception of ENDF/B. Traditionally, the ENDF/B database for photon-atom interactions contained cross sections for coherent and incoherent scattering, pair production as well as photoelectric absorption. In addition it contained form factors and scattering functions to describe the angular distributions of coherent and incoherently scattered photons. In some applications when needed, these traditional data are not sufficient to .uniquely define the emission of secondary photons following photoelectric effects, e.g., fluorescence. ~he assumption in the past has been that when a photoelectric event takes place all of the energy of the incident photons is deposited at the point of interaction. In fact, in the case of photons with energies near the K photoelectric edge of lead almost 88% of the energy will be re-radiated as fluorescence

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x-rays. The trad1tional data also did not include the effect of anomalous scattering on coherent scattering. Including this effect predicts a coherent scattering cross-section which. approaches zero at low energy, as opposed to the constant low energy llmit predicted simply by using form factors. Lastly the traditional data did not differentiate between pair and triplet production. The traditional assumption concerning energy deposition at a point breaks down when one requires more and more spatial effects in the answer. For example, in medical applications one should not expect to be able to use the traditional method to define the image which. appear on a thin x-ray film; in this case the results can be dominated by the transport of electrons into, out of and within the volume. Here the size scale is too small to allow the traditional method to be used. There are industrial applications where the answers are extremely sensitive to the distribution of energy deposited by electrons, as in the production of microchips for computers or for that matter the deposition of energy or charge in el~ctronic equipment (including that used to control nuclear reactors). 8.

BENCHMARK EXPERIMENTS:

A variety of available integrql benchmark experiments (See for example, Table 1) are analyzed and with the obtained results, the user qualifies the working libraries for neutronicsand shielding applications. Before we go to the current state of art, some general remarks (Ref. 1) on the evolution of shielding experiments from a historical point of view will be desirable. The evolution of shielding experiments (See for example, Ref.l) has been closely coupled with the development of neutron sources and detectors and to changing requirements for testing transport calculations. Reactor sources, accelerator sources and pulsed electron accelerators have been used. The earliest experiments consisted of dose measurements made behind shields. See Fig. 4a. In this case, the measurement involved integration over the energy and angular distributions of the source neutrons, over the shield volume, and over the energy and angular distributions of the leakage neutrons. A second generation experiment (Fig. 4a) resulted when the integration over the energy distribution of leakage neutrons was eliminated by measuring this energy spectrum by using a differential proton-recoil spectrometer. The Time of Flight experiments have been performed at Lawrence Livermore to measure leakage neutrons from spherical shells of various shielding materials surrounding a small, pulsed 14-MeV neutron source. In most of these second generation experiments . involving differential proton recoil and TOF spectrometers, integration over the angular distribution of the leakage neutrons has also been eliminated by the use of collimation. In the third generation shielding experiment (Fig. 4a and Fig.4b), integration over the energy and angular distributions of the source neutrons was effectively removed for source neutrons above 2 MeV. This was accomplished using a Linac Pulsed neutron source and, in contrast to earlier pulsed neutron integral

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shielding experiments, the shield under study was j located at the detector end of the flight path instead of at the source enti. In the experiment by Harris and Kendrick (Ref.1), a Linac was used to produce intense 50-nsec pulses of electrons. A TalBe target converted the electrons into photo-neutron pulses essentially of the same duration. These neutrons traversed an evacuated 50 metre long flight path to strike the concrete slabs. The backscatter shield shown in the figure prevented neutrons backscattered by the concrete from reaching the detector. A 13 x 13 cm cylindrical NE-213 scintillator detected both fast neutrons (E > 1 MeV) and secondary gamma rays produced in the concrete by (n,xg) and (n,g) reactions. Pulse-shape discrimination was used to separate neutron and gamma-ray counts. Three parameters, time, pulse height and pulse shape, were recorded for each detection event using an online computer. As part of data reduction, neutron and gamma-ray counts detected in the 0.9- and 2.9 microseconds time range (corresponding to incident neutron energies from 2 to 20 MeV) were sorted into 18 and 9 time bins, respectively. Gamma-ray counts detected at longer times were sorted into 9 time bins. Using measured neutron and gamma ray response matrices, the pulse height spectrum for each time bin was unfolded using the FERDOR code to obtain time-dependent neutron and gamma~ray energy' spectra. Benchmark tests of gamma-ray production cross section in evaluated nuclear data libraries are highly required to verify the data experimentally. For example, in order to provide benchmark data related to gamma-ray production cross section, two series of integral experiment have been conducted (Ref.20) by utilizing D-T neutron source facilities; FNS in JAERI' and OKTAVIAN in Osaka University. In the FNS experiment, gamma-ray spectra and· gamma-ray heating rates were measured at various positions in Fe, Cu, Wand type 316 stainless steel assemblies as a function of depth in the assemblies. Neutron spectra and reaction rates were also measured in the experiment. Gamma-ray data not only for threshold reactions by D-T neutrons but also for radiative neutron capture reactions by low energy neutrons can be verified with the experimental data. In the OKTAVIAN experiment, gamma-ray spectra leaking from spherical piles made of LiF, CF2, AI, Si, Ti, Cr, Mn, Co, Cu, Nb, Mo., Wand Pb were measured along with leakage neutron spectra. This experiment focusses on gamma-rays produced mostly by threshold reactions with D-T neutrons. The compilation of experimental and 'calculational benchmarks (Refs. 21-23) is being made now-a-days in electronic format. The compilation "SINBAD". (Shielding, Integral, Benchmark, Archive and Database) is one such example (Ref. 22). The SINBAD is expected to be an user-friendly compilation of integral benchmarks for radiation shielding for all types of fission, fusion and accelerator systems. H. T. Hunter reported (Ref. 22) that this electronic database has already a compilation of 68 shielding benchmarks excluding accelerator benchmarks in this counting. The SINBAD under deve~opment is expected to provide accurate, fast, space efficient and complete benchmark information. The benchmark description has, in addition to a

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general description, details of radiation source used in the experiment, geometry, materials, detectors, integral results, error analyses and response functions. The electronic document contains formatted text and equations, with graphics stored as a popular desktop pUblisher. Quick review of benchmarks via prepared abstracts is 'available. The computerized search by the user can be done at various levels. One can quickly search experimental facility of interest, experiment type (penetration, streaming, skyshine, energy deposition or radiation damage, induced activity in shield etc), source spectrum, source particles (at present neutron, gamma, proton and electron), primary material, range of thickness, measured particle and data type. The electronic compilation of the IAEA benchmarks (Ref. 21) have also been included in the database being developed by H. T. Hunter (Ref .22).

9.

REMARKS ON UPDATING COUPLED NEUTRON GAMMA MULTIGROUP LIBRARY FOR SHIELDING ANALYSES OF FAST REACTORS:

The excellent work being done (Ref.8) at Kalpakkam is designed to take care of needs of the Indian Fast Reactor Programme. R. Indira (Ref. 25) has reported the use of DLC-37 library (100 neutron groups and 21 gamma groups both derived from ENDF/B-IV) in shielding studies with the ANISN code in S8-P1 approximation. In addition to using such libraries of DLC-37 type in transport calculations, efforts are underway (Ref. 8) to produce indigenous working nuclear data libraries by processing basic data files in ENDF/B format. The data processing system in Kalpakkam starts from basic data files and makes use of Red Cullen's pre-processing codes (Ref.26). The neutron multigroup library has' been generated from JENDL-2 and has been validated against pUblished data of critical experiments. The studies performed at Kalpakkam clearly established the superiority JENDL2 derived IGC cross section set for the design of larger fast reactors as compared to the old Cadarache cross section set used successfully in the design and operation of Fast Breeder Test Reactor. The generation and integral validation of an indigenous coupled neutron-photon library for applications to fast reactors has been reported (Ref. 36). In Ref. 36, the basic data of neutron interactions and neutron induced photon production cross sections derived from ENDF/B-IV and the photon-atom interaction cross sections from DLC-99 have been processed and integrally tested successfully. See the flow chart ,in Fig. 5.

10.

REMARKS ON PC VERSION

o~

SHIELDING CODES:

The currently available personal computers in the market such as 486 or Pentium based Personal Computers can be used for most of the shielding calculations if the specifications of the PC are properly chosen for its speed, memory, hard disk space and Fortran compiler. Interesting experience in using a 486 PC for 2-D transport calculations by Keshavamurthy and Indira (Ref. 32) at Kalpakkam throw some light on the use of PC for such calculations. In the late eighties, the porting of codes to the PC were often hampered, however, by Fortran compiler limitations. Since June 1990, however, success has been reported with the use

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of Lahey F77L-EM/32 compiler starting with version 3.0 and currently at verSlon 5.2. This Lahey compiler produces 32 bit code that runs only on PCs, accepts large source files, and comes with a DOS extender. It has been reported (Ref. 33) that the Lahey compiler based executable files run twice as fast as the one based on the Microsoft Fortran compiler. The radiation Shielding Information centre, USA distributes pc versions of several codes, for example, the NJOY code system, the MCNP code system, the ANISN code etc. The current trend is to make the reading of the output files very fast and efficient by making use of graphical post-processors created for each of the application codes. See for example, Ref. 34. The plotting routines in the NJOY code in its latest version (Ref. 7) has postscript capabilities to ensure portability. 11.

ON THE SHIELDING ANALYSIS DONE FOR THE MONJU FAST REACTOR

K. Chatani et al., (Ref. 35) reported a brief description of the shielding analysis for their MONJU fast reactor (714 MWth/ 280 MWe). The flow diagram shown in Fig. 6 and taken from Ref. 35 is the standard shielding analysis system used in Japan in 1992. It should be noted that the ANISN/DOT package has become the universal tool for the shield design. In the coming decades, the impact of parallel architecture and improvements in computing performance will make the Monte Carlo code such as the MCNP the universal tool for all shielding calculations. It may be noted that the benchmark specialists (Ref. 24) recommend analysis of deep penetration problem$ by Monte Carlo calculations with continuously represented cross section data than by discrete ordinates calculations with mUlti-group data. 12.

REMARKS ON NUCLEAR DATA FOR FUSION APPLICATIONS:

It should be noted that the FENDL system (Ref. 23, 27-31) of nuclear data goes beyond the scope of the classical data operations in a sense that it also includes processed mUlti-group libraries FENDL/MG in 175 VITAMIN-J group structure derived from the basic files for use by discrete ordinate codes and FENDL/MC in pointwise format compatible with the Monte Carlo code MCNP. The processed libraries were generated by R. E. MacFarlane of Los Alamos National Laboratory (LASL) using the comprehensive processing code NJOY developed over the last 22 years at LASL. Some of the interesting experiences in QA studies, while preparing the processed libraries, that surfaced in resonance reconstruction, Doppler broadening, group averaging etc were very educative (Refs. 23, 27-31). The experiences in creating the FENDL brought out the fact that processability of the basic data file is a pre-requisite for the user (ITER and other fusion designers in this case) and nuclear data processing steps are an integral part of QA checks on the basic file and the data services for the actual designer who sees only the processed library as data in the design calculations. 13.

LESSONS FROM THE ANALYSIS OF FUSION SHIELDING BENqHMARKS: U. Fischer from

Forschungszentrum, Karlsruhe (Ref.

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24) has

recently presented a summary report of the results of analy~es of a variety of existing 14 MeV neutron benchmark experiments contributed by various laboratories (Japan, Europe, U.S.A., Russia) and compiled by the IAEA (Ref. 21). The benchmark analyses have been performed for the FENDL-l data file in an international effort. Benchmark calculations have been performed with the working libraries FENDL/MG-l.0 for discrete ordinates calculations and FENDL/MC-l.0 for Monte Carlo calculations with the MCNP code. It was pointed out that the transport problems involving neutron thermalization cannot be properly accounted for in discrete ordinates calculations with multi-group data in the VITAMIN-J group structure which was understood due to missing up-scattering capabilities while splitting the thermal energy range into two groups. In the case of use of internationally available multi-group data, the user has to be aware of the following consistency problems, such as the problem of the required resolution of the energy group structure, spatially varying weighting functions, the resonance self-shielding and the associated mUlti-group data processing. In addition, the discrete ordinates technique is deficient in simulating the strongly forward peaked neutron transport in a deep penetration problem. It was pointed out (Ref. 24) that although the two approaches gave the same results for analyses of spherical shell and slab transmission experiments, care has to be taken in applying the available processed multigroup libraries to neutronics and shielding problems where it may not be appropriate. It was noted that as a result of the detailed analysis of benchmarks that the FENDL-l data library has achieved a high confidence level with few exceptions. The benchmark specialists (Ref.24) point out that deep penetration problems can be better described by Monte Carlo calculations with continuously represented cross section data than by discrete ordinates calculations with mUlti-group data. 14.

REMARKS ON SOME IMPROVEMENTS IN BASIC DATA

In Ref. 37 and Ref. 38, some discussions on the improvements that have been seen in the results of analyses of integral shielding experiments have been presented.. The nuclear data of most of the shielding materials in ENDF/B-VI performed as good as in ENDF/B-V. A dramatic improvement was seen in characterizing the pressure vessel dosimetry in Light water Reactors facilitated by the use of new data for Fe-56 and B-ll from ENDF/B-VI. Generally the new data is superior to the use of older data if correct transport analyses is carried out. Kawai et al., report the applicability of iron data in JENDL-3 in deep penetration problems. More discussions on the assessment of the improvements in data have been presented in the literature. See for example, Refs. 2, 16, 23 and 24. 15.

CONCLUSIONS AND RECOMMENDATIONS:

We have looked at some status of nuclear data for

aspects of the current international shielding applications with special

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reference to the scene in India. The advanced shielding methodology is essentially a fundamental approach without depending much on pure empiricism. Our national efforts start from "cross sec1:ion processing" mentioned in section 2 since the ENDF/B data files which are freely available are not usable directly as input to core design and shielding or other applied calculations. Efforts that are being made in developing indigenous processing codes should be nurtured, supported and continued. Analyses of a variety of available integral benchmark experiments should continue. Efforts should be initiated to perform our own shielding benchmark experiments. Pentium based Personal Computers can be used for most of the shielding calculations if the specifications of the PC are properly chosen for its speed, memory, hard disk space and Fortran compiler. Although the ANISN/DOT package has become the universal tool for the shield design, in the coming decades, the impact of parallel architecture and improvements in computing perform~nce will make the Monte Carlo code such as the MCNP the universal tool for all shielding calculations. India's past credibility and future potential should put us on par with the advanced countries in the years to come. For instance, the calculational tools similar to those used in the shielding analysis (Ref. 35) for the MONJU fast reactor (714 MWth/ 280 MWe) which is the standard shielding analysis system in Japan since 1992 should be achievable by us. REFERENCES 1.

F. R.. Mynatt and M. L. Gri tzner, "Fast Reactor Shielding Methods Development'" pp. 480-498, in Book 1, NATIONAL TOPICAL MEETING ON NEW DEVELOPMENTS IN REACTOR PHYSICS AND SHIELDING, Sept. 12-15, 1972, Kiamesha Lake, NY,ONF-720901, US Atomic Energy Commission, (1972)

2.

Papers presented at the International Conference on Nuclear Data for Science and Technology, Gatlinburg, Tennessee, May 9-13, 1994, J.K. Dickens (Editor), American Nuclear society (1994).

3.

C. L. Dunford and T. W. Burrows, "Online Nuclear data Service'" IAE..~-NDS-150 (1995) Nuclear Data section, International Atomic Energy Agency, vienna,Austria.

4.

H.D. Lemmel, "Index of Data Libraries available on magnetic tape from the IAEA Nuclear Data Section," IAEA-NDS-7 (1995).

5.

"National Nuclear Data Needs of tpe 1990's. "A report by the Nuclear science Advisory Committee of the U.S. DOE. Unpublished 1994. See also, "A strategic View on Nuclear Data Needs," Report by the NEA secretariat, OECD, Paris, september 1993.

6.

S. Ganesan, "IAEA Nuclear data services," in A. Gandini, S. Ganesan and J. J. Schmidt (Editors), "Proceedings of the Workshop on NUCLEAR REACTORS: PHYSICS, DESIGN AND SAFETY, ICTP, Trieste, Italy 11 April-13 May 1994," World Scientific (1995).

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7.

R.E. MacFarlane, "NJOY94, A Code System for producing Pointwise and Multigroup Neutron and Photon Cross Sections from ENDF/B Evaluated Data," PSR-171, Radiation Shielding Information Centre, USA, 1994.

8.

V. Gopalakrishnan (Editor), "Activity Report of Reactor physics Division-1994," Report IGC-165, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu, India (1995).

9.

S. Ganesan (Compiler), "Progress Report on Nuclear Data Activities in India for the Period July 1992 to March 1995," BARC/1995/E/005(1995), Bhabha Atomic Research Centre, Bombay

10.

R.E. MacFarlane, "TRANSX 2: A Code for Interfacing MATXS Cross section Libr~ries to Nuclear Transport Codes," LA12312-MS (1992). See also J. Stepanek, "TRAMIX: A Code for Interfacing MATXS Cross section Libraries To Nuclear Transport Code for all Type Fission Reactor, Fusion Blanket as well as Shielding Analysis," Paul Scherer Institute (Draft 15-4-1988).

11.

S. Ganesan (Editor), "Preparation of processed Nuclear Data Libraries for Thermal, Fast and Fusion Research and Power Reactor Applications," Summary report of the IAEA Consultants' Meeting organized by the International Atomic Energy Agency and held at IAEA headquarters,' Vienna, Austria, 8-10 Dec. 1993, INDC(NDS)-299 (1994).

12.

J. E. White, "AMPX77 Modules for Checking and Validating Processed Multigroup Cross Sections'" paper presented at the IAEA ~onsultants' Meeting mentioned in Ref. 11 (1993).

13.

D.L. Smith, Probability, statistics and Data Uncertainties in Nuclear Science and Technology (American Nuclear Society, LeGrange Park, Illinois 1991).

14.

M. Wagner (Editor): "Proceedings of a Specialists' Meeting on Evaluation and Processing of Covariance Data", Oak Ridge National Laboratory, USA, 7-9 October 1992. Report NEA/ NSC/ DOC(93) 3. OECD, Paris (1993).

15.

E. Fort and M. Salvatores: "Fast Reactor Benchmarks and Integral Data Testing and Feedback into JEF2". In Ref. 16 below (1992).

16.

C.L. Dunford (Editor), "International symposium on Nuclear Data Evaluation Methodology," 12-16 October 1992, Brookhaven National Laboratory, World Scientific (1993).

17.

S.T. Perkins, D.E. Cullen, M.H. Chen, J.H. Hubbell, J.Rathkopf, J. Scofield, "Tables and Graphs of Atomic SUbshel1 and Relaxation Data Derived from the LLNL Evaluated Atomic Data Library (EADL) Z=1-100," Lawrence Livermore National Laboratory, Report UCRL-50400 Vol. 30 (1991).

175

18.

S.T. Perkins, D.E. Cullen, S.M. Seltzer, "Tables and Graphs of Electron Interaction Cross sections from 10 GeV to 100 GeV Derived from LLNL Evaluated Electron Data Library (EEDL), Z=1-100," Lawrence Livermore National Laboratory, Report, UCRL-50400 Vol.31 (1992).

19.

D.E. Cullen, S.T. Perkins and J.A. Rathkopf, "The 1989 Livermore Evaluated Photon Data Library (EPDL), Lawrence Livermore National Laboratory, Report UCRL-ID-103424 (1990).

20.

Fujio Maekawa, "Benchmark Experiments for Validation of Gamma- Ray Production Cross section Data," JAERI-M 94-19 pp. 369-379, Fusion Neutronics Laboratory, Japan Atomic Energy Research Institute (1994)

21.

S. Ganesan (Editor), "Preparation of Fusion Benchmarks in Electronic Format for Nuclear data Validation Studies, "Summary report of the IAEA Consultants' Meeting organized by the International Atomic Energy Agency and held at IAEA headquarters, Vienna, Austria, 13-16 Dec. 1993, INDC(NDS)-298 (1994).

22.

H.T. Hunter, "Display and Implementation of the SINBAD Electronic Data Base for Future Benchmark Needs," Paper presented at the IAEA Advisory Group Meeting mentioned in Ref. 23 below.

23.

Papers presented at the Advisory Group Meeting organized by the International Atomic Energy Agency (IAEA) on "Completion of FENDL~l and Start of FENDL-2," December 5-9, 1996. See summary report of the IAEA AGM prepared by A. B. Pashchenko, Report INDC(NDS)-352, February 1996.

24.

U. Fischer (Editor), "Integral Data Tests of the FENDL-1 Nuclear 'data Library for Fusion Applications," Summary Report of the International Working Group on Experimental and Calculational Benchmarks on Fusion Neutronics for FENDL Validation, Forschungszentrum, Karlsruhe (March 1996). To be published as an INDC report.

25.

D.E. Cullen, "The ENDF Pre-processing Codes," IAEA-NDS-39, Nuclear Data Section, International Atomic Energy Agency, Vienna,Austria (1994).

26.

R. Indira, "Neutron Flux Profile~ and Heating in radial Shields of FBTR MARK II Core," pp. 48-51 in Ref. 8 above.

27.

D.W. Muir, S. Ganesan and A.B. Pashchenko, "FENDL: A reference nuclear data library for fusion applications," Paper presented at the International Conference on Nuclear Data for Sci. and Tech., May 13-17, 1991, Juelich, Fed. Rep. of Germany.

176

28.

D.W. Muir, S. Ganesan and A.B. Pashchenko, "status, Plans and International co-operation in the preparation of the International Fusion Evaluated Nuclear data File (FENDL)", paper presented at the International Workshop on Fusion Neutronics,. 7 June 1992, KarlsrUhe, Fed. Rep.of Germany. JAERI-memo 03-305, Ed. Fusion reactor Physics laboratory, JAERI, Japan, September 1991.

29.

S. Ganesan and D.W. Muir, "IAEA Activities in Nuclear Data Processing for Thermal, Fast and Fusion Reactor Applications using the NJOY System," Paper presented in the OECD / NEA Seminar on NJOY-91 and THEMIS, OECD/NEA Data Bank, Saclay 79 April 1992.

30.

S. Ganesan, "Review of Uncertainty Files and Improved MUltigroup Cross section Files for FENDL," Summary report of the IAEA Advisory Group Meeting organized by the IAEA in cooperation with the Japan Atomic Research Institute and held at Tokai Research Establishment, JAERI, Japan, 8-12 Noyember 1993, INDC (NDS)-297 (1994).

31.

S. Ganesan, "Improved evaluations and integral data testing for FENDL," Summary report of the IAEA Advisory Group Meeting organized by the International Atomic Energy Agency in cooperation with Max-Plank Institute fuer Plasma Physik, Garching, Germany, 12-16 Sept. 1994, INDC(NDS)-312 (995).

32.

V. Gopalakrishnan (Editor), "Activity Report of Reactor Physics Division-1995," Report IGC-171,Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu, India (1996). See the contribution in this report by R. S. Keshavamurthy and R. Indira, "2-D Transport Calculations for FBTR," pp. 41-42.

33.

Thomas. M. Jordan, "Porting of Radiation Shielding Codes to Personal Computers," pp.38-40, in NEW HORIZONS IN RADIATION Topical Meeting Proceedings, PROTECTION AND SHIELDING, Pasco, Washington, April 26- May 1, 1992, American Nuclear society (1992).

34.

R. A. Schwarz, J. H. Lu, and N. Shrivastava, "Graphical post-processor created for the ORIGIN2 code," pp. 56-63 in the Meeting Proceedings of Ref. 33.

35.

K. Chatani et al., "Experiment and Analysis of Neutron streaming through an Axial Shield in an FBR Fuel SUbassembly," pp'. 292-296 in the Meeting proceedings of Ref. 33.

36.

K. Devan et al., "Generation and validation of a New 121 Group Coupled (n, g) Cross section Library for Fast Reactor Applications., Annals of Nuclear Energy (In press). See also p. 1 in report IGC-171, Ref. 32 above.

37.

R. D. McKnight and

M. L. Williams, "Reactor Benchmarks and Integral Data Testing and Feedback into ENDF/B-VI," pp. 146155 in Proceedings mentioned in Ref. 16.

177

38 ..

Me Kawai et al., "systematic Shielding Benchmark Tests Of JENDL-3 and Their Feedback to Nuclear Data Evaluation," Pp .. 180-187 in Proceedings mentioned in Ref .. 16. See also Naoki Yamano and Kohtaro UEKI, "Validation of Gamma-Ray Production Data of Iron in JENDL-3 .. 2 with Shielding Benchmark," J .. Nucl. Sci & Tech 32 (7), 614-621 (July 1995) ..

39 ..

D.. A.. Resler and R. M. White, " The LLNL Interactive System for Nuclear Data Evaluation, p. 724-733 in Proceedings mentioned in Ref. 16.

40..

T. Zidi, "Nuclear. Data Processing and Applications," M.. S .. Thesis , Ministry of National Education, Algeria (1993).

ACKNOWLEDGEMENT: The author wishes to express his sincere thanks to Dr. L. V. Krishnan, Dr. S. M. Lee and the organizing committee of RASP-96 for the invitation to present this paper. This paper was prepared to share some of my thoughts regarding the evolution and current status of nuclear data used in radiation shielding methodology in the world with particular emphasis on the Indian scenario.. I have suggested the directions in which the future activity in the use of better data for radiation shielding should possibly evolve in the coming years. The views expressed here are my own .. I have extensively used the information available in the literature. Special thanks are due to the IAEA Nuclear data section for fruitful correspondence.

178

TABLE 1. (AFTER REF. 12)

SOME EXAMPLES OF DATA TESTING BENCHMARKS

1.

SOT1-5:

"Broomstick" Experiments for Iron, Oxygen, Nitrogen,Sodium and Stainless Steel.

2.

SOT11:

ORNL Benchmark for Iron and Stainless Steel.

3.

SB2:

Secondary Gamma-Ray Production for Thermal Neutron Spectrum.

4.

S83.

Secondary Gamma-Ray Production for 1st Neutron Spectrum.

5.

S85.

ORNL14-MeV Stainless Steel/Borated Polyethylene Slab Experiment.

6.

S86.

ORNL 14-MeV Iron Duct Experiment.

7.

Winfroth Iron Benchmark Experiment.

8.

University of Illionois Iron Sphere Benchmarks (14-MeV and Cf-252)

9.

PCA-PV "Blind Test" Benchmark.

10.

Winfrith NESDIP2 and NESOIP3 Radial Shield and Cavity Expenments.

11.

PWR Shielding Benchmark (Computational)

12.

LMFBR Shielding Benchmark (Computational)

13.

LWR Shielding Benchmark (Computational)

14.

CTR Standard Blanket Benchmark (Computational)

179

I

li.BASIC tmClEAR DATA

2. SOURCES

Nuclear reection

Measurements,systematics

probabilities (cross sections)

Nuclear modeling.theory

IJ·DATA EVAlUAnON}fILES

1-------.• 14.

Very critical link

ENOL, ENDflB-VI, JEtmL, JEF,BRONO,CNOF, FENDl, etc,

I~.

APPLICATION FILES

DATA PROCESSING

- - - - - - -....... It

I

G. APPLICATION CODES

1

User interest starts here Users usually assume t~o.1

when they are actually using No.5

INTEGRAL DATA TESTING WITH MONTE CARLO PLUS

EXPERIMENTAL DATA

fig. 1 Various stages in the preparation of nuclear data for applications. [After Ref.39]

180

Basic Data

reconstructs pointwis;e

(energy dependent) cross sections from fNDfresonance

parameters and interpolation schemes. Doppler broadens and thins pointwise cross sections.

computes effective self-shielded pointwise cross

sections in the unresolved resonance enerwnmge.

produces cross sections for bounded or free scatterers in thermal energy range.

scs TTERJtlG LAW DATA

generates spectrunl

weighted self-shielded cross sections and scattering matrices.

Fig. 2 Flow chart for nuclear data processing (See Ref. 40J

181

GROUPR

PEHDf

EHDf GENDf

MATXSR

MATXS

TRAMIX

RMZfLUX

CARDS

aao

AMISH

o ISOTXS

GOXS

Fig. 2 (Continued): Flow chart for nuclear data processing

182

NUCLEAR DATA FOR RADIATION TRANSPORT (CROSS-SECTIONSJ

TIME

ACTIVITY

Needs Recognition and Establishment of Accuracy Requirements

MONTHS

Nuclear Data McaSUfcmciiits and

YEARS/NUCLIDE

Nuclear Model Calculations

Evaluation Process and Generation of Evaluated Nuclear Dahl File

YEARS/NUCUDE

Programmatic Requirements

Define Specific Data in a

WEEKs-

Well- Defined Sequence

Multigruup CiOSS Section library

Ag. 3 : Cross Sedion data required for applications usually evolve from a national effort to provide state-of-the-art data based on established needs

and uncertainities. The process requires an iterative sequence of events to yield 3D quality assured transport cross-sedion library. (After J.E..Wh ite , Hel. 12J

183

liME

ACTIVllY

Processing Methods Development

YEARS

YEARS

Preliminary Multigroup library

Data Testing

MONTHS""

final Multigroup Library

Fig. 3: [Continued]

184

fiRST GENERATION:

DOSE MEASUREMENTS ----~

I

SEconD GENERATION:

ENERGV SPECTRUM MEASUREMEtITS

I

I

EVACUATED fLIGHT PATH

SPECTROMETER

I I

THIRD GENERATION:

DETECTOR

DETECTOR

TIME-DEPENDENT DOSE RATE MEASUREMENTS!

TIME-DEPENDENT ENERGV SPECTRUM MEASUREMENTS

I

EVACUATED FUGHT PATH

fig. 4a:

J

EJ

DETECTOR! SPECTROMETER

THREE GENERATIONS Of SHIELDING BENCHMARK EXPERIMENTS (AFTER REF. 1)

185

PULSED ELECTRON

BEAM BACKSCATTER T8 ELECTRON TARGET Be PHOTOHEUTRON

SOURCE

SHIEUl

o!

Jl-!!

EVACUATED 50-METER FUGHT PATH

LR

~~

~~ 0

~ ~

13:11 U-cm NE-2'13

CONCRETE SLABS

EVACUATED 50-METER fLIGHT PATH

15-cmDlAM. SPHERICAL SAMPlE (N AND C)

Fig. Jib: THIRD GENERATION SHIELDING EXPERIMENT (AfTER REF. 1)

186

ENOFM - VI

OLe - 99

MODER

MODER

RECONR

RECONR

GROUPR

GAMINR

OTfR

OTFR

NGCOUP

NUETRON - GAMMA COUPLED lIBRARV

fig. 5

Flow chart for the preparation of coupled neutron-photon library at Kalpakkam (After Ref. 8]

187

l 1

Neutron Cross-section

Gamma Cross-section \( PROOUCTlOt~

'{ TRAt~SPORT

INFINITE DILUTE

CROSS-SECTIOU

CROSS-SECTIOt~

CROSS-SECTION ( JSD 100) (JSUJ2)

SElf SHIELmtlG fACTOR (JFT 2(0)

1

1

PI'- I J2)

I

I

1

I

1

1

2-D

Production of

Core Cal&..

Group Constants

1

(RADHEAT ~\l3)

(CITATION etc)

1

,

1

Effective Macroscopic Cross-section n:'1 Gr '(= 7 Gr

Neutron

Source Distribution

1

1

1

t

1

1 - D

Transport Calc, (ANISN)

1

1

Collapsed Croassection

1

1

n: 21 Gr

'(= 7 Gr 1

1

3 [) Detailed Calc.

2 -0

Transport Calc.

~

1

(DOT3.5,DORT)

(MORSE, TORT)

1

t

Neutron flue nee for Structural Material (fast nut.dpa)

t

+

Generation of Radiation sources (;MNoa,22Na,CPlP,

Dose Rate (Rem,SU)

t

Detector Response (tUS Response)

1

1

1

1

1

Fig. 6: SHIELDING METHODOLOGY FOLLOWED IN THE DESIGN OF THE MONJU REACTOR (AFTER REF. 35)

1

1

188