Showing papers in "Transactions of the American Nuclear Society in 1996"

ATTILA: A three-dimensional, unstructured tetrahedral mesh discrete ordinates transport code

Abstract: Many applications of radiation transport require the accurate modeling of complex three-dimensional geometries. Historically, Monte Carlo codes have been used for such applications. Existing deterministic transport codes were not applied to such problems because of the difficulties of modeling complex three-dimensional geometries with rectangular meshes. The authors have developed a three-dimensional discrete ordinates (S{sub n}) code, ATTILA, which uses linear-discontinuous finite element spatial differencing in conjunction with diffusion-synthetic acceleration (DSA) on an unstructured tetrahedral mesh. This tetrahedral mesh capability enables the authors to efficiently model complex three-dimensional geometries. One interesting and challenging application of neutron and/or gamma-ray transport is nuclear well-logging applications. Nuclear well-logging problems usually involve a complex geometry with fixed sources and one or more detectors. Detector responses must generally be accurate to within {approx}1%. The combination of complex three-dimensional geometries and high accuracy requirements makes it difficult to perform logging problems with traditional S{sub n} differencing schemes and rectangular meshes. Hence, it is not surprising that deterministic S{sub n} codes have seen limited use in nuclear well-logging applications. The geometric modeling capabilities and the advanced spatial differencing of ATTILA give it a significant advantage, relative to traditional S{sub n} codes, for performing nuclear well-logging calculations.

The table of isotopes-8th edition and beyond

Abstract: A new edition of the Table of Isotopes has been published this year by John Wiley and Sons, Inc. This edition is the eighth in a series started by Glenn T. Seaborg in 1940. The two-volume, 3168-page, cloth-bound edition is twice the size of the previous edition published in 1978. It contains nuclear structure and decay data, based mainly on the Evaluated Nuclear Structure Data File (ENSDF), for >3100 isotopes and isomers. Approximately 24000 references are cited, and the appendices have been updated and extended. The book is packaged with an interactive CD-ROM that contains the Table of Isotopes in Adobe Acrobat Portable Document Format for convenient viewing on personal computer (PC) and UNIX workstations. The CD-ROM version contains a chart of the nuclides graphical index and separate indices organized for radioisotope users and nuclear structure physicists. More than 100000 hypertext links are provided to move the user quickly through related information free from the limitations of page size. Complete references with keyword abstracts are provided. The CD-ROM also contains the Table of Super-deformed Nuclear Bands and Fission Isomers; Tables of Atoms, Atomic Nuclei, and Subatomic Particles by Ivan P. Selinov; the ENSDF and nuclear structure reference (NSR) databases; themore » ENSDF manual by Jagdish K. Tuli; and Abode Acrobat Reader software.« less

Problems in the radon versus lung cancer test of the linear no-threshold theory and a procedure for resolving them

Abstract: It has been shown that lung cancer rates in U.S. counties, with or without correction for smoking, decrease with increasing radon exposure, in sharp contrast to the increase predicted by the linear-no-threshold (LNT) theory. The discrepancy is by 20 standard deviations, and very extensive efforts to explain it were not successful. Unless a plausible explanation for this discrepancy (or conflicting evidence) can be found, continued use of the LNT theory is a violation of {open_quotes}the scientific method.{close_quotes} Nevertheless, LNT continues to be accepted and used by all official and governmental organizations, such as the International Commission on Radiological Protection, the National Council on Radiation Protection and Measurements, the Council on Radiation Protection and Measurements, the National Academy of Sciences - U.S. Nuclear Regulatory Commission Board of Radiation Effects Research, Environmental Protection Agency etc., and there has been no move by any of these bodies to discontinue or limit its use. Assuming that they rely on the scientific method, this clearly implies that they have a plausible explanation for the discrepancy. The author has made great efforts to discover these `plausible explanations` by inquiries through various channels, and the purpose of this paper is to describe and discuss them.

Experimental investigation of filtered epithermal-photoneutron Beams for BNCT

Abstract: The Idaho National Engineering Laboratory (INEL) has been investigating the feasibility of a concept for an accelerator-based source of epithermal neutrons for BNCT that is based on the use of a two-stage photoneutron production process driven by an electron accelerator. In this concept, relativistic electron beams impinge upon heavily-shielded tungsten targets located at the outer radius of a small cylindrical tank of circulating heavy water (D{sub 2}0). A fraction of the energy of the electrons is converted in the tungsten targets into radially-inward-directed bremsstrahlung radiation. Neutrons subsequently generated by photodisintegration of deuterons in the D{sub 2}O within the tank are directed to the patient through a suitable beam tailoring system. Initial proof-of-principal tests using a low-current benchtop prototype of this concept have been conducted. Testing has included extensive measurements of the unfiltered photoneutron source as well as initial measurements of filtered epithermal-neutron spectra produced using two different advanced neutron filtering assemblies, as described here.

Site Restoration: Estimation of Attributable Costs From Plutonium-Dispersal Accidents

Abstract: A nuclear weapons accident is an extremely unlikely event due to the extensive care taken in operations. However, under some hypothetical accident conditions, plutonium might be dispersed to the environment. This would result in costs being incurred by the government to remediate the site and compensate for losses. This study is a multi-disciplinary evaluation of the potential scope of the post-accident response that includes technical factors, current and proposed legal requirements and constraints, as well as social/political factors that could influence decision making. The study provides parameters that can be used to assess economic costs for accidents postulated to occur in urban areas, Midwest farmland, Western rangeland, and forest. Per-area remediation costs have been estimated, using industry-standard methods, for both expedited and extended remediation. Expedited remediation costs have been evaluated for highways, airports, and urban areas. Extended remediation costs have been evaluated for all land uses except highways and airports. The inclusion of cost estimates in risk assessments, together with the conventional estimation of doses and health effects, allows a fuller understanding of the post-accident environment. The insights obtained can be used to minimize economic risks by evaluation of operational and design alternatives, and through development of improved capabilities for accident response.

PENTRAN: A three-dimensional scalable transport code with complete phase-space decomposition

Abstract: Three-dimensional S{sub N} transport modeling is gaining importance, with a rising need for detailed fluence calculations in geometries with inherent asymmetries. These calculations often cannot be adequately performed due to immense memory and storage requirements. New scalable, parallel discrete ordinates codes that take advantage of distributed memory parallel architectures are required. In general, scalable parallel algorithms minimize serial operations, maximize algorithm and data parallelism, and maximize computation to communication ratio. A leading issue in scalability is the minimization of serial input/output (I/O) in parallel algorithms. All I/O tasks should be handled efficiently by each individual processor. Further, memory-intensive arrays should be partitioned locally, distributed among processors. Regardless of parallel speedup, use of a local, partitioned memory structure is the only way that larger problems can be solved at all. In addition to using local memory allocation, efficient interprocess communicator groups for specific sets of processors should be used, allowing as needed message passing between necessary groups of processors, avoiding wasteful communications. Acceleration methods are needed because, without them, convergence can be difficult in some problems. Complete parallel decomposition of the problem should also be available; recent analysis of the transport source iteration indicates that significant parallel speedups are possible withmore » various decomposition strategies of the phase-space (angle, energy, and spatial dimensions) over a large processor array.« less

Validation of RETRAN-03 against a wide range of rod bundle void fraction data

Abstract: The work reported in this paper is a validation of the RETRAN-03 code`s ability to calculate the void fraction in a reactor core and so presents an assessment of the code against a wide range of void fraction measurements in rod bundles. The current version of RETRAN (RETRAN-03) uses the Chexal-Lellouche void fraction correlation as the basis for the calculation of the liquid vapor slip, either in the framework of a drift-flux model or as the basis of its dynamic slip model. The current assessment is based on data sets not included in the original correlation database. The data sets are from experimental facilities in Switzerland (NEPTUN-111), France (PERICLES), the United Kingdom (THETIS), and Japan: the Large-Scale Test Facility (LSTF) and the Two-Phase Flow Test Facility (TPTF).

High-fidelity, real-time simulation with RELAP5/NESTLE

Abstract: Recent advances in the computational speed of engineering workstations have enabled the development of a real-time version of the RELAP5 nuclear plant simulation code with laboratory discretionary research and development funding. In addition, the Idaho National Engineering Laboratory (INEL) is also funding the development of an enhanced real-time version of the existing three-dimensional nodal neutron kinetics package via its University Research Consortium (URC) at Purdue University and North Carolina State University (NCSU). This paper focuses on the enhancements to RELAP5 to achieve real-time performance.

Decommissioning of the thorium high-temperature reactor, THTR 300

Abstract: The pebble bed thorium high-temperature reactor THTR 300 was shut down on September 1, 1989, after > 16 000 h operation time. The THTR 300 is a project that is jointly sponsored by the Federal Republic of Germany, North Rhine-Westphalia, and Hochtemperatur-Kernkraftwerk GmbH. The public financers of this prototype reactor and the operators could not solve the financial problems for continued operation of this technically intact plant. The decommissioning decision had not been expected by the operator. This is why safe enclosure, the German term for SAFSTOR, turned out to be the only technical solution for quick decommissioning of the plant, apart from financial reasons and the nonavailability of a final storage facility. The plant is intended to be dismantled after {approximately}30 yr of safe enclosure, provided respective funds are available. The decommissioning was and will be performed in three steps, which are largely scheduled one after the other.

BNCT filter design studies for the ORNL Tower Shielding Facility

Abstract: Boron Neutron Capture Therapy (BNCT) in the United States has entered into a new phase with the initiation of clinical trials using neutron sources at the Brookhaven National Laboratory and the Massachusetts Institute of Technology. If these trials are successful at demonstrating the efficacy of BNCT as a viable treatment for glioblastoma multiforme, then there will be an immediate demand for several additional neutron sources in order to treat the several thousand patients currently diagnosed with glioblastomas in the U.S. each year. However, the requirements for an acceptable neutron source for BNCT are rather severe in terms of the need to provide a sufficient number of epithermal neutrons to a patient-accessible location in a reasonable time with minimal thermal-neutron, fast- neutron, and gamma-ray background. A recent study of potential neutron sources at Oak Ridge National Laboratory (ORNL) has been completed, which concludes that the Tower Shielding Facility (TSF), also appears very well suited for BNCT. The light-water-cooled reactor is contained in an aluminum pressure vessel and located in a large concrete `bunker` referred to as the Big Beam Shield (BBS). The BBS contains a 77-cm-diameter beam collimator, which permits access to a broad beam neutron flux exceeding 4 x 10[supmore » ll] Cm[sup -2]s[sup- 1] at the operational power of 1 MW. The collimated beam emerges horizontally onto an unenclosed test pad area on which shield mockups were assembled. The appropriate beam filter and collimator system can be easily constructed in the expansive area previously used for the large shield mockups. Additional engineering of the beam shutter mechanism and the construction of treatment support facilities will be needed but can be easily accommodated on the remote dedicated site. The filter design analysis is provided.« less

Adaptive kernel density estimation and Monte Carlo sampling

Abstract: The simulation of electron transport is computationally intensive. In principle, conventional (analog) Monte Carlo methods can be employed to analyze complex configurations to arbitrary accuracy, but in practice, the amount of computation is prohibitive for most applications. Condensed history Monte Carlo (CHMC) methods, which sample the outcome after a large number of electron collisions, have become the standard method for electron transport calculations but suffer well-known deficiencies. This report describes a method based on the kernel density estimator proposed as an alternative to histogramming.

An exponential discontinuous scheme for X-Y-Z geometry transport problems

Abstract: The recently developed exponential discontinuous spatial differencing scheme for the discrete-ordinate equations has been extended to x-y-z geometry with hexahedral cells. This scheme produces strictly positive angular fluxes given positive discrete-ordinate sources. The exponential discontinuous scheme has been developed and implemented into the three-dimensional, discrete-ordinate code. THREEDANT. Numerical results are given which show that the exponential discontinuous scheme is very accurate for deep-penetration transport problems with optically thick spatial meshes.

The MAAP code: What is it, what has it accomplished, and how can it be used in the future?

Abstract: The Modular Accident Analysis Program (MAAP) represents a substantial effort by the nuclear industry to provide the analytical capabilities for analyzing the response of light-water-cooled nuclear power plants to possible accident conditions. This capability has also been developed for both current and advanced light water reactor designs and has been applied to the Ontario Hydro CANDU reactors as well as the VVER Russian pressurized water reactor (PWR) designs. The most recent version, i.e., MAAP4, substantially improves the representation of the primary system and containment and also adds a dynamic graphical representation of the plant response through the use of MAAP4-GRAAPH.

Cyclotron production of molybdenum-99 via proton-induced uranium-238 fission

Abstract: Technetium-99m ({sup 99m}Tc; 6.02 h) is the most widely used radioisotope in nuclear medicine worldwide. It is currently supplied from elutions of a {sup 99}Mo {r_arrow} {sup 99m}Tc generator and used for regional distribution or locally for institutional use. The parent {sup 99}Mo (66.02 h) radioactivities are being produced commercially in reactors using the {sup 235}U(n, fission){sup 99}Mo (preferred method) or the {sup 98}Mo(n,{gamma}){sup 99}Mo (less desirable) methods. The production of {sup 99}Mo is based on the operation of a small number of nuclear reactors, most of which have reached decommissioning age. Two new reactors to be dedicated to radioisotope production are now being planned in Canada. Accelerator-based methods for producing {sup 99}Mo and/or {sup 99m}Tc would then provide new alternatives to the current reliance on reactor-based technologies.

MacNCTPlan: An improved Macintosh-based BNCT treatment planning program

Abstract: The physical and mathematical principles, architecture, operation, and application of the Monte Carlo-based boron neutron capture therapy (BNCT) treatment planning code NCTPLAN have been previously described. The original version of the NCTPLAN code, written in FORTRAN on a VAX platform under the VMS operating system, has been used to plan irradiations for four subjects who have completed a phase-1 BNCT protocol for melanoma of the extremities. With this background of experience in its use, a number of deficiencies were identified in this code which, to rectify, would have required an extensive rewrite. To modernize the code platform, to make the code more portable, and to facilitate future modifications, the code has been completely rewritten for the Power Macintosh platform using pascal units nested within the widely available image-processing code NIH Image. This paper describes the operation of MacNCTPlan.

Adjoint Monte Carlo methods for radiotherapy treatment planning

Abstract: For the past two decades, clinical procedures using highly collimated radiation beams, especially photons, have been used routinely. The main idea in radiation cancer therapy has been to maximize the dose in each point of the tumor without affecting the surrounding healthy tissue and especially the vital organs like the spine and the liver, using individually nonlethal beams that intersect at the tumor. Currently, the selection of the best set of beams (or fields) for a particular patient is determined by an iterative procedure that includes in each step a three-dimensional dose calculation for each beam configuration. The geometry is defined on the information obtained from the patient`s computed tomography or magnetic resonance imaging images. Current clinical dose calculation codes generally rely on semiempirical methods that are fast and work well for geometrically simple problems but are less accurate for practical, geometrically complex problems. The best-known method that can cope with that kind of physical and geometric complexity is the Monte Carlo method. However, to solve dose calculation problems with reasonable statistical errors in individual voxels, the needed computation time is excessively large. As a result, Monte Carlo codes are not routinely used for clinical treatment planning. In this paper,more » we sketch a new approach for the three-dimensional dose computations designed for radiotherapy treatment planning based on the adjoint Monte Carlo method. The proposed approach is more accurate than empirical techniques and has the potential to be faster than current methods.« less

Boron-10 analyses using prompt gamma neutron activation and ICP-AES

Abstract: Boron-10 quantification is a vital part of the phase I clinical trials that are in progress under the Massachusetts Institute of Technology (MIT)/New England Deaconess Hospital (NEDH) joint BNCT project. Accurate knowledge of the {sup 10}B content of blood as a function of time and immediately prior to irradiation is needed to calculate the total dose delivered to healthy tissue. The MIT/NEDH group continually seeks to improve and refine quantification methods to achieve greater sensitivity and faster analysis time. Rapid analysis is desirable so that irradiation can be initiated quickly, while high sensitivity will be needed to analyze very small samples. Brain tumor biopsies may be quite small, especially in cases where the tumor is not resectable. An upgraded version of the MIT prompt gamma neutron activation analysis (PGNAA) facility and an inductively coupled plasma atomic emission spectroscopy (ICP-AES) machine with a high-efficiency nebulizer (HEN) have been used to analygze {sup 10}B content in small samples down to the size range of needle biopsies ({approx} 0.03 ml).

Passive heat removal in the AP600 for LOCA and non-LOCA events

Abstract: The AP600 reactor is a pressurized water reactor being designed to utilize a natural circulation heat exchanger as one of the safety-grade means of removing primary system heat following certain design-basis events. The passive residual heat removal (PRHR) heat exchanger removes primary system heat to mitigate loss-of-coolant-accident (LOCA) and non-LOCA events and functionally replaces the safety-grade auxiliary feed-water system used in current operating plants. Tests were performed to characterize the PRHR heat exchanger performance and to provide a basis for heat transfer correlations to be used in the safety analysis of the AP600.

Theory of Westinghouse advanced nodal code for MOX applications

Abstract: The Westinghouse Advanced Nodal Code (ANC) has been modified to incorporate new theoretical improvements in its nodal solution method. These theoretical improvements include the employment of a hybrid of the Polynomial Nodal Expansion Method and the Analytic Nodal Expansion Method in solving the nodal diffusion equation and the use of the burnup gradient treatment to accurately capture uneven assembly burnup. In today`s environment, nuclear electric utilities, especially in Europe, are used to loading their cores with mixed oxides (MOX) of enrichments that give the same lifetime average reactivity of regular UO{sub 2} fuel assembly. This did not pose any difficulties for core design methodologies that are used to analyze those cores. However, the usage of weapons-grade plutonium, or even recycled plutonium with higher plutonium, in core design will challenge existing design methods. Also, there has been a strong inclination among utilities to run their reactors longer and burn their fuel higher to improve their fuel cycle economy. These two challenges and many others have put a lot of pressure on core analysis methods in their attempt to maintain the same superior performance that they enjoyed in the past. This paper outlines Westinghouse efforts in the area of nodal method development.more » In particular, ANC has several theoretical enhancements that enable accurate predictions of design parameters for cores that have mixed loadings of regular UO{sub 2} and MOX and/or high-burnup fuel. These enhancements include a consistent corner-point flux calculation for pin-power reconstruction and spectral interaction correction of assembly cross sections for accurate environment effects evaluation. The basic theoretical foundation of these enhancements is outlined in this paper« less

Implementation and assessment of improved models and options in TRAC-BF1

Abstract: Extensive assessment of the TRAC-BF1 code has revealed a number of shortcomings of the code in relation to the modeling of the heat transfer coefficient (HTC) during reflooding, the annular flow interfacial shear, as well as some numerics-related problems, and a number of modifications have already been reported and tested. In this work, the author summarizes all the final model changes and options implemented in the code, thereby extending and complementing the work reported. The author has demonstrated with sufficient confidence that the main modifications (the special wall-to-liquid HTC during reflooding, restoring the Wallis annular flow interfacial shear correlation, and removing the wall shear package of the code in favor of the one of TRAC-PF1) greatly improve the predictive capabilities of the code during the analysis of separate-effects bottom-flooding tests, benchmark top-flooding tests, and TLTA tests. The predictions of the modified code are not only close to the measurements but are also free of spurious and unphysical numerical oscillations.

Assessment of uncertainties for PWR pressure vessel surveillance-French Experience

Abstract: A characteristic of the French nuclear installations that differs from those in most other countries with an important nuclear industry is their high degree of standardization. Two main types of pressurized water reactors (PWRs) are the 900-MW(electric) CPY and the 1300-MW(electric) P4 reactors produced by a single manufacturer, Electricite de France (EdF). Loading schemes are very standardized, although greater diversification has been introduced in recent years due to implementation of some new loading schemes (LLLP, mixed oxide, extended fuel cycle). This report describes the pressure vessel surveillance program.

FORMOSA-P multiple adaptive penalty function methodology

Abstract: The nuclear fuel management optimization code FORMOSA-P has been developed to determine the family of near-optimum loading patterns for pressurized water reactors. The FORMOSA-P code utilizes the nonlinear stochastic optimization approach of simulated annealing, wherein an objective function is calculated to evaluate the suitability of the candidate loading patterns. During the course of a simulated annealing cooling cycle, the penalty function multiplier is adaptively increased to eliminate the penalty constraint violations by the end of the cooling cycle. In early versions of FORMOSA-P, only one adaptive penalty function, for power peaking, was available. All other design constraints were handled via true-false constraints. The introduction of generalized perturbation theory (GPT) error control which greatly increased the design space accessible in a single cooling cycle, and modern core designs, which are increasingly tightly constrained by multiple design constraints, however, have made these true-false constraints undesirable; thus, a methodology for inclusion of multiple adaptive penalty constraints has now been developed.

Clinical treatment planning of subjects undergoing boron neutron capture therapy

Abstract: The sequence of treatment planning operations for clinical boron neutron capture therapy (BNCT) in the New England Deaconess Hospital-Massachusetts Institute of Technology (MIT) BNCT program consists of multiple computational and experimental steps. The course of action for each subject has been categorized into several steps, each of which is introduced and summarized in this paper. Those steps are: 1. Calibration of epithermal neutron beam via in-phantom dose measurements and calibration of beam monitoring system; 2. Administration of test dose of BPA-f (boron compound) to subject; 3. Analyses of blood and tissue samples for boron content; 4. CT scanning of subject; 5. Calculation of treatment plan; 6. Irradiation of subject in the epithermal neutron beam at the MIT research reactor.

ANSI/ANS-8.7-1975(R87): Guide for nuclear criticality safety in the storage of fissile materials

Abstract: Basic parameters and guidance for the nuclear criticality safety control outside reactors are described in American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1-1983(R88), {open_quotes}Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.{close_quotes} Because ANS-8.1 provides general guidance and limited single-unit parameter limits, ANS standards subcommittees developed the current ANSI/ANS-8.7-1975, {open_quotes}Guide for Nuclear Criticality Safety in the Storage of Fissile Materials,{close_quotes} which was reaffirmed in 1987. The standard is being revised, and the new revision is to be completed by December 1997. It is therefore appropriate that the standard and its technical bases be summarized, the applications highlighted, and the plans for further development explained.

Development of an advanced boron injection tank

Abstract: Mitsubishi has developed a hybrid safety system. This is an optimum combination of active and passive safety systems that provides improved safety, higher reliability, and better economy. As one option of the passive safety systems, Mitsubishi is studying a passive boron injection system that uses an advanced boron injection tank (BIT). The boron injection system to be developed in this study is passive and does not use nitrogen gas as a driving force. These features realize the higher reliability and eliminate a bad influence of the nitrogen gas during natural circulation cooling in the reactor coolant system (RCS). The driving force of the boric acid water injection in our advanced BIT is the boiling and steam expansion due to the depressurization inside the tank. Mitsubishi carried out tests to verify that the injection mechanism of the advanced BIT is basically feasible.

ANSI/ANS-8.15-1981(R87): Nuclear criticality control of special actinide elements

Abstract: The American National Standard, {open_quotes}Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactotors{close_quotes} American National Standards Institute/American Nuclear Society (ANSI/ANS)-8.1-1983(R88) provides guidance for the nuclides {sup 233}U, {sup 235}U, and {sup 239}Pu. These three nuclides are of primary interest in out-of-reactor criticality safety since they are the most commonly encountered in the vast majority of operations. However, some operations can involve nuclides other than {sup 233}U, {sup 235}U, and {sup 239}Pu in sufficient quantities that their effect on criticality safety could be of concern. ANSI/ANS-8.15-1981(R87) {open_quotes}Nuclear Criticality Control of Special Actinide Elements,{close_quotes} provides guidance for 15 such nuclides. The standard was approved for use on November 9, 1981. When it received its first 5-yr review, no changes were made, and it was reaffirmed effective October 30, 1987. The standard was again reviewed and reaffirmed without changes in December 1995. The next 5-yr review of the standard is due in December 2000. The affected nuclides are {sup 237}Np, {sup 238}Pu, {sup 240}Pu, {sup 242}Pu, {sup 241}Am, {sup 243}Am, {sup 244}Cm, {sup 239}Pu, {sup 241}Pu, {sup 242m}Am, {sup 243}Cm, {sup 245}Cm, {sup 247}Cm, {sup 249}Cf, and {sup 251}Cf.

TGBLA/PANACEA and CASMO/MICROBURN analyses of GE9B/GE10 fuel in the Quad Cities

Abstract: Quad Cities station is a dual-unit site with boiling water reactor (BWR)/3 reactors GE913 fuel was loaded into cycles 12 and 13 of Quad Cities unit 1 and cycles 11, 12, and 13 of unit 2 GE10 fuel was introduced in cycle 14 of both units In 1996 Quad Cities will be transitioning from General Electric (GE) fuel to Siemens Power Corporation (SPC) fuel An extensive benchmark program has been underway at Commonwealth Edison Company (ComEd) to support this transition to SPC fuel and analytical methods Results of this benchmark for hot and cold eigenvalue trends are presented for the TGBLA04V/PANAC08V and CASMO-3G/MICROBURN code packages Comparisons of calculated thermal limits between the off-line simulator and the plant process computer are also presented for the TGBLA04V/PANAC08V package

Uncertainties in transport theory pressure vessel neutron fluence calculations

Abstract: The desire for increasing safety margins (increased power production and plant availability), the fact that many existing plants are approaching the final years of their design lifetime, and the consideration of life extension have resulted in the need for performing more accurate estimation of parameters that affect plant performance and safety. This means that more accurate calculational and experimental methodologies for estimating the neutron and/or gamma fluence distributions are needed to ensure the structural integrity of the pressure vessel in pressurized water reactors (PWRs) and vessel internals in boiling water reactors. Such an estimation requires the use of particle transport theory methods and analysis; because of the existing uncertainties in the input parameters and numerical methods, it is necessary to estimate the uncertainties in the calculated results.

Modeling of water hammer with column separation

Abstract: The objective of this paper is to compare the accuracy of different two-phase flow codes and their capability of fast transient prediction Water-hammer transient following instantaneous closure of a valve is a useful benchmark for two-phase flow codes for comparison with the available experimental results Important phenomena are pressure waves, flashing near the valve, and condensation of the steam

Safety culture in nuclear installations: Summary of an international topical meeting

Abstract: An international topical meeting, Safety Culture in Nuclear Installations, was organized by the American Nuclear Society (ANS) Austria Local Section, cosponsored by the ANS Nuclear Reactor Safety and Human Factors Divisions in cooperation with the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (NEA/OECD) and held in Vienna April 24-28, 1995. Some 250 experts from 30 different countries and organizations took part in the 85 paper presentations and two workshops. The concept of safety culture was initially used in the first International Nuclear Safety Advisory Group (INSAG) report on the Chernobyl accident analysis report in 1986. Although some elements of safety culture have been used over the years in nuclear safety activities, the new phrase safety culture and the concept were found interesting as highlighting the `soft` aspects of safety and as encompassing more than human errors. Unfortunately, for many years it was used more in the way of identifying lack of safety culture. Conscious of this application, INSAG further developed the safety culture concept in the INSAG 4 report: The report contains a definition, the universal aspects of safety culture, the two main components of safety culture management and individual behaviour, and performance indicators of amore » good safety culture. This report is now quite famous and adopted with some additions or complementary definitions by many institutes and organizations for their daily activities.« less