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Technical Competencies for the Safe Interim Storage and Management of 233U at U.S. Department of Energy Facilities

Description: Uranium-233 (with concomitant {sup 232}U) is a man-made fissile isotope of uranium with unique nuclear characteristics which require high-integrity alpha containment biological shielding, and remote handling. The special handling considerations and the fact that much of the {sup 233}U processing and large-scale handling was performed over a decade ago underscore the importance of identifying the people within the DOE complex who are currently working with or have worked with {sup 233}U. The availability of these key personnel is important in ensuring safe interim storage, management and ultimate disposition of {sup 233}U at DOE facilities. Significant programs are ongoing at several DOE sites with actinides. The properties of these actinide materials require many of the same types of facilities and handling expertise as does {sup 233}U.
Date: February 17, 1999
Creator: Campbell, D. O.; Krichinsky, A. M.; Laughlin, S. S.; Van Essen, D. C. & Yong, L. K.
Partner: UNT Libraries Government Documents Department

Practical and cost effective solution to the need for shielding penetrations against photons and neutrons from normal and accident losses

Description: The Thomas Jefferson National Accelerator Facility (Jefferson Lab) houses a 4 GeV, 200 {micro}A continuous wave (CW) recirculating electron accelerator. This underground accelerator is made up of two superconducting linear accelerators (linacs), two arcs, a beam switch yard (BSY), and three end stations. Each linac has the capability of accelerating electrons to a kinetic energy of 400 MeV. The arcs contain four (on the west) and five (on the east) beamlines to transport the beams of differing energies back into the linacs. The BSY steers the desired beams into the end stations as needed for nuclear physics experiments. The accelerator is connected to the control and diagnostic electronics in the above-ground service buildings via 30 cm and 51 cm diameter penetrations that travel through 4.6 m of soil and concrete. As a result, there exists the potential for personnel exposure to radiation scattering up the penetrations. It was desired that some of these buildings become Uncontrolled Areas, so that persons in the buildings would not require dosimetry. The Jefferson Lab Beam Containment Policy also requires that effective dose rates to workers be limited to 150 mSv in one hour if a maximum beam power loss accident was to continue unabated.
Date: January 1, 1997
Creator: Schwahn, S.
Partner: UNT Libraries Government Documents Department

Experiment proposal for the determination of neutron spectra from targeted electron beams

Description: There is a dearth of experimental data on the production and yields of neutrons from targeted electron beams; yet, for accelerator radiation protection these data are of the greatest importance in setting up methods of shielding and other means for protecting people against ionizing radiation. Although adequate for simple cases and lateral production angles, empirical analytical methods are not suitable for the more complicated geometries or source configurations often met with in practice. Monte Carlo (MC) methods that model the transport of neutrons provide far better results in many cases but rely on the random generation of the energy of a source particle selected for any beam condition, production angle and target configuration. A number of theoretical approaches to the derivation of a model for the production of particle events at energies greater than the giant resonance region have been made. Many of these are based on the quasi deuteron model of the nucleus and operate over photon energies in the range 30 MeV to 400 MeV. A method is also available, based on the vector meson dominance model which is designed to work above the photopion resonance region where the cross section levels off at a few GeV (Ranft 1987). Both of these models are limited in utility to a certain energy range and both show some discrepancies with existing empirical methods. More recently a new fragmentation model was developed, which could be used over a large energy range and modeled all production processes. This new method also showed differences from the traditional approaches and a thorough comparison indicated that the event generator in conjunction with conventional MC transport codes produced results a factor two to three higher than the results using the empirical methods. This unsatisfactory situation can only be resolved by making measurements of proper physical quantities ...
Date: October 1, 1996
Creator: Degtyarenko, P. & Stapleton, G.
Partner: UNT Libraries Government Documents Department

Topics in radiation at accelerators: Radiation physics for personnel and environmental protection

Description: In the first chapter, terminology, physical and radiological quantities, and units of measurement used to describe the properties of accelerator radiation fields are reviewed. The general considerations of primary radiation fields pertinent to accelerators are discussed. The primary radiation fields produced by electron beams are described qualitatively and quantitatively. In the same manner the primary radiation fields produced by proton and ion beams are described. Subsequent chapters describe: shielding of electrons and photons at accelerators; shielding of proton and ion accelerators; low energy prompt radiation phenomena; induced radioactivity at accelerators; topics in radiation protection instrumentation at accelerators; and accelerator radiation protection program elements.
Date: October 1, 1996
Creator: Cossairt, J.D.
Partner: UNT Libraries Government Documents Department

Parameterizations for shielding electron accelerators based on Monte Carlo studies

Description: Numerous recipes for designing lateral slab neutron shielding for electron accelerators are available and each generally produces rather similar results for shield thicknesses of about 2 m of concrete and for electron beams with energy in the 1 to 10 GeV region. For thinner or much thicker shielding the results tend to diverge and the standard recipes require modification. Likewise for geometries other than lateral to the beam direction further corrections are required so that calculated results are less reliable and hence additional and costly conservatism is needed. With the adoption of Monte Carlo (MC) methods of transporting particles a much more powerful way of calculating radiation dose rates outside shielding becomes available. This method is not constrained by geometry, although deep penetration problems need special statistical treatment, and is an excellent approach to solving any radiation transport problem providing the method has been properly checked against measurements and is free from the well known errors common to such computer methods. This present paper utilizes the results of MC calculations based on a nuclear fragmentation model named DINREG using the MC transport code GEANT and models them with the normal two parameter shielding expressions. Because the parameters can change with electron beam energy, angle to the electron beam direction and target material, the parameters are expressed as functions of some of these variables to provide a universal equations for shielding electron beams which can used rather simply for deep penetration problems in simple geometry without the time consuming computations needed in the original MC programs. A particular problem with using simple parameterizations based on the uncollided flux is that approximations based on spherical geometry might not apply to the more common cylindrical cases used for accelerator shielding. This source of error has been discussed at length by Stevenson and others. ...
Date: October 1, 1996
Creator: Degtyarenko, P. & Stapleton, G.
Partner: UNT Libraries Government Documents Department

SSC workshop on environmental radiation

Description: The Superconducting Super Collider is a 20 TeV-on-20 TeV proton beam collider where two 20-TeV proton accelerators whose beams, rotating in opposite senses, are brought into collision to provide 40 TeV in the center of mass. The scale of the project is set by the 6.6 tesla magnet guide field for the protons which results in a roughly circular machine with a circumference of 83 km (51.5 mi.). The energy scale of the proton beams and the physical scale of the machine are an order of magnitude greater than for any presently operating or contemplated proton accelerator yet the facility must be operated within the same strict radiological guidelines as existing accelerators in the US and Europe. To ensure that the facility conforms to existing and projected guidelines both in design and operation, the Workshop was charged to review the experience and practices of existing accelerator laboratories, to determine the relevant present and projected regulatory requirements, to review particle production and shielding data from accelerators and cosmic rays, to study the design and operational specifications of the Collider, to examine the parameters set forth in the Siting Parameters Document, and to evaluate the computational tools available to model the radiation patterns arising under various operational and failure scenarios. This report summarizes the extensive and intensive presentations and discussions of the Workshop. A great deal of material, much of it in the form of internal reports from the various laboratories and drafts of works in preparation, was provided by the participants for the various topics. This material, including the viewgraphs used by the presenters, forms the background and basis for the conclusions of the Workshop and, as such, is an important part of the Workshop. An introduction to the material and a catalog by topic are presented as section 6 of ...
Date: January 9, 1986
Partner: UNT Libraries Government Documents Department

Beam transport radiation shielding for branch lines 2-ID-B and 2-ID-C

Description: The x-ray radiation shielding requirements beyond the first optics enclosure have been considered for the beam transport of the 2-ID-B and 2-ID-C branch lines of Sector 2 (SRI-CAT) of the APS. The first three optical components (mirrors) of the 2-ID-B branch are contained within the shielded first optics enclosure. Calculations indicate that scattering of the primary synchrotron beam by beamline components outside the enclosure, such as apertures and monochromators, or by gas particles in case of vacuum failure is within safe limits for this branch. A standard 2.5-inch-diameter stainless steel pipe with 1/16-inch-thick walls provides adequate shielding to reduce the radiation dose equivalent rate to human tissue to below the maximum permissible limit of 0.25 mrem/hr. The 2-ID-C branch requires, between the first optics enclosure where only two mirrors are used and the housing for the third mirror, additional lead shielding (0.75 mm) and a minimum approach distance of 2.6 cm. A direct beam stop consisting of at least 4.5 mm of lead is also required immediately downstream of the third mirror for 2-ID-C. Finally, to stop the direct beam from escaping the experimental station, a beam stop consisting of at least 4-mm or 2.5-mm steel is required for the 2-ID-B or 2-ID-C branches, respectively. This final requirement can be met by the vacuum chambers used to house the experiments for both branch lines.
Date: August 1, 1995
Creator: Feng, Y.P.; Lai, B.; McNulty, I.; Dejus, R.J.; Randall, K.J. & Yun, W.
Partner: UNT Libraries Government Documents Department

A two-dimensional point-kernel model for dose calculations in a glovebox array

Description: An associated paper details a model of a room containing gloveboxes using the industry standard dose equivalent (dose) estimation tool MCNP. Such tools provide an excellent means for obtaining relatively reliable estimates of radiation transport in a complicated geometric structure. However, creating the input deck that models the complicated geometry is equally complicated. Therefore, an alternative tool is desirable that provides reasonable accurate dose estimates in complicated geometries for use in engineering-scale dose analyses. In the past, several tools that use the point-kernel model for estimating doses equivalent have been constructed (those referenced are only a small sample of similar tools). This new tool, the Photon and Neutron Dose Equivalent Model Of Nuclear materials Integrated with an Uncomplicated geometry Model (PANDEMONIUM), combines point-kernel and diffusion theory calculation routines with a simple geometry construction tool. PANDEMONIUM uses Visio{trademark} to draw a glovebox array in the room, including hydrogenous shields, sources and detectors. This simplification in geometric rendering limits the tool to two-dimensional geometries (and one-dimensional particle transport calculations).
Date: June 1, 1999
Creator: Kornreich, D.E. & Dooley, D.E.
Partner: UNT Libraries Government Documents Department

Upgraded safety analysis document including operations policies, operational safety limits and policy changes. Revision 2

Description: The National Synchrotron Light Source Safety Analysis Reports (1), (2), (3), BNL reports {number_sign}51584, {number_sign}52205 and {number_sign}52205 (addendum) describe the basic Environmental Safety and Health issues associated with the department`s operations. They include the operating envelope for the Storage Rings and also the rest of the facility. These documents contain the operational limits as perceived prior or during construction of the facility, much of which still are appropriate for current operations. However, as the machine has matured, the experimental program has grown in size, requiring more supervision in that area. Also, machine studies have either verified or modified knowledge of beam loss modes and/or radiation loss patterns around the facility. This document is written to allow for these changes in procedure or standards resulting from their current mode of operation and shall be used in conjunction with the above reports. These changes have been reviewed by NSLS and BNL ES and H committee and approved by BNL management.
Date: March 1, 1996
Creator: Batchelor, K.
Partner: UNT Libraries Government Documents Department

Examination of criticality accident alarm coverage on the operating floors of Buildings X-333, X-330, and X-326 at the Portsmouth Gaseous Diffusion Plant. Revision 1

Description: The diffusion cascade processing equipment at the Portsmouth Gaseous Diffusion Plant (PORTS) is located in Buildings X-333, X-330, and X-326. These buildings were referred to as the cascade buildings. Because enriched uranium operations are performed within the cascade buildings, the potential for a criticality accident in these buildings exists. A Criticality Accident Alarm System (CAAS) is in place to alarm in the event of a criticality accident. The CAAS is required to be designed to immediately detect the minimum accident-of-concern. A minimum accident-of-concern in an area with nominal shielding delivers the equivalent of an absorbed dose rate in free air of 20 rads per minute at a distance of 2 meters from the reacting material [Am86]. This report summarizes the analysis that was performed to evaluate the CAAS response to selected minimum accidents-of-concern on the operating floor of the cascade buildings. Selection of potential accident locations was based, in part, on the maximum distance to the closest CAAS detector. The other factor in selecting potential accident locations for analysis was the amount of intervening shielding between the accident location and CAAS detector. If the CAAS was predicted to alarm under conditions of significant shielding, then the system presumably would alarm in response to all accidents greater than the minimum accident-of-concern, at closer distances, and with less shielding.
Date: March 1, 1997
Creator: Brown, A.S.; Tayloe, R.W. Jr.; Wollard, J. & Dobelbower, M.C.
Partner: UNT Libraries Government Documents Department

Evaluation of gamma radiation shielding for nuclear waste shipping casks

Description: A method has been developed for evaluating gamma radiation shielding of shipping casks that are used to transport nuclear waste with ill-defined radionuclide contents. The method is based on calculations that establish individual limits for a comprehensive list of radionuclides in the waste, assuming that each radionuclide is uniformly distributed in a volumetric source in the cask. For multiple radionuclide mixtures, a linear fraction rule is used to restrict the total amount of radionuclides such that the sum of the fractions does not exceed 1. As long as the radionuclide limits and the linear fraction rule are followed, it can be shown that the regulatory dose rate requirements for a cask will be satisfied under normal conditions of transport and in a hypothetical accident during which the shielding thickness of the cask has been reduced by 40%.
Date: May 1, 1998
Creator: Liu, Y.Y.; Carlson, R.D.; Primeau, S.J. & Wangler, M.E.
Partner: UNT Libraries Government Documents Department

MARS code developments, benchmarking and applications

Description: Recent developments of the MARS Monte Carlo code system for simulation of hadronic and electromagnetic cascades in shielding, accelerator and detector components in the energy range from a fraction of an electron volt up to 100 TeV are described. The physical model of hadron and lepton interactions with nuclei and atoms has undergone substantial improvements. These include a new nuclear cross section library, a model for soft prior production, a cascade-exciton model, a dual parton model, deuteron-nucleus and neutrino-nucleus interaction models, a detailed description of negative hadron and muon absorption, and a unified treatment of muon and charged hadron electro-magnetic interactions with matter. New algorithms have been implemented into the code and benchmarked against experimental data. A new Graphical-User Interface has been developed. The code capabilities to simulate cascades and generate a variety of results in complex systems have been enhanced. The MARS system includes links to the MCNP code for neutron and photon transport below 20 MeV, to the ANSYS code for thermal and stress analyses and to the STRUCT code for multi-turn particle tracking in large synchrotrons and collider rings. Results of recent benchmarking of the MARS code are presented. Examples of non-trivial code applications are given for the Fermilab Booster and Main Injector, for a 1.5 MW target station and a muon storage ring.
Date: March 24, 2000
Creator: Mokhov, N.V.
Partner: UNT Libraries Government Documents Department

Ceramicrete: A novel ceramic packaging system for spent-fuel transport and storage

Description: This presentation summarizes efforts to develop and apply chemically bonded phosphate ceramic (Ceramicrete{trademark}) technology for radiation shielding applications. The specific application being targeted is a packaging system for spent-fuel transport and storage. Using Ceramicrete technology under ambient conditions, the authors can produce dense and hard ceramic forms that incorporate second-phase material. Ceramicrete inherently is a superior shielding material because it contains large amounts of bound water in its crystal structure and can be cast in any shape. A parametric study was conducted on Ceramicrete that contained second-phase additions of metals and other ceramic powders. Results of various standardized tests that included mechanical performance and shielding from neutrons are presented. The fabrication of complex shapes and structures by Ceramicrete technology is discussed. Ceramicrete is compared with other currently available shielding systems that are based on concrete and polymers.
Date: February 25, 2000
Creator: Singh, D.; Jeong, S. Y.; Dwyer, K. & Abesadze, T.
Partner: UNT Libraries Government Documents Department

Global shielding analysis of the 2-element ANS core and reflector with photoneutrons

Description: This paper describes the initial global 2-D shielding analyses for the 2-element, heavy-water cooled and reflected Advanced Neutron Source reactor which was to have been built in Oak Ridge, Tennessee. The portion of the system analyzed encompassed the highly enriched core, the 1.5-m-thick heavy-water reflector, the aluminum reflector vessel, and the first 0.2 m of light water beyond the reflector vessel. While some results are presented, this paper focuses primarily on the lessons learned during the analysis of this rather unique system.
Date: April 1, 1996
Creator: Bucholz, J.A.
Partner: UNT Libraries Government Documents Department

Summary report for ITER task - D4: Activation calculations for the lithium vanadium ITER design

Description: Detailed activation analysis for ITER has been performed as a part of ITER Task D4. The calculations have been performed for the shielding blanket (SS/water) and for the breeding blanket (Li/V) options. The activation code RACC-P, which has been modified under ITER Task-D-10 for pulsed operation, has been used in this analysis. The spatial distributions of the radioactive inventory, decay heat, biological hazard potential, and the contact dose were calculated for the two designs for different operation modes and targeted fluences. A one-dimensional toroidal geometrical model has been utilized to determine the neutron fluxes in the two designs. The results are normalized for an inboard and outboard neutron wall loading of 0.91 and 1.2 MW/m{sup 2}, respectively. The point-wise distributions of the decay gamma sources have been calculated everywhere in the reactor at several times after the shutdown of the two designs and are then used in the transport code ONEDANT to calculate the biological dose everywhere in the reactor. The point-wise distributions of all the responses have also been calculated. These calculations have been performed for neutron fluences of 3.0 MWa/m{sup 2}, which corresponds to the target fluence of ITER, and 0.1 MWa/m{sup 2}, which is anticipated to correspond to the beginning of the extended maintenance period. The decay heat results show that a large fraction of this energy (50 to 90%) Is produced by photons. This implies that this energy would be transported to different parts of the reactor, thus relieving the energy concentration at high intensity source locations such as the first wall. Accurate modeling for the decay gamma transport is required to produce realistic spatial distribution of the decay heat which may be used in LOCA and LOFA analyses. The results of the pulsed operation, using the new version of RACC, show large reductions in the ...
Date: February 1, 1995
Creator: Attaya, H.
Partner: UNT Libraries Government Documents Department

Decontamination and decommissioning of the Experimental Boiling Water Reactor at Argonne National Laboratory

Description: The Experimental Boiling Water Reactor (EBWR), located on the Argonne National Laboratory-East (ANL-E) site, started operations in 1957. The initial rating was 20 MW(t). The rating was eventually increased to 70 MW(t) in 1959 and 100 MW(t) in 1962. The reactor was shut down in 1967 and all of the fuel was removed from the facility. The facility was placed in dry lay-up until 1986. ANL-E personnel started the decontamination and decommissioning (D&D) effort in 1986. Supporting equipment such as the external steam system and some of the upper reactor components, the core riser and the top fuel shroud, were removed at that time. Characterization of the facility was also undertaken. The contract to complete the EBWR D&D Project was issued in December 1993. The initial schedule called for the final effort to be divided into five phases that were to be completed over a four year period. However, this schedule was subsequently consolidated, at the request of ANL-E, to a thirteen month period, with the on-site work to be completed by the end of 1994. The EBWR D&D Project is approximately 88% complete. A small quantity of reactor internals remains to be volume reduced along with the removal of the SFSP water treatment system. Upon completion of this work the facility will be decontaminated and a final survey completed. The planned completion of on-site work is scheduled for July 1995.
Date: July 1, 1995
Creator: Sears, L.F. & Fellhauer, C.
Partner: UNT Libraries Government Documents Department

Reactor Refueling - Interim Decay Storage (FFTF)

Description: The IDS facility is located between the CLEM rails and within the FFTF containment building. It is located in a rectangular steel-lined concrete cell which lies entirely below the 550 ft floor level with the top flush with the 550 ft floor level. The BLTC rails within containment traverse the IDS cover (H-4-38001). The facility consists of a rotatable storage basket submerged in liquid sodium which is contained in a stainless steel tank. The storage positions within the basket are arranged so that it is not physically possible to achieve a critical array. The primary vessel is enclosed in a secondary guard tank of such size and arrangement that, should a leak develop in the primary tank, the sodium level would not fall below the top of the fueled section of the stored core components or test assemblies. The atmosphere outside the primary vessel, but within the concrete cell, is nitrogen which also serves as a heat transfer medium to control the cell temperature. To provide space for the storage of test assemblies such as the OTA and CLIRA, 10 storage tubes (each approximately 43-1/4 ft long) are included near the center of the basket. This arrangement requires that the center of the primary vessel be quite deep. In this region, the primary vessel extends downward to elevation 501 ft 6 inches while the guard tank reaches 500 ft 4 inches. The floor of the cell is at 499 ft a inches which is 51 ft below the operating room floor. Storage positions are provided for 112 core components in the upper section of the storage basket. These positions are arranged in four circles, all of which are concentric with the test element array and the storage basket. The primary vessel and the guard tank are shaped to provide the ...
Date: June 18, 1990
Creator: Mcfadden, N. R. & Omberg, R. P.
Partner: UNT Libraries Government Documents Department

New facility shield design criteria

Description: The purpose of the criteria presented here is to provide standard guidance for the design of nuclear radiation shields thoughout new facilities. These criteria are required to assure a consistent and integrated design that can be operated safely and economically within the DOE standards. The scope of this report is confined to the consideration of radiation shielding for contained sources. The whole body dose limit established by the DOE applies to all doses which are generally distributed throughout the trunk of the body. Therefore, where the whole body is the critical organ for an internally deposited radionuclide, the whole body dose limit applies to the sum of doses received must assure control of the concentration of radionuclides in the building atmosphere and thereby limit the dose from internal sources.
Date: July 1, 1981
Creator: Howell, W.P.
Partner: UNT Libraries Government Documents Department

BUGLE-96: A revised multigroup cross section library for LWR applications based on ENDF/B-VI Release 3

Description: A revised multigroup cross-section library based ON ENDF/B-VI Release 3 has been produced for light water reactor shielding and reactor pressure vessel dosimetry applications. This new broad-group library, which is designated BUGLE-96, represents an improvement over the BUGLE-93 library released in February 1994 and is expected to replace te BUGLE-93 data. The cross-section processing methodology is the same as that used for producing BUGLE-93 and is consistent with ANSI/ANS 6.1.2. As an added feature, cross-section sets having upscatter data for four thermal neutron groups are included in the BUGLE-96 package available from the Radiation Shielding Information Center. The upscattering data should improve the application of this library to the calculation of more accurate thermal fluences, although more computer time will be required. The incorporation of feedback from users has resulted in a data library that addresses a wider spectrum of user needs.
Date: May 1, 1996
Creator: White, J.E.; Ingersoll, D.T.; Slater, C.O. & Roussin, R.W.
Partner: UNT Libraries Government Documents Department

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation: Functional modules F1-F8

Description: This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This volume consists of the section of the manual dealing with eight of the functional modules in the code. Those are: BONAMI - resonance self-shielding by the Bondarenko method; NITAWL-II - SCALE system module for performing resonance shielding and working library production; XSDRNPM - a one-dimensional discrete-ordinates code for transport analysis; XSDOSE - a module for calculating fluxes and dose rates at points outside a shield; KENO IV/S - an improved monte carlo criticality program; COUPLE; ORIGEN-S - SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay, and associated radiation source terms; ICE.
Date: March 1, 1997
Partner: UNT Libraries Government Documents Department

SCALE: A modular code system for performing standardized computer analyses for licensing evaluation

Description: This Manual represents Revision 5 of the user documentation for the modular code system referred to as SCALE. The history of the SCALE code system dates back to 1969 when the current Computational Physics and Engineering Division at Oak Ridge National Laboratory (ORNL) began providing the transportation package certification staff at the U.S. Atomic Energy Commission with computational support in the use of the new KENO code for performing criticality safety assessments with the statistical Monte Carlo method. From 1969 to 1976 the certification staff relied on the ORNL staff to assist them in the correct use of codes and data for criticality, shielding, and heat transfer analyses of transportation packages. However, the certification staff learned that, with only occasional use of the codes, it was difficult to become proficient in performing the calculations often needed for an independent safety review. Thus, shortly after the move of the certification staff to the U.S. Nuclear Regulatory Commission (NRC), the NRC staff proposed the development of an easy-to-use analysis system that provided the technical capabilities of the individual modules with which they were familiar. With this proposal, the concept of the Standardized Computer Analyses for Licensing Evaluation (SCALE) code system was born. This manual covers an array of modules written for the SCALE package, consisting of drivers, system libraries, cross section and materials properties libraries, input/output routines, storage modules, and help files.
Date: March 1, 1997
Partner: UNT Libraries Government Documents Department

Workshop on radiological aspects of SSC operations

Description: Integral to the design of an accelerator facility is the provision of adequate shielding to contain any radiation arising from operation of the facility. Complementary to the questions of environmental shielding are a number of radiation questions related to operation of the completed facility. One obvious need is the specification of systems for monitoring environmental emissions to ensure consistency between the design criteria and the actual levels during operation. Another question is the effect on the components of the machine of the radiation within the environmental shield. These questions were examined at the workshop. This report is a summary of the materials presented at the workshop.
Date: May 1, 1987
Creator: Toohig, T. E.
Partner: UNT Libraries Government Documents Department