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SCALE-4 (Standardized Computer Analyses for Licensing Evaluation): An improved computational system for spent-fuel cask analysis

Description: The purpose of this paper is to provide specific information regarding improvements available with Version 4.0 of the SCALE system and discuss the future of SCALE within the current computing and regulatory environment. The emphasis focuses on the improvements in SCALE-4 over that available in SCALE-3. 10 refs., 1 fig., 1 tab.
Date: January 1, 1989
Creator: Parks, C.V.
Partner: UNT Libraries Government Documents Department

Adjoint-based sensitivity analysis for reactor accident codes

Description: This paper summarizes a recently completed study that identified and investigated the difficulties and limitations of applying first-order adjoint sensitivity methods to reactor accident codes. The work extends earlier adjoint sensitivity formulations and applications to consider problem/model discontinuities in a general fashion, provide for response (R) formulations required by reactor safety applications, and provide a scheme for accurately handling extremely time-sensitive reactor accident responses. The scheme involves partitioning (dividing) the model into submodels (with spearate defining equations and initial conditions) at the location of discontinuity. Successful partitioning moves the problem dependence on the discontinuity location from the whole model system equations to the initial conditions of the second submodel.
Date: January 1, 1985
Creator: Parks, C.V.
Partner: UNT Libraries Government Documents Department

Plutonium Production Using Natural Uranium From the Front-End of the Nuclear Fuel Cycle

Description: This report explores the potential for producing weapons-grade plutonium using conventional industrial resources, oxides of natural uranium (namely UO{sub 3}), and either heavy water or reactor-grade graphite. Using established codes and data for nuclear analysis, it is demonstrated that physics-based reactor models capable of producing kilogram quantities of weapons-grade plutonium can be readily conceived. The basic assumptions and analysis approach are discussed together with the results of the analysis.
Date: August 29, 2002
Creator: Parks, C.V.
Partner: UNT Libraries Government Documents Department

Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel

Description: This report has been prepared to review relevant background information and provide technical discussion that will help initiate a PIRT (Phenomena Identification and Ranking Tables) process for use of burnup credit in light-water reactor (LWR) spent fuel storage and transport cask applications. The PIRT process will be used by the NRC Office of Nuclear Regulatory Research to help prioritize and guide a coordinated program of research and as a means to obtain input/feedback from industry and other interested parties. The review and discussion in this report are based on knowledge and experience gained from work performed in the United States and other countries. Current regulatory practice and perceived industry needs are also reviewed as a background for prioritizing technical needs that will facilitate safe practice in the use of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation is given. Finally, phenomena that need to be better understood for effective licensing, together with technical issues that require resolution, are presented and discussed in the form of a prioritization ranking and initial draft program plan.
Date: March 13, 2000
Creator: Parks, C. V.
Partner: UNT Libraries Government Documents Department

Point kernel versus radiation transport for iron deep penetration problems

Description: The purpose of this study was to determine the relative merits and adequacy of the QAD-CG and QAD-CGGP point kernel codes for spent fuel shielding problems by comparing gamma dose results with those of several radiation transport codes - DOT IV, MORSE-SGC, XSDRNPM-S, and MCNP. An intercomparison of results from the radiation transport codes was also of secondary interest. The problem considered was an R-Z spent fuel cask model consisting of a homogenized source region (UO/sub 2/ fuel + basket) surrounded by a 38 cm thick cast iron body. The total cask height is 530 cm. Photon source strengths for eight different discrete energy lines (0.6 to 2.8 MeV) were specified. 8 refs., 1 tab.
Date: January 1, 1987
Creator: Broadhead, B.L. & Parks, C.V.
Partner: UNT Libraries Government Documents Department

Characterization of neutron sources from spent fuel casks. [Skyshine]

Description: In the interim period prior to the acceptance of spent fuel for disposal by the USDOE, utilities are beginning to choose dry cask storage as an alternative to pool re-racking, transshipments, or new pool construction. In addition, the current MRS proposal calls for interim dry storage of consolidated spent fuel in concrete casks. As part of the licensing requirements for these cask storage facilities, calculations are typically necessary to determine the yearly radiation dose received at the site boundary. Unlike wet facilities, neutron skyshine can be an important contribution to the total boundary dose from a dry storage facility. Calculation of the neutron skyshine is in turn heavily dependent on the source characteristics and source model selected for the analysis. This paper presents the basic source characteristics of the spent fuel stored in dry casks and discusses factors that must be considered in evaluating and modeling the radiation sources for the subsequent skyshine calculation. 4 refs., 1 tab.
Date: January 1, 1987
Creator: Parks, C.V. & Pace, J.V. III
Partner: UNT Libraries Government Documents Department

Detector-selection technique for Monte Carlo transport in azimuthally symmetric geometries

Description: Many radiation transport problems contain geometric symmetries which are not exploited in obtaining their Monte Carlo solutions. An important class of problems is that in which the geometry is symmetric about an axis. These problems arise in the analyses of a reactor core or shield, spent fuel shipping casks, tanks containing radioactive solutions, radiation transport in the atmosphere (air-over-ground problems), etc. Although amenable to deterministic solution, such problems can often be solved more efficiently and accurately with the Monte Carlo method. For this class of problems, a technique is described in this paper which significantly reduces the variance of the Monte Carlo-calculated effect of interest at point detectors.
Date: January 1, 1982
Creator: Hoffman, T.J.; Tang, J.S. & Parks, C.V.
Partner: UNT Libraries Government Documents Department

Evaluation of spent fuel isotopics, radiation spectra and decay heat using the scale computational system

Description: In order to be a self-sufficient system for transport/storage cask shielding and heat transfer analysis, the SCALE system developers included modules to evaluate spent fuel radiation spectra and decay heat. The primary module developed for these analyses is ORIGEN-S which is an updated verision of the original ORIGEN code. The COUPLE module was also developed to enable ORIGEN-S to easily utilize multigroup cross sections and neutron flux data during a depletion analysis. Finally, the SAS2 control module was developed for automating the depletion and decay via ORIGEN-S while using burnup-dependent neutronic data based on a user-specified fuel assembly and reactor history. The ORIGEN-S data libraries available for depletion and decay have also been significantly updated from that developed with the original ORIGEN code.
Date: January 1, 1986
Creator: Parks, C.V.; Hermann, O.W. & Ryman, J.C.
Partner: UNT Libraries Government Documents Department

Review of the international conference on nuclear criticality-issues, discussions, and challenges

Description: The Fifth International Conference on Nuclear Criticality Safety (ICNC`95) was held September 17-22, 1995, in Albuquerque, New Mexico, USA. Organization and support for the conference was provided by the Sandia National Laboratories (SNL), Los Alamos National Laboratory (LANL), the University of New Mexico, and the Organization for Economic Cooperation and Development (OECD). This conference traces its history back to 1981 when a group of select criticality safety specialists (mostly experimentalists) from France, Germany, Japan, the United Kingdom, and the United States participated in a small conference at LANL in the United States. The motivation for the conference had been provided by Dr. J. C. Manaranche of France who had asked D. Smith and G. E. Whitesides of the United States if it would be possible for the French experimentalists to be able to visit the experimental facilities at LANL. This first conference was followed by a similar conference held in Dijon, France, in 1993. Then in 1987 the conference was hosted by the Japanese and opened to much wider participation by criticality safety specialists involved in experiments, methods development and analysis, and operations. With the 1987 conference in Japan and the fourth conference (ICNC`91) held in the United Kingdom, the interest and international participation by the criticality safety community has grown rapidly. With this background, the occasion of ICNC`95 was one of much expectation.
Date: December 31, 1995
Creator: Parks, C.V. & Whitesides, G.E.
Partner: UNT Libraries Government Documents Department

Phenomena and Parameters Important to Burnup Credit

Description: Since the mid-1980s, a significant number of studies have been directed at understanding the phenomena and parameters important to implementation of burnup credit in out-of-reactor applications involving pressurized-water-reactor (PWR) spent fuel. The efforts directed at burnup credit involving boiling-water-reactor (BWR) spent fuel have been more limited. This paper reviews the knowledge and experience gained from work performed in the US and other countries in the study of burnup credit. Relevant physics and analysis phenomenon are identified, and an assessment of their importance to burnup credit implementation for transport and dry cask storage is given.
Date: January 10, 2001
Creator: Parks, C. V.
Partner: UNT Libraries Government Documents Department

U.S. Regulatory Research Program for Implementation of Burnup Credit in Transport Casks

Description: In 1999 the U.S. Nuclear Regulatory Commission (U.S. NRC) initiated a research program to support the development of technical bases and guidance that would facilitate the implementation of burnup credit into licensing activities for transport and dry cask storage. This paper reviews the following major areas of investigation: (1) specification of axial burnup profiles, (2) assumption on cooling time, (3) allowance for assemblies with fixed and removable neutron absorbers, (4) the need for a burnup margin for fuel with initial enrichments over 4 wt %, and (5) evaluation of assay data and critical experiments. The capabilities of a new computational tool that facilitates the performance and coupling of the depletion and criticality analyses needed for burnup credit are also discussed.
Date: September 10, 2001
Creator: Parks, C. V.
Partner: UNT Libraries Government Documents Department

Review of criticality safety and shielding analysis issues for transportation packages

Description: The staff of the Nuclear Engineering Applications Section (NEAS) at Oak Ridge National Laboratory (ORNL) have been involved for over 25 years with the development and application of computational tools for use in analyzing the criticality safety and shielding features of transportation packages carrying radioactive material (RAM). The majority of the computational tools developed by ORNL/NEAS have been included within the SCALE modular code system (SCALE 1995). This code system has been used throughout the world for the evaluation of nuclear facility and package designs. With this development and application experience as a basis, this paper highlights a number of criticality safety and shielding analysis issues that confront the designer and reviewer of a new RAM package. Changes in the types and quantities of material that need to be shipped will keep these issues before the technical community and provide challenges to future package design and certification.
Date: December 31, 1995
Creator: Parks, C.V. & Broadhead, B.L.
Partner: UNT Libraries Government Documents Department

Issues related to criticality safety analysis for burnup credit applications

Description: Spent fuel transportation and storage cask designs based on a burnup credit approach must consider issues that are not relevant in casks designed under a fresh fuel loading assumption. Parametric analyses are required to characterize the importance of fuel assembly and fuel cycle parameters on spent fuel composition and reactivity. Numerical models are evaluated to determine the sensitivity of criticality safety calculations to modeling assumptions. This paper discusses the results of studies to determine the effect of two important modeling assumptions on the criticality analysis of pressurized-water reactor (PWR) spent fuel: (1) the effect of assumed burnup history (i.e., specific power during and time-dependent variations in operational power) during depletion calculations, and (2) the effect of axial burnup distributions on the neutron multiplication factor calculated for a three-dimensional (3-D) conceptual cask design.
Date: December 1, 1995
Creator: DeHart, M.D. & Parks, C.V.
Partner: UNT Libraries Government Documents Department

End effects in the criticality analysis of burnup credit casks

Description: A study to evaluate the effect of axially dependent burnup on k{sub eff} has been performed as part of an effort to qualify procedures to be used in establishing burnup credit in shipping cask design and certification. This study was performed using a generic 31-element modular cast-iron cask (wall thickness 33.1 cm) with a 1-cm-thick borated stainless-steel basket for reactivity control. Fuel isotopics used here are those of the 17 {times} 17 Westinghouse assemblies from the North Anna Unit 1 reactor. Virginia Power (VP) provided detailed spatial isotopics for the fuel assemblies in-core at beginning-of-cycle 5 (BOC-5) as generated from their PDQ analyses. Twenty-two axial planes were defined in the original VP data. The isotopics used in this study were for a 3.41 initial wt % {sup 235}U and an average burnup of 31.5 GWd/MTU.
Date: January 1, 1990
Creator: Brady, M.C. & Parks, C.V.
Partner: UNT Libraries Government Documents Department

Reactor critical benchmark calculations for burnup credit applications

Description: In the criticality safety analyses for the development and certification of spent fuel casks, the current approach requires the assumption of fresh fuel'' isotopics. It has been shown that the removal of the fresh fuel'' assumption and the use of spent fuel isotopics ( burnup credit'') greatly increases the payload of spent fuel casks by reducing the reactivity of the fuel. Regulatory approval of burnup credit and the requirements of ANSI/ANS 8.1 specify that calculational methods for away-from-reactor criticality safety analyses be validated against experimental measurements. Criticality analyses for low-enriched lattices of fuel pins using the fresh fuel isotopics'' assumption have been widely benchmarked against applicable critical experiments. However, the same computational methods have not been benchmarked against criticals containing spent fuel because of the non-existence of spent fuel critical experiments. Commercial reactors offer an excellent and inexhaustible source of critical configurations against which criticality analyses can be benchmarked for spent fuel configurations. This document provides brief descriptions of the benchmarks and the computational methods for the criticality analyses. 8 refs., 1 fig., 1 tab.
Date: January 1, 1990
Creator: Renier, J.P. & Parks, C.V.
Partner: UNT Libraries Government Documents Department

Differential sensitivity theory applied to movement of maxima responses. [LMFBR]

Description: Differential sensitivity theory (DST) is a recently developed methodology to evaluate response derivatives dR/d..cap alpha.. by using adjoint functions which correspond to the differentiated (with respect to an arbitrary parameter ..cap alpha..) linear or nonlinear physical system of equations. However, for many problems, where responses of importance are local maxima such as peak temperature, power, or heat flux, changes in the phase space location of the peak itself are of interest. This summary will present the DST procedure for predicting phase space shifts of maxima responses as applied to the MELT-III fast reactor safety code. An FFTF protected transient involving a $.23/s ramp reactivity insertion with scram on high power was selected for investigation.
Date: January 1, 1981
Creator: Maudlin, P.J.; Parks, C.V. & Cacuci, D.G.
Partner: UNT Libraries Government Documents Department

Radiation dose rates from consolidated fuel in current generation shipping casks

Description: As part of a DOE study of public risk from cask transport of consolidated and unconsolidated spent fuel cooled beyond five years, radiation dose rates from two current generation shipping casks were evaluated. The IF300 rail shipping cask (with a capacity of seven PWR assemblies) and NLI 1/2 truck shipping cask (with a capacity of one PWR assembly) were selected as appropriate cask models. A Westinghouse 17 x 17 fuel assembly with a 3.3 wt % /sup 235/U enrichment, operated at a specific power of 37.5 MW/MTU for 880 full-power days (20% downtime in history), and discharged at a burnup of 33 GWD/MTU was specified. Results for spent fuel cooling times of 5, 10, 15, and 25 years are reported here.
Date: January 1, 1985
Creator: Parks, C.V.; Hermann, O.W. & Knight, J.R.
Partner: UNT Libraries Government Documents Department

Parametric study of radiation dose rates from rail and truck spent fuel transport casks

Description: Neutron and gamma dose rates from typical rail and truck spent fuel transport casks are reported for a variety of spent PWR fuel sources and cask conditions. The IF 300 rail cask and NLI 1/2 truck cask were selected for use as appropriate cask models. All calculations (cross section preparation, generation of spent fuel source terms, radiation transport calculations, and dose evaluation) were performed using various modules of the SCALE computational system. Conditions or parameters for which there were variations between cases include: detector distance from cask, spent fuel cooling time, the setting of fuel or neutron shielding cavities to either wet or dry, the cobalt content of assembly materials, normal fuel assemblies and consolidated cannisters, the geometry mesh interval size, and the order of the angular quadrature set. 13 refs., 6 figs., 9 tabs.
Date: August 1, 1985
Creator: Parks, C.V.; Hermann, O.W. & Knight, J.R.
Partner: UNT Libraries Government Documents Department

Thermal-hydraulic differential sensitivity theory. [LMFBR]

Description: Differential sensitivity theory is applied to a nonlinear, three-dimensional space- and time-dependent description of the thermal-hydraulic conservation equations. The resulting sensitivity equations, which are derived using adjoint functions, can be readily utilized for input parameter sensitivity analysis of large or long-running thermal-hydraulic computer codes without any prior engineering judgement. The procedure for applying this sensitivity theory is illustrated using several classical analytical problems.
Date: September 1, 1981
Creator: Maudlin, P.J.; Parks, C.V. & Weber, C.F.
Partner: UNT Libraries Government Documents Department

Application of differential sensitivity theory to transients with scram. [LMFBR]

Description: Differential sensitivity theory (DST) based on adjoint functions has been applied to various reactor safety problems. The most comprehensive application of DST sensitivity analysis has addressed the coupled thermal-hydraulic equations of the MELT-III fast reactor safety code, where a power ramp was imposed to eliminate the neutron point kinetics equations. In extending the above work to include realistic neutronic coupling, a DST procedure was developed for dealing with parameter discontinuities induced by dependent variables.
Date: January 1, 1980
Creator: Parks, C.V.; Maudlin, P.J. & Weber, C.F.
Partner: UNT Libraries Government Documents Department

ORIGEN-S (. cap alpha. ,n) neutron source spectra in borosilicate glass containing HLW (high-level waste)

Description: There is growing interest in the methodology and computational software for evaluating the (..cap alpha..,n) source spectra produced in mixtures of high-level waste (HLW) and borosilicate glass. The need for this development has been seen in previous work involving the analysis of HLW in borosilicate glass. Descriptions and applications of the ORIGEN-S method of computing neutron source spectra by both (..cap alpha..,n) reactions and spontaneous fission of UO/sub 2/ spent fuel have been reported previously. This summary presents a significant expansion of the ORIGEN-S (..cap alpha..,n) model to include ..cap alpha..-interactions with the light elements of borosilicate glass.
Date: January 1, 1987
Creator: Hermann, O.W.; Parks, C.V. & Ludwig, S.B.
Partner: UNT Libraries Government Documents Department

Automated shielding analysis sequences for spent fuel casks

Description: Two important Shielding Analysis Sequences (SAS) have recently been developed within the SCALE computational system. These sequences significantly enhance the existing SCALE system capabilities for evaluating radiation doses exterior to spent fuel casks. These new control module sequences (SAS1 and SAS4) and their capabilities are discussed and demonstrated, together with the existing SAS2 sequence that is used to generate radiation sources for spent fuel. Particular attention is given to the new SAS4 sequence which provides an automated scheme for generating and using biasing parameters in a subsequent Monte Carlo analysis of a cask.
Date: January 1, 1987
Creator: Tang, J.S.; Parks, C.V. & Hermann, O.W.
Partner: UNT Libraries Government Documents Department