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Sample records for rensselaer critical facility

  1. Reactor Sharing at Rensselaer Critical Facility

    International Nuclear Information System (INIS)

    D. Steiner, D. Harris, T. Trumbull

    2006-01-01

    This final report summarizes the reactor sharing activities at the Rensselaer Critical Facility. An example of a typical tour is also included. Reactor sharing at the RCF brings outside groups into the facility for a tour, an explanation of reactor matters, and a reactor measurement. It has involved groups ranging from high school classes to advanced college groups and in size from a few to about 50 visitors. The RCF differs from other university reactors in that its fuel is like that of large power reactors, and its research and curriculum are dedicated to power reactor matters

  2. Sharing of Rensselaer Polytechnic Institute Reactor Critical Facility (RCF)

    International Nuclear Information System (INIS)

    1995-01-01

    The RPI Reactor Critical Facility (RCF) operated successfully over the period fall 1994 - fall 1995. During this period, the RCF was used for Critical Reactor Laboratory spring 1995 (12 students); Reactor Operations Training fall 1994 (3 students); Reactor Operations Training spring 1995 (3 students); and Reactor Operations Training fall 1995 (3 students). Thirty-two Instrumentation and Measurement students used the RCF for one class for hands-on experiments with nuclear instruments. In addition, a total of nine credits of PhD thesis work were carried out at the RCF. This document constitutes the 1995 Report of the Rensselaer Polytechnic Institute's Reactor Critical Facility (RCF) to the USNRC, to the USDOE, and to RPI management

  3. Nuclear blenders: blended learning from Rensselaer's Reactor Critical Facility

    International Nuclear Information System (INIS)

    Haley, T.C.

    2011-01-01

    Rensselaer's senior level undergraduate nuclear engineering course 'Critical Reactor Laboratory' is highly regarded and much loved. If you can get in, that is. But now it's a required course, nuclear engineering enrollment is up, and others are knocking on our door to get in. How might one offer such a unique course to the masses, without losing the whole point of a laboratory experience? This presentation looks at the costs and benefits of the transition to a 'blended learning' mode -- the merging of traditional, face-to-face instruction and web-based instruction as a solution. As part of the presentation, the course and the facility will be highlighted by short excepts from the 50 minute movie 'Everything You Always Wanted to Know about Neutron Chain Reactions (but were afraid to ask)'.

  4. Safety Evaluation Report, related to the renewal of the operating license for the critical experiment facility of the Rensselaer Polytechnic Institute (Docket No. 50-225)

    International Nuclear Information System (INIS)

    1983-10-01

    This Safety Evaluation Report for the application filed by the Rensselaer Polytechnic Institute (RPI) for a renewal of operating license CX-22 to continue to operate a critical experiment facility has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by Rensselaer Polytechnic Institute and is located at a site in the city of Schenectady, New York. The staff concludes that this critical facility can continue to be operated by RPI without endangering the health and safety of the public

  5. Education and training at the Rensselaer Polytechnic Institute reactor critical facility

    International Nuclear Information System (INIS)

    Harris, D.R.

    1989-01-01

    The Rensselaer Polytechnic Institute (RPI) Reactor Critical Facility (RCF) has provided hands-on education and training for RPI and other students for almost a quarter of a century. The RCF was built in the 1950s by the American Locomotive Company (ALCO) as a critical facility in which to carry out experiments in support of the Army Package power Reactor (APPR) program. A number of APPRs were built and operated. In the middle 1960s, ALCO went out of business and provided the facility to RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although limited amounts of reactor research, activation analysis, and reactivity assays have been carried out as well. Recently, a U.S. Department of Energy (DOE) upgrade program supported refueling the RCF with 4.81 wt% enriched UO 2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective approach to conversion from high-enrichment bombgrade fuel to low-enrichment fuel. More important, however, is the fact that the new fuel is of current interest for light water power reactors with extended lifetime fuel. Thus, not only are critical reactor experiments being carried out on the fuel but, more importantly, the quality of the education and training has been enhanced

  6. Nuclear blenders: blended learning from Rensselaer's Reactor Critical Facility

    Energy Technology Data Exchange (ETDEWEB)

    Haley, T.C. [Rensselaer Polytechnic Inst., Troy, NY (United States)

    2011-07-01

    Rensselaer's senior level undergraduate nuclear engineering course 'Critical Reactor Laboratory' is highly regarded and much loved. If you can get in, that is. But now it's a required course, nuclear engineering enrollment is up, and others are knocking on our door to get in. How might one offer such a unique course to the masses, without losing the whole point of a laboratory experience? This presentation looks at the costs and benefits of the transition to a 'blended learning' mode -- the merging of traditional, face-to-face instruction and web-based instruction as a solution. As part of the presentation, the course and the facility will be highlighted by short excepts from the 50 minute movie 'Everything You Always Wanted to Know about Neutron Chain Reactions (but were afraid to ask)'.

  7. Combined use of the RPI [Rensselaer Polytechnic Institute] reactor for training and critical experiments

    International Nuclear Information System (INIS)

    Harris, D.R.; Rohr, R.R.; Rodriguez-Vera, F.

    1990-01-01

    The Rensselaer Polytechnic Institute (RPI) reactor critical facility (RCF) has provided educational and research opportunities for RPI and other students for >25 yr. The RCF was built by the American Locomotive Company (ALCO) in the 1950s as a critical facility in support of the army package power reactor program, and, when ALCO went out of business in 1964, the RCF was acquired by RPI. Since that time, RPI has operated the RCF primarily in a teaching mode in the nuclear engineering department, although reactor research, activation analyses, and reactivity assays have been carried out as well. Until recently, the RCF was fueled by plates containing highly enriched uranium as a cermet in stainless steel. This highly enriched uranium (HEU) fuel was replaced recently by 4.81 wt% enriched UO 2 high-density pellets clad in stainless steel rods. The use of these SPERT (F1) fuel rods in the RCF provided a cost-effective method for conversion of the core from HEU to low-enriched uranium and for enhancement of the RCF training and research program. The RCF is the only facility in the United States that provides reactor training on a core containing fuel that is similar to that used in power industry light water reactors (LWRs). Moreover, the RCF is the only facility in the United States currently available for supplying critical experimental data in support of the LWR power industry. Thus, the RCF is in a unique position to carry out important training and research services consistent with RPI's nuclear engineering objectives

  8. Sharing of the RPI Reactor Critical Facility (RCF). Final summary report, January 1988--September 1995

    International Nuclear Information System (INIS)

    Harris, D.R.

    1995-01-01

    Rensselaer Polytechnic Institute (RPI) has participated for a number of years in Sharing of the Reactor Critical Facility (RCF) under the U.S. Department of Energy University Reactor Sharing Program. In September of each year a Sharing invitation is sent to 92 public and private high schools and to 74 colleges and universities within about a 3 hour drive to the RCF (Appendix B). Each year about 10 such educational institutions send groups to share the RCF

  9. Critical experiments with mixed oxide fuel

    International Nuclear Information System (INIS)

    Harris, D.R.

    1997-01-01

    This paper very briefly outlines technical considerations in performing critical experiments on weapons-grade plutonium mixed oxide fuel assemblies. The experiments proposed would use weapons-grade plutonium and Er 2 O 3 at various dissolved boron levels, and for specific fuel assemblies such as the ABBCE fuel assembly with five large water holes. Technical considerations described include the core, the measurements, safety, security, radiological matters, and licensing. It is concluded that the experiments are feasible at the Rensselaer Polytechnic Institute Reactor Critical Facility. 9 refs

  10. Dedicated Laboratory Setup for CO2 TEA Laser Propulsion Experiments at Rensselaer Polytechnic Institute

    International Nuclear Information System (INIS)

    Salvador, Israel I.; Kenoyer, David; Myrabo, Leik N.; Notaro, Samuel

    2010-01-01

    Laser propulsion research progress has traditionally been hindered by the scarcity of photon sources with desirable characteristics, as well as integrated specialized flow facilities in a dedicated laboratory environment. For TEA CO 2 lasers, the minimal requirements are time-average powers of >100 W), and pulse energies of >10 J pulses with short duration (e.g., 0.1 to 1 μs); furthermore, for the advanced pulsejet engines of interest here, the laser system must simulate pulse repetition frequencies of 1-10 kilohertz or more, at least for two (carefully sequenced) pulses. A well-equipped laser propulsion laboratory should have an arsenal of sensor and diagnostics tools (such as load cells, thrust stands, moment balances, pressure and heat transfer gages), Tesla-level electromagnet and permanent magnets, flow simulation facilities, and high-speed visualization systems, in addition to other related equipment, such as optics and gas supply systems. In this paper we introduce a cutting-edge Laser Propulsion Laboratory created at Rensselaer Polytechnic Institute, one of the very few in the world to be uniquely set up for beamed energy propulsion (BEP) experiments. The present BEP research program is described, along with the envisioned research strategy that will exploit current and expanded facilities in the near future.

  11. Critical experiments facility and criticality safety programs at JAERI

    International Nuclear Information System (INIS)

    Kobayashi, Iwao; Tachimori, Shoichi; Takeshita, Isao; Suzaki, Takenori; Miyoshi, Yoshinori; Nomura, Yasushi

    1985-10-01

    The nuclear criticality safety is becoming a key point in Japan in the safety considerations for nuclear installations outside reactors such as spent fuel reprocessing facilities, plutonium fuel fabrication facilities, large scale hot alboratories, and so on. Especially a large scale spent fuel reprocessing facility is being designed and would be constructed in near future, therefore extensive experimental studies are needed for compilation of our own technical standards and also for verification of safety in a potential criticality accident to obtain public acceptance. Japan Atomic Energy Research Institute is proceeding a construction program of a new criticality safety experimental facility where criticality data can be obtained for such solution fuels as mainly handled in a reprocessing facility and also chemical process experiments can be performed to investigate abnormal phenomena, e.g. plutonium behavior in solvent extraction process by using pulsed colums. In FY 1985 detail design of the facility will be completed and licensing review by the government would start in FY 1986. Experiments would start in FY 1990. Research subjects and main specifications of the facility are described. (author)

  12. Construction of STACY (Static Experiment Critical Facility)

    International Nuclear Information System (INIS)

    Murakami, Kiyonobu; Onodera, Seiji; Hirose, Hideyuki

    1998-08-01

    Two critical assemblies, STACY (Static Experiment Critical Facility) and TRACY (Transient Experiment Critical Facility), were constructed in NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) to promote researches on the criticality safety at a reprocessing facility. STACY aims at providing critical data of uranium nitrate solution, plutonium nitrate solution and their mixture while varying concentration of solution fuel, core tank shape and size and neutron reflecting condition. STACY achieved first criticality in February 1995, and passed the licensing inspection by STA (Science and Technology Agency of Japan) in May. After that a series of critical experiments commenced with 10 w/o enriched uranium solution. This report describes the outline of STACY at the end of FY 1996. (author)

  13. Nobel Prize winning physicist to speak at Rensselaer Nov. 20 Leon Lederman to discuss pre college science education

    CERN Multimedia

    2002-01-01

    Leon Lederman, Nobel Prize-winner, will offer some radical ideas for improving pre-college science education when he delivers the annual Robert Resnick Lecture at Rensselaer Polytechnic Institute's Russell Sage Laboratory (1/2 page).

  14. Construction of new critical experiment facilities in JAERI

    International Nuclear Information System (INIS)

    Takeshita, Isao; Itahashi, Takayuki; Ogawa, Kazuhiko; Tonoike, Kotaro; Matsumura, Tatsuro; Miyoshi, Yoshinori; Nakajima, Ken; Izawa, Naoki

    1995-01-01

    Japan Atomic Energy Research Institute (JAERI) has promoted the experiment research program on criticality safety since early in 1980s and two types of new critical facilities, Static Experiment Critical Facility (STACY) and Transient Experiment Critical Facility (TRACY) were completed on 1994 in Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) of JAERI Tokai Research Establishment. STACY was designed so as to obtain critical mass data of low enriched uranium and plutonium solution which is extensively handled in LWR fuel reprocessing plant. TRACY is the critical facility where critical accident phenomenon is demonstrated with low enriched uranium nitrate solution. For criticality safety experiments with both facilities, the Fuel Treatment System is attached to them, where composition and concentration of uranium and plutonium nitrate solutions are widely varied so as to obtain experiments data covering fuel solution conditions in reprocessing plant. Design performances of both critical facilities were confirmed through mock-up tests of important components and cold function tests. Hot function test has started since January of 1995 and some of the results on STACY are to be reported. (author)

  15. SRTC criticality technical review: Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility

    International Nuclear Information System (INIS)

    Rathbun, R.

    1993-01-01

    Separate review of NMP-NCS-930058, open-quotes Nuclear Criticality Safety Evaluation 93-18 Uranium Solidification Facility's Waste Handling Facility (U), August 17, 1993,close quotes was requested of SRTC Applied Physics Group. The NCSE is a criticality assessment to determine waste container uranium limits in the Uranium Solidification Facility's Waste Handling Facility. The NCSE under review concludes that the NDA room remains in a critically safe configuration for all normal and single credible abnormal conditions. The ability to make this conclusion is highly dependent on array limitation and inclusion of physical barriers between 2x2x1 arrays of boxes containing materials contaminated with uranium. After a thorough review of the NCSE and independent calculations, this reviewer agrees with that conclusion

  16. Criticality accident of nuclear fuel facility. Think back on JCO criticality accident

    International Nuclear Information System (INIS)

    Naito, Keiji

    2003-09-01

    This book is written in order to understand the fundamental knowledge of criticality safety or criticality accident of nuclear fuel facility by the citizens. It consists of four chapters such as critical conditions and criticality accident of nuclear facility, risk of criticality accident, prevention of criticality accident and a measure at an occurrence of criticality accident. A definition of criticality, control of critical conditions, an aspect of accident, a rate of incident, damage, three sufferers, safety control method of criticality, engineering and administrative control, safety design of criticality, investigation of failure of safety control of JCO criticality accident, safety culture are explained. JCO criticality accident was caused with intention of disregarding regulation. It is important that we recognize the correct risk of criticality accident of nuclear fuel facility and prevent disasters. On the basis of them, we should establish safety culture. (S.Y.)

  17. The NBER-Rensselaer Scientific Papers Database: Form, Nature, and Function. NBER Working Paper No. 14575

    Science.gov (United States)

    Adams, James D.; Clemmons, J. Roger

    2008-01-01

    This article is a guide to the NBER-Rensselaer Scientific Papers Database, which includes more than 2.5 million scientific publications and over 21 million citations to those papers. The data cover an important sample of 110 top U.S. universities and 200 top U.S.-based R&D-performing firms during the period 1981-1999. This article describes the…

  18. Criticality safety and facility design considerations

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1991-06-01

    Operations with fissile material introduce the risk of a criticality accident that may be lethal to nearby personnel. In addition, concerns over criticality safety can result in substantial delays and shutdown of facility operations. For these reasons, it is clear that the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The emphasis of this report will be placed on engineering design considerations in the prevention of criticality. The discussion will not include other important aspects, such as the physics of calculating limits nor criticality alarm systems

  19. Los Alamos Critical Experiments Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-01-01

    The Critical Experiments Facility of the Los Alamos National Laboratory has been in existence for 45 years. In that period of time, thousands of measurements have been made on assemblies containing every fissionable material in various configurations that included bare metal and compounds of the nitrate, sulfate, fluoride, carbide, and oxide. Techniques developed or applied include Rossi-α, source-jerk, rod oscillator, and replacement measurements. Many of the original measurements of delay neutrons were performed at the site, and a replica of the Hiroshima weapon was operated at steady state to assist in evaluating the relative biological effectiveness (RBE) of neutrons. Solid, liquid, and gas fissioning systems were run at critical. Operation of this original critical facility has demonstrated the margin of safety that can be obtained through remote operation. Eight accidental excursions have occurred on the site, ranging from 1.5 x 10 16 to 1.2 x 10 17 fissions, with no significant exposure to personnel or damage to the facility beyond the machines themselves -- and in only one case was the machine damaged beyond further use. The present status of the facility, operating procedures, and complement of machines will be described in the context of programmatic activity. New programs will focus on training, validation of criticality alarm systems, experimental safety assessment of process applications, and dosimetry. Special emphasis will be placed on the incorporation of experience from 45 years of operation into present procedures and programs. 3 refs

  20. STACY and TRACY: nuclear criticality experimental facilities under construction

    International Nuclear Information System (INIS)

    Kobayashi, I.; Takeshita, I.; Yanagisawa, H.; Tsujino, T.

    1992-01-01

    Japan Atomic Energy Research Institute is constructing a Nuclear Fuel Cycle Safety Engineering Research Facility, NUCEF, where the following research themes essential for evaluating safety problems relating to back-end technology in nuclear fuel cycle facilities will be studied: nuclear criticality safety research; research on advanced reprocessing processes and partitioning; and research on transuranic waste treatment and disposal. To perform nuclear criticality safety research related to the reprocessing of light water reactor spent fuels, two criticality experimental facilities, STACY and TRACY, are under construction. STACY (Static Criticality Facility) will be used for the study of criticality conditions of solution fuels, uranium, plutonium and their mixtures. TRACY (Transient Criticality Facility) will be used to investigate criticality accident phenomena with uranium solutions. The construction progress and experimental programmes are described in this Paper. (author)

  1. DRY TRANSFER FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation updates the previous criticality evaluation for the fuel handling, transfer, and staging operations to be performed in the Dry Transfer Facility (DTF) including the remediation area. The purpose of the calculation is to demonstrate that operations performed in the DTF and RF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Dry Transfer Facility Description Document'' (BSC 2005 [DIRS 173737], p. 3-8). A description of the changes is as follows: (1) Update the supporting calculations for the various Category 1 and 2 event sequences as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2005 [DIRS 171429], Section 7). (2) Update the criticality safety calculations for the DTF staging racks and the remediation pool to reflect the current design. This design calculation focuses on commercial spent nuclear fuel (SNF) assemblies, i.e., pressurized water reactor (PWR) and boiling water reactor (BWR) SNF. U.S. Department of Energy (DOE) Environmental Management (EM) owned SNF is evaluated in depth in the ''Canister Handling Facility Criticality Safety Calculations'' (BSC 2005 [DIRS 173284]) and is also applicable to DTF operations. Further, the design and safety analyses of the naval SNF canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. Also, note that the results for the Monitored Geologic Repository (MGR) Site specific Cask (MSC) calculations are limited to the

  2. Criticality Safety Evaluation of Hanford Tank Farms Facility

    Energy Technology Data Exchange (ETDEWEB)

    WEISS, E.V.

    2000-12-15

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste.

  3. Criticality Safety Evaluation of Hanford Tank Farms Facility

    International Nuclear Information System (INIS)

    WEISS, E.V.

    2000-01-01

    Data and calculations from previous criticality safety evaluations and analyses were used to evaluate criticality safety for the entire Tank Farms facility to support the continued waste storage mission. This criticality safety evaluation concludes that a criticality accident at the Tank Farms facility is an incredible event due to the existing form (chemistry) and distribution (neutron absorbers) of tank waste. Limits and controls for receipt of waste from other facilities and maintenance of tank waste condition are set forth to maintain the margin subcriticality in tank waste

  4. Criticality management of Tokai reprocessing facility

    Energy Technology Data Exchange (ETDEWEB)

    Nojiri, Ichiro [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan)

    2001-01-01

    In fuel cycle centers a number of equipment and vessels of various types and of complex design are used in several processes, i.e. dissolution of spent fuels, separation and storage of uranium and plutonium from fission products and transuranium elements. For each processes, Monte Carlo codes are frequently applied to manage the fuel criticality. Safety design depends largely on specific features of each facilities. The present report describes status of criticality management for main processes in Tokai Reprocessing Facility, JNC, and the criticality conditions specifically existing there. The guiding principle throughout consists of mass control, volume control, design (form) control, concentration control, and control due to employment of neutron poisons. (S. Ohno)

  5. Criticality safety considerations. Integral Monitored Retrievable Storage (MRS) Facility

    International Nuclear Information System (INIS)

    1986-09-01

    This report summarizes the criticality analysis performed to address criticality safety concerns and to support facility design during the conceptual design phase of the Monitored Retrievable Storage (MRS) Facility. The report addresses the criticality safety concerns, the design features of the facility relative to criticality, and the results of the analysis of both normal operating and hypothetical off-normal conditions. Key references are provided (Appendix C) if additional information is desired by the reader. The MRS Facility design was developed and the related analysis was performed in accordance with the MRS Facility Functional Design Criteria and the Basis for Design. The detailed description and calculations are documented in the Integral MRS Facility Conceptual Design Report. In addition to the summary portion of this report, explanatary notes for various terms, calculation methodology, and design parameters are presented in Appendix A. Appendix B provides a brief glossary of technical terms

  6. Four critical facilities: their capabilities and programs

    International Nuclear Information System (INIS)

    Whitesides, G.E.

    1980-01-01

    Information is presented on the critical experiments facilities at Babcock and Wilcox, Lynchburg, Virginia; at Battelle Pacific Northwest Laboratory in Hanford, Washington; at Rockwell-International in Rocky Flats, Colorado; and at Los Alamos Scientific Laboratory in New Mexico. It is noted that the critical mass facilities which still exist in this country represent a bare minimum for maintaining a measurement program sufficient for meeting data requirements

  7. Earthquake research for the safer siting of critical facilities

    Energy Technology Data Exchange (ETDEWEB)

    Cluff, J.L. (ed.)

    1980-01-01

    The task of providing the necessities for living, such as adequate electrical power, water, and fuel, is becoming more complicated with time. Some of the facilities that provide these necessities would present potential hazards to the population if serious damage were to occur to them during earthquakes. Other facilities must remain operable immediately after an earthquake to provide life-support services to people who have been affected. The purpose of this report is to recommend research that will improve the information available to those who must decide where to site these critical facilities, and thereby mitigate the effects of the earthquake hazard. The term critical facility is used in this report to describe facilities that could seriously affect the public well-being through loss of life, large financial loss, or degradation of the environment if they were to fail. The term critical facility also is used to refer to facilities that, although they pose a limited hazard to the public, are considered critical because they must continue to function in the event of a disaster so that they can provide vital services.

  8. Nuclear criticality safety program at the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Fujita, E.K.; Tracy, D.B.; Klann, R.T.; Imel, G.R.; Benedict, R.W.; Rigg, R.H.

    1994-01-01

    The Fuel Cycle Facility (FCF) is designed to demonstrate the feasibility of a novel commercial-scale remote pyrometallurgical process for metallic fuels from liquid metal-cooled reactors and to show closure of the Integral Fast Reactor (IFR) fuel cycle. Requirements for nuclear criticality safety impose the most restrictive of the various constraints on the operation of FCF. The upper limits on batch sizes and other important process parameters are determined principally by criticality safety considerations. To maintain an efficient operation within appropriate safety limits, it is necessary to formulate a nuclear criticality safety program that integrates equipment design, process development, process modeling, conduct of operations, a measurement program, adequate material control procedures, and nuclear criticality analysis. The nuclear criticality safety program for FCF reflects this integration, ensuring that the facility can be operated efficiently without compromising safety. The experience gained from the conduct of this program in the Fuel cycle Facility will be used to design and safely operate IFR facilities on a commercial scale. The key features of the nuclear criticality safety program are described. The relationship of these features to normal facility operation is also described

  9. Critical Facilities for Coastal Geographies

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The critical facilities data are derived from the USGS Structures Inventory Database (June, 2016). The structures in the derived dataset displays aggregated totals...

  10. Operating procedures for the Pajarito Site Critical Assembly Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1983-03-01

    Operating procedures consistent with DOE Order 5480.2, Chapter VI, and the American National Standard Safety Guide for the Performance of Critical Experiments are defined for the Pajarito Site Critical Assembly Facility of the Los Alamos National Laboratory. These operating procedures supersede and update those previously published in 1973 and apply to any criticality experiment performed at the facility

  11. Criticality experiment for No.2 core of DF-VI fast neutron criticality facility

    International Nuclear Information System (INIS)

    Yang Lijun; Liu Zhenhua; Yan Fengwen; Luo Zhiwen; Chu Chun; Liang Shuhong

    2007-01-01

    At the completion of the DF-VI fast neutron criticality facility, its core changed, and it was restarted and a series of experiments and measurements were made. According to the data from 29 criticality experiments, the criticality element number and mass were calculated, the control rod reactivity worth were measured by period method and rod compensate method, reactivity worth of safety rod and safety block were measured using reactivity instrument; the reactivity worth of outer elements and radial distribution of elements were measured too. Based on all the measurements mentioned above, safety operation parameters for core 2 in DF-VI fast neutron criticality facility were conformed. (authors)

  12. 76 FR 34770 - Rensselaer Polytechnic Institute Critical Experiments Facility; Environmental Assessment and...

    Science.gov (United States)

    2011-06-14

    ... the doses represent conservative estimates for the MHA. As discussed in the NRC staff's safety... Vicinity of the RCF--According to 2000 Census data, approximately 23,000 families and 123,000 individuals... family of four. According to Census data in the 2006-2008 American Community Survey 3-Year Estimates, the...

  13. CSER 94-012: Criticality safety evaluation report for 340 Facility

    International Nuclear Information System (INIS)

    Altschuler, S.J.

    1995-01-01

    This Criticality Safety Evaluation Report (CSER) covers the 340 Facility which acts as a collecting point for liquid and solid waste from various facilities in the 300 Area. Criticality safety is achieved by controlling the amount and concentration of the fissionable material sent to the 340 Facility from the originating facilities in the 300 Area, a method similar to that used elsewhere at Hanford for the waste tank farms. Unlike those, however, the waste received at the 340 Facility will be far less radioactive. It is concluded that present operations meet the two contingency criterion. The facility will still be safely subcritical even after two independent and concurrent failures (either of equipment or administrative controls). The solid waste storage and liquid waste will be managed separately. The solid waste storage area is classified as exempt because it contains less than 15 grams of fissionable materials. The Radioactive Liquid Waste System is classified as isolated because it contains less than one third of a minimum critical mass. The criticality safety of the 340 Facility devoted to the Radioactive Liquid Waste System (RLWS) is assured by the form and concentration of the fissile material and could also be classified as a limited control facility. However, the 340 Facility has been operated as an isolated facility which results in a more conservative limit

  14. Critical experiment study on uranyl nitrate solution experiment facility

    International Nuclear Information System (INIS)

    Zhu Qingfu; Shi Yongqian; Wang Jinrong

    2005-01-01

    The Uranyl Nitrate Solution Experiment Facility was constructed for the research on nuclear criticality safety. In this paper, the configuration of the facility is introduced; a series of critical experiments on uranyl nitrate solution is described later, which were performed for various uranium concentrations under different conditions, i.e. with or without neutron absorbers in the core and with or without water-reflector outside the core. Critical volume and the minimum 235U critical mass for different uranium concentrations are presented. Finally, theoretical analysis is made on the experimental results. (authors)

  15. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  16. Concepts for inventory verification in critical facilities

    International Nuclear Information System (INIS)

    Cobb, D.D.; Sapir, J.L.; Kern, E.A.; Dietz, R.J.

    1978-12-01

    Materials measurement and inventory verification concepts for safeguarding large critical facilities are presented. Inspection strategies and methods for applying international safeguards to such facilities are proposed. The conceptual approach to routine inventory verification includes frequent visits to the facility by one inspector, and the use of seals and nondestructive assay (NDA) measurements to verify the portion of the inventory maintained in vault storage. Periodic verification of the reactor inventory is accomplished by sampling and NDA measurement of in-core fuel elements combined with measurements of integral reactivity and related reactor parameters that are sensitive to the total fissile inventory. A combination of statistical sampling and NDA verification with measurements of reactor parameters is more effective than either technique used by itself. Special procedures for assessment and verification for abnormal safeguards conditions are also considered. When the inspection strategies and inventory verification methods are combined with strict containment and surveillance methods, they provide a high degree of assurance that any clandestine attempt to divert a significant quantity of fissile material from a critical facility inventory will be detected. Field testing of specific hardware systems and procedures to determine their sensitivity, reliability, and operational acceptability is recommended. 50 figures, 21 tables

  17. Safety analysis of the Los Alamos critical experiments facility

    International Nuclear Information System (INIS)

    Paxton, H.C.

    1975-10-01

    The safety of Pajarito Site critical assembly operations depends upon protection built into the facility, upon knowledgeable personnel, and upon good practice as defined by operating procedures and experimental plans. Distance, supplemented by shielding in some cases, would protect personnel against an extreme accident generating 10 19 fissions. During the facility's 28-year history, the direct cost of criticality accidents has translated to a risk of less than $200 per year

  18. Broadening of nuclear engineering programs: An engineering physics approach at Rensselaer

    International Nuclear Information System (INIS)

    Malaviya, B.K.

    1990-01-01

    With the maturing of nuclear engineering as an academic discipline and the uncertainty surrounding the nuclear industry, attention is being increasingly turned to ways in which the base of traditional nuclear engineering programs in universities can be broadened to make them more attractively useful to a wider class of potential students and employers while maintaining the strengths in mainstream areas of nuclear technology. An approach that seems to provide a natural evolution combining the existing programmatic strengths, infrastructure, and resources with the trending needs of a broad segment of diversified industries is the development and initiation of an engineering physics degree program as an adjunct to an established nuclear engineering curriculum. In line with these developments, a new comprehensive academic program offering baccalaureate, master's, and doctoral degrees in engineering physics has been developed and formally instituted at Rensselaer Polytechnic Institute (RPI). It provides a valuable opportunity for students to pursue education and research that cuts across traditional disciplinary lines, leading to a wide variety of career opportunities in industry, government, national research and defense laboratories, and academia

  19. Criticality safety training at the Hot Fuel Examination Facility

    International Nuclear Information System (INIS)

    Garcia, A.S.; Courtney, J.C.; Thelen, V.N.

    1983-01-01

    HFEF comprises four hot cells and out-of-cell support facilities for the US breeder program. The HFEF criticality safety program includes training in the basic theory of criticality and in specific criticality hazard control rules that apply to HFEF. A professional staff-member oversees the implementation of the criticality prevention program

  20. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  1. Experimental facilities of Valduc critical station

    International Nuclear Information System (INIS)

    Mangin, D.; Maubert, L.

    1975-01-01

    The critical facility of Valduc and its experimentation possibilities are described. The different experimental programs carried out since 1962 are briefly reviewed. The last program involving a plutonium nitrate solution (18.9wt% 240 Pu) in a large parallelepipedic tank is presented and main results given [fr

  2. International safeguards for critical facilities

    International Nuclear Information System (INIS)

    Ney, J.F.; Todd, J.L.

    1979-01-01

    A study was undertaken to investigate various approaches to provide international safeguards for critical facilities and to select an optimized system. Only high-inventory critical facilities were considered. The goal of the study was to detect and confirm the protracted or abrupt diversion of 8kg of plutonium or 25kg of the uranium isotope 235 within approximately a week of the diversion. The general safeguards alternatives considered were (1) continuous inspections by resident inspectors, with varying degrees of comprehensiveness, (2) periodic inspections by regional inspectors at varying time intervals, (3) unattended containment/surveillance measures, and (4) various combinations of the above. It was concluded that a practical and effective international safeguards system can be achieved by employing a method of continuously monitoring facility activities which could lead to diversion. This is in addition to the routine inspections typical of current international safeguards. Monitoring detects inventory discrepancies and violations of agreed-upon procedural restrictions, as well as unauthorized removal of Special Nuclear Materials (SNM). A special inventory is used following detection to confirm any suspected diversion. Comparison of 28 safeguards options led to the selection of a system for further development which uses a combination of surveillance and inspection by resident IAEA personnel, containment/surveillance by unattended equipment, and routine inventory sampling. A development programme is described which is intended to demonstrate the feasibility of several containment and surveillance measures proposed in the study. Included are a personnel portal and an instrument/material pass-through as well as associated recording and tamper-protection features. (author)

  3. Criticality safety research on nuclear fuel cycle facility

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2004-07-01

    This paper present d s current status and future program of the criticality safety research on nuclear fuel cycle made by Japan Atomic Energy Research Institute. Experimental research on solution fuel treated in reprocessing plant has been performed using two critical facilities, STACY and TRACY. Fundamental data of static and transient characteristics are accumulated for validation of criticality safety codes. Subcritical measurements are also made for developing a monitoring system for criticality safety. Criticality safety codes system for solution and power system, and evaluation method related to burnup credit are developed. (author)

  4. Design and construction of a fast critical facility

    International Nuclear Information System (INIS)

    Kato, W.Y.; Dates, L.R.

    1962-01-01

    Design and construction of a fast critical facility. In a fast-power-reactor development programme, a critical facility is found to be a highly useful tool to ascertain calculational techniques, to verify neutron cross-section sets, and to obtain integral reactor-physics parameters necessary for the nuclear design of a power system. Since it is primarily a physics instrument, the design of a fast critical facility itself poses a number of different problems not found in the design of a power reactor. In addition to usual questions of site, containment, core design and instrumentation , there arise such problems as: how to obtain a large degree of flexibility consistent with safety, the determination of the size and type of facility to meet the experimental physics requirements, the determination of the number and location of control and safety rods minimizing perturbation effects and the specification of the reproducibility of control rods and other movable components to obtain the accuracy required in reactivity measurements. These are some of the problems which are discussed in this paper based on recent experience at the Argonne National Laboratory which has under construction a fast critical facility, ZPR-VI at its Lemont, Illinois site for fast-reactor-physics studies. The ZPR-VI is a movable half- or split-table-type machine similar to ZPR-III. It has a matrix about two and a half times the volume of the earlier machine and will be used to investigate the physics of large, highly dilute, metal and cermet, unmoderated and partially moderated systems having core volumes up to about 1500 l. A detailed description of the ZPR-VI with a discussion on the criteria used in the design of its various components from the point of view of reactor physics is presented. In addition, such topics as management and operating procedures, potential hazards during operation, experimental techniques to be used and construction costs are also included. (author) [fr

  5. Technical specifications for the Oak Ridge Critical Experiments Facility

    International Nuclear Information System (INIS)

    Stinnett, R.M.

    1986-01-01

    These Technical Specifications for the Oak Ridge Critical Experiments Facility (CEF) delineate limiting conditions of operation for the facility. The CEF is used primarily for testing the High Flux Isotope Reactor (HFIR) fuel assemblies. Specifically, the Criticality Testing Unit, Liquid (CTUL), located in the CEF, is used for the HFIR fuel assembly test. The test is performed to satisfy the surveillance requirements of the HFIR Technical Specifications. The test is used to determine the water-submerged shutdown margin for each fuel assembly. 11 refs

  6. Los Alamos Critical Assemblies Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1981-06-01

    The Critical Assemblies Facility of the Los Alamos National Laboratory has been in existence for thirty-five years. In that period, many thousands of measurements have been made on assemblies of 235 U, 233 U, and 239 Pu in various configurations, including the nitrate, sulfate, fluoride, carbide, and oxide chemical compositions and the solid, liquid, and gaseous states. The present complex of eleven operating machines is described, and typical applications are presented

  7. Design Guide for Category I reactors critical facilities

    International Nuclear Information System (INIS)

    Brynda, W.J.; Powell, R.W.

    1978-08-01

    The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned critical facilities be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission

  8. Initial tank calibration at NUCEF critical facility. 2

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi

    1994-07-01

    Analyses on initial tank calibration data were carried out for the purpose of the nuclear material accountancy and control for critical facilities in NUCEF: Nuclear Fuel Cycle Safety Engineering Research Facility. Calibration functions to evaluate volume of nuclear material solution in accountancy tanks were determined by regression analysis on the data considering dimension and shape of the tank. The analyses on dip-tube separation (probe separation), which are necessary to evaluate solution density in the tanks, were also carried out. As a result, regression errors of volume calculated with the calibration functions were within 0.05 lit. (0.01%) at a nominal level of Pu accountancy tanks. Errors of the evaluated dip-tube separations were also small, e.g. within 0.2mm (0.11%). Therefore, it was estimated that systematic errors of bulk measurements would satisfy the target value of NUCEF critical facilities (0.3% for Pu accountancy tanks). This paper summarizes the data analysis methods, results of analysis and evaluated errors. (author)

  9. The Pajarito Site operating procedures for the Los Alamos Critical Experiments Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1991-12-01

    Operating procedures consistent with DOE Order 5480.6, and the American National Standard Safety Guide for the Performance of Critical Experiments are defined for the Los Alamos Critical Experiments Facility (LACEF) of the Los Alamos National Laboratory. These operating procedures supersede and update those previously published in 1983 and apply to any criticality experiment performed at the facility. 11 refs

  10. Criticality safety analysis for mockup facility

    International Nuclear Information System (INIS)

    Shin, Young Joon; Shin, Hee Sung; Kim, Ik Soo; Oh, Seung Chul; Ro, Seung Gy; Bae, Kang Mok

    2000-03-01

    Benchmark calculations for SCALE4.4 CSAS6 module have been performed for 31 UO 2 fuel, 15MOX fuel and 10 metal material criticality experiments and then calculation biases of the SCALE 4.4 CSAS6 module have been revealed to be 0.00982, 0.00579 and 0.02347, respectively. When CSAS6 is applied to the criticality safety analysis for the mockup facility in which several kinds of nuclear material components are included, the calculation bias of CSAS6 is conservatively taken to be 0.02347. With the aid of this benchmarked code system, criticality safety analyses for the mockup facility at normal and hypothetical accidental conditions have been carried out. It appears that the maximum K eff is 0.28356 well below than the critical limit, K eff =0.95 at normal condition. In a hypothetical accidental condition, the maximum K eff is found to be 0.73527 much lower than the subcritical limit. For another hypothetical accidental condition the nuclear material leaks out of container and spread or lump in the floor, it was assumed that the nuclear material is shaped into a slab and water exists in the empty space of the nuclear material. K eff has been calculated as function of slab thickness and the volume ratio of water to nuclear material. The result shows that the K eff increases as the water volume ratio increases. It is also revealed that the K eff reaches to the maximum value when water if filled in the empty space of nuclear material. The maximum K eff value is 0.93960 lower than the subcritical limit

  11. Seismic evaluation of critical facilities at the Lawrence Livermore Laboratory

    International Nuclear Information System (INIS)

    Murray, R.C.; Tokarz, F.J.

    1976-01-01

    The performance of critical facilities at the Lawrence Livermore Laboratory (LLL) are being evaluated for severe earthquake loading. Facilities at Livermore, Site-300 and the Nevada Test Site are included in this study. These facilities are identified, the seismic criteria used for the analysis are indicated, the various methods used for structural analysis are discussed and a summary of the results of facilities analyzed to date are presented

  12. Construction of the two-phase critical flow test facility

    International Nuclear Information System (INIS)

    Chung, C. H.; Chang, S. K.; Park, H. S.; Min, K. H.; Choi, N. H.; Kim, C. H.; Lee, S. H.; Kim, H. C.; Chang, M. H.

    2002-03-01

    The two-phase critical test loop facility has been constructed in the KAERI engineering laboratory for the simulation of small break loss of coolant accident entrained with non-condensible gas of SMART. The test facility can operate at 12 MPa of pressure and 0 to 60 C of sub-cooling with 0.5 kg/s of non- condensible gas injection into break flow, and simulate up to 20 mm of pipe break. Main components of the test facility were arranged such that the pressure vessel containing coolant, a test section simulating break and a suppression tank inter-connected with pipings were installed vertically. As quick opening valve opens, high pressure/temperature coolant flows through the test section forming critical two-phase flow into the suppression tank. The pressure vessel was connected to two high pressure N2 gas tanks through a control valve to control pressure in the pressure vessel. Another N2 gas tank was also connected to the test section for the non-condensible gas injection. The test facility operation was performed on computers supported with PLC systems installed in the control room, and test data such as temperature, break flow rate, pressure drop across test section, gas injection flow rate were all together gathered in the data acquisition system for further data analysis. This test facility was classified as a safety related high pressure gas facility in law. Thus the loop design documentation was reviewed, and inspected during construction of the test loop by the regulatory body. And the regulatory body issued permission for the operation of the test facility

  13. Nuclear Criticality Calculation for Determining the Bach Size in a Pyroprocessing Facility

    International Nuclear Information System (INIS)

    Ko, Won Il; Lee, Ho Hee; Chang, Hong Rae; Song, Dae Yong; Kwon, Eun Ha; Jung, Chang Jun; Yoon, Suk Kyun

    2009-01-01

    The criticality analysis in a pyroprocessing facility is very important element for the R and D and the facility design in terms of the determination of batch size of the sub-processes as well as facility safety. Particularly, the determining the batch size is essential at the beginning stage of the R and D. In this report, the criticality analysis was carried out for the subprocesses such as voloxidation, electrolytic reduction, electrorefining and electrowinning process in order to estimate the maximum batch size of each process by using Monte Carlo code (MCNP4/C2). On the whole, the criticality problem could not give a big effect on the batch sizes in the voloxidation, electrolytic reduction and electrorefining. However, it was resulted that permissible amount of nuclear material to prevent the criticality accident in the electrowinning process was about 10kgHM

  14. Nuclear Criticality Calculation for Determining the Bach Size in a Pyroprocessing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Lee, Ho Hee; Chang, Hong Rae; Song, Dae Yong; Kwon, Eun Ha; Jung, Chang Jun; Yoon, Suk Kyun [KAERI, Daejeon (Korea, Republic of)

    2009-01-15

    The criticality analysis in a pyroprocessing facility is very important element for the R and D and the facility design in terms of the determination of batch size of the sub-processes as well as facility safety. Particularly, the determining the batch size is essential at the beginning stage of the R and D. In this report, the criticality analysis was carried out for the subprocesses such as voloxidation, electrolytic reduction, electrorefining and electrowinning process in order to estimate the maximum batch size of each process by using Monte Carlo code (MCNP4/C2). On the whole, the criticality problem could not give a big effect on the batch sizes in the voloxidation, electrolytic reduction and electrorefining. However, it was resulted that permissible amount of nuclear material to prevent the criticality accident in the electrowinning process was about 10kgHM

  15. Adverse Condition and Critical Event Prediction in Cranfield Multiphase Flow Facility

    DEFF Research Database (Denmark)

    Egedorf, Søren; Shaker, Hamid Reza

    2017-01-01

    , or even to the environment. To cope with these, adverse condition and critical event prediction plays an important role. Adverse Condition and Critical Event Prediction Toolbox (ACCEPT) is a tool which has been recently developed by NASA to allow for a timely prediction of an adverse event, with low false...... alarm and missed detection rates. While ACCEPT has shown to be an effective tool in some applications, its performance has not yet been evaluated on practical well-known benchmark examples. In this paper, ACCEPT is used for adverse condition and critical event prediction in a multiphase flow facility....... Cranfield multiphase flow facility is known to be an interesting benchmark which has been used to evaluate different methods from statistical process monitoring. In order to allow for the data from the flow facility to be used in ACCEPT, methods such as Kernel Density Estimation (KDE), PCA-and CVA...

  16. Very high temperature gas-cooled reactor critical facility for Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    Ishihara, Noriyuki

    1985-01-01

    The outline of the critical facility, its construction, the results of the basic studies and experiments on the graphite material, and the results obtained from the test conducted on the overall functions of the critical facility were reported. With the completion of the critical facility, it has been made possible to demonstrate the establishment of the manufacturing techniques and product-quality guarantee for extremely pure isotropic graphite in addition to the reliability of the structural design and analytical techniques for the main unit of the critical facility. It is expected that the present facility will prove instrumental in the verification of the nuclear safety of the very high temperature gas-cooled nuclear reactor and in the acquisition of experimental data on the reactor physics pertaining to the improvement of the reactor characteristics. The tasks which remain to be accomplished hereafter are the improvements of the performance and quality features with regard to the oxidization of graphite, the heat-resisting structural materials, and the welded structures. (Kubozono, M.)

  17. SYNER-G buildings, lifelines, transportation networks and critical facilities

    CERN Document Server

    Pitilakis, K; Kaynia, AM

    2014-01-01

    This volume collates serviceable fragility functions from an international selection of seismic risk probability research, developing and adapting them to a European context to improve damage forecasts for buildings, infrastructure, and critical facilities.

  18. A Study of Critical Flowrate in the Integral Effect Test Facilities

    International Nuclear Information System (INIS)

    Kim, Yeongsik; Ryu, Sunguk; Cho, Seok; Yi, Sungjae; Park, Hyunsik

    2014-01-01

    In earlier studies, most of the information available in the literature was either for a saturated two-phase flow or a sub-cooled water flow at medium pressure conditions, e. g., up to about 7.0 MPa. The choking is regarded as a condition of maximum possible discharge through a given orifice and/or nozzle exit area. A critical flow rate can be achieved at a choking under the given thermo-hydraulic conditions. The critical flow phenomena were studied extensively in both single-phase and two-phase systems because of its importance in the LOCA analyses of light water reactors and in the design of other engineering areas. Park suggested a modified correlation for predicting the critical flow for sub-cooled water through a nozzle. Recently, Park et al. performed an experimental study on a two-phase critical flow with a noncondensable gas at high pressure conditions. Various experiments of critical flow using sub-cooled water were performed for a modeling of break simulators in thermohydraulic integral effect test facilities for light water reactors, e. g., an advanced power reactor 1400MWe (APR1400) and a system-integrated modular advanced reactor (SMART). For the design of break simulators of SBLOCA scenarios, the aspect ratio (L/D) is considered to be a key parameter to determine the shape of a break simulator. In this paper, an investigation of critical flow phenomena was performed especially on break simulators for LOCA scenarios in the integral effect test facilities of KAERI, such as ATLAS and FESTA. In this study, various studies on the critical flow models for sub-cooled and/or saturated water were reviewed. For a comparison among the models for the selected test data, discussions of the comparisons on the effect of the diameters, predictions of critical flow models, and break simulators for SBLOCA in the integral effect test facilities were presented

  19. A Study of Critical Flowrate in the Integral Effect Test Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeongsik; Ryu, Sunguk; Cho, Seok; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In earlier studies, most of the information available in the literature was either for a saturated two-phase flow or a sub-cooled water flow at medium pressure conditions, e. g., up to about 7.0 MPa. The choking is regarded as a condition of maximum possible discharge through a given orifice and/or nozzle exit area. A critical flow rate can be achieved at a choking under the given thermo-hydraulic conditions. The critical flow phenomena were studied extensively in both single-phase and two-phase systems because of its importance in the LOCA analyses of light water reactors and in the design of other engineering areas. Park suggested a modified correlation for predicting the critical flow for sub-cooled water through a nozzle. Recently, Park et al. performed an experimental study on a two-phase critical flow with a noncondensable gas at high pressure conditions. Various experiments of critical flow using sub-cooled water were performed for a modeling of break simulators in thermohydraulic integral effect test facilities for light water reactors, e. g., an advanced power reactor 1400MWe (APR1400) and a system-integrated modular advanced reactor (SMART). For the design of break simulators of SBLOCA scenarios, the aspect ratio (L/D) is considered to be a key parameter to determine the shape of a break simulator. In this paper, an investigation of critical flow phenomena was performed especially on break simulators for LOCA scenarios in the integral effect test facilities of KAERI, such as ATLAS and FESTA. In this study, various studies on the critical flow models for sub-cooled and/or saturated water were reviewed. For a comparison among the models for the selected test data, discussions of the comparisons on the effect of the diameters, predictions of critical flow models, and break simulators for SBLOCA in the integral effect test facilities were presented.

  20. Mapping the Characteristics of Critical Care Facilities: Assessment, Distribution, and Level of Critical Care Facilities from Central India.

    Science.gov (United States)

    Saigal, Saurabh; Sharma, Jai Prakash; Pakhare, Abhijit; Bhaskar, Santosh; Dhanuka, Sanjay; Kumar, Sanjay; Sabde, Yogesh; Bhattacharya, Pradip; Joshi, Rajnish

    2017-10-01

    In low- and middle-income countries such as India, where health systems are weak, the number of available Critical Care Unit (Intensive Care Unit [ICU]) beds is expected to be low. There is no study from the Indian subcontinent that has reported the characteristics and distribution of existing ICUs. We performed this study to understand the characteristics and distribution of ICUs in Madhya Pradesh (MP) state of Central India. We also aimed to develop a consensus scoring system and internally validate it to define levels of care and to improve health system planning and to strengthen referral networks in the state. We obtained a list of potential ICU facilities from various sources and then performed a cross-sectional survey by visiting each facility and determining characteristics for each facility. We collected variables with respect to infrastructure, human resources, equipment, support services, procedures performed, training courses conducted, and in-place policies or standard operating procedure documents. We identified a total of 123 ICUs in MP. Of 123 ICUs, 35 were level 1 facilities, 74 were level 2 facilities, and only 14 were level 3 facilities. Overall, there were 0.17 facilities per 100,000 population (95* confidence interval [CI] 0.14-0.20 per 100,000 populations). There were a total of 1816 ICU beds in the state, with an average of 2.5 beds per 100,000 population (95* CI 2.4-2.6 per 100,000 population). Of the total number of ICU beds, 250 are in level 1, 1141 are in level 2, and 425 are in level 3 facilities. This amounts to 0.34, 1.57, and 0.59 ICU beds per 100,000 population for levels 1, 2, and 3, respectively. This study could just be an eye opener for our healthcare authorities at both state and national levels to estimate the proportion of ICU beds per lac population. Similar mapping of intensive care services from other States will generate national data that is hitherto unknown.

  1. Status and future program of reactor physics experiments in JAERI Critical facilities, FCA and TCA

    International Nuclear Information System (INIS)

    Okajima, Shigeaki; Osugi, Toshitaka; Nakajima, Ken; Suzaki, Takenori; Miyoshi, Yoshinori

    1999-01-01

    The critical facilities in JAERI, FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly), have been used to provide integral data for evaluation of nuclear data as well as for development of various types of reactor since they went critical in 1960's. In this paper a review is presented on the experimental programs in both facilities. And the experimental programs in next 5 years are also shown. (author)

  2. Neutronic computational modeling of the ASTRA critical facility using MCNPX

    International Nuclear Information System (INIS)

    Rodriguez, L. P.; Garcia, C. R.; Milian, D.; Milian, E. E.; Brayner, C.

    2015-01-01

    The Pebble Bed Very High Temperature Reactor is considered as a prominent candidate among Generation IV nuclear energy systems. Nevertheless the Pebble Bed Very High Temperature Reactor faces an important challenge due to the insufficient validation of computer codes currently available for use in its design and safety analysis. In this paper a detailed IAEA computational benchmark announced by IAEA-TECDOC-1694 in the framework of the Coordinated Research Project 'Evaluation of High Temperature Gas Cooled Reactor (HTGR) Performance' was solved in support of the Generation IV computer codes validation effort using MCNPX ver. 2.6e computational code. In the IAEA-TECDOC-1694 were summarized a set of four calculational benchmark problems performed at the ASTRA critical facility. Benchmark problems include criticality experiments, control rod worth measurements and reactivity measurements. The ASTRA Critical Facility at the Kurchatov Institute in Moscow was used to simulate the neutronic behavior of nuclear pebble bed reactors. (Author)

  3. First start-up of nuclear criticality safety experiment facility for uranyl nitrate solution

    International Nuclear Information System (INIS)

    Zhu Qingfu; Shi Yongqian; Shen Leisheng; Hu Dingsheng; Zhao Shouzhi; He Tao; Sun Zheng; Lin Shenghuo; Yao Shigui

    2005-01-01

    The uranyl nitrate solution experiment facility for the research on nuclear criticality safety is described. The nuclear fuel loading steps in the first start-up for water-reflected core are presented. During the experiments, the critical volume of uranyl nitrate solution was determined as 20479.62 mL with count rate inverse extrapolation method, reactivity interpolation method, and steady power method. By calculation, critical mass of 235 U was derived as 1579.184 g from experimental data. The worth of control rods was also calibrated in the first start-up of the facility. (authors)

  4. Preliminary concepts for materials measurement and accounting in critical facilities

    International Nuclear Information System (INIS)

    Cobb, D.D.; Sapir, J.L.

    1978-01-01

    Preliminary concepts are presented for improved materials measurement and accounting in large critical facilities. These concepts will be developed as part of a study that will emphasize international safeguarding of critical facilities. The major safeguards problem is the timely verification of in-reactor inventory during periods of reactor operation. This will require a combination of measurement, statistical sampling, and data analysis techniques. Promising techniques include integral measurements of reactivity and other reactor parameters that are sensitive to the total fissile inventory, and nondestructive assay measurements of the fissile material in reactor fuel drawers and vault storage canisters coupled with statistical sampling plans tailored for the specific application. The effectiveness of proposed measurement and accounting strategies will be evaluated during the study

  5. Structural integrity monitoring of critical components in nuclear facilities

    International Nuclear Information System (INIS)

    Roth, Maria; Constantinescu, Dan Mihai; Brad, Sebastian; Ducu, Catalin; Malinovschi, Viorel

    2007-01-01

    Full text: The paper presents the results obtained as part of the Project 'Integrated Network for Structural Integrity Monitoring of Critical Components in Nuclear Facilities', RIMIS, a research work underway within the framework of the Ministry of Education and Research Programme 'Research of Excellence'. The main objective of the Project is to constitute a network integrating the national R and D institutes with preoccupations in the structural integrity assessment of critical components in the nuclear facilities operating in Romania, in order to elaborate a specific procedure for this field. The degradation mechanisms of the structural materials used in the CANDU type reactors, operated by Unit 1 and Unit 2 at Cernavoda (pressure tubes, fuel elements sheaths, steam generator tubing) and in the nuclear facilities relating to reactors of this type as, for instance, the Hydrogen Isotopes Separation facility, will be investigated. The development of a flexible procedure will offer the opportunity to extend the applications to other structural materials used in the nuclear field and in the non-nuclear fields as well, in cooperation with other institutes involved in the developed network. The expected results of the project will allow the integration of the network developed at national level in the structures of similar networks operating within the EU, the enhancement of the scientific importance of Romanian R and D organizations as well as the increase of our country's contribution in solving the major issues of the nuclear field. (authors)

  6. Validation the methodology calculate critical position of control rods to the critical facility IPEN/MB-01

    International Nuclear Information System (INIS)

    Lopez Aldama, D.; Rodriguez Gual, R.

    1998-01-01

    Presently work intends to validate the models and programs used in the Nuclear Technology Center for calculating the critical position of control rods by means of the analysis of the measurements performed at the critical facility IPEN/MB-01. The lattice calculations were carried out with the WIMS/D4 code and for the global calculations the diffusion code SNAP-3D was used

  7. Development of PSA procedure for a criticality in reprocessing facilities

    International Nuclear Information System (INIS)

    Endo, Shigeki; Takanashi, Mitsuhiro; Ueda, Yoshinori

    2012-08-01

    Utilization of risk information for the nuclear safety regulation is being discussed in Japan. The development of probabilistic safety assessment (PSA) procedure is indispensable for the utilization of risk information. The Japan Nuclear Energy Safety Organization (JNES) has been conducting trial PSA to a model plant for major events, i.e. hydrogen explosion, solution boiling, rapid decomposition of TBP complexes, criticality, solvent fire, leakage of molten glass, leakage of high active concentrated liquid waste, loss of all AC electricity, drop of a fuel assembly, for the purpose of developing the PSA procedure for reprocessing facilities. For criticality events results of trial PSA were summarized as a report in which how to evaluate an amount of radioactive materials released from a facility and a health effect on the public were emphasized. Therefore, for criticality events the results of trial PSA were summarized in this report to emphasize procedures from making event progression scenarios to quantifying event sequences, which were not handled in the previous report, in a style of a document describing PSA procedures. (author)

  8. Development of likelihood estimation method for criticality accidents of mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tamaki, Hitoshi; Yoshida, Kazuo; Kimoto, Tatsuya; Hamaguchi, Yoshikane

    2010-01-01

    A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility. (author)

  9. Results of the critical experiments concerning OTTO loading at the critical HTR-test facility KAHTER

    International Nuclear Information System (INIS)

    Drueke, V.; Litzow, W.; Paul, N.

    1982-12-01

    Critical experiments concerning OTTO loading are described. In the KAHTER facility an OTTO loading has been simulated, therefore the original KAHTER assembly was reconstructed. The determination of critical masses and reactivity worths of control rods and of additional absorber rods in the top reflector and in the upper cavity was of main interest for comparison with reactor following calculations. Besides this, reaction rates in different energy regions were measured in the upper part of the core, in the cavity and top reflector. (orig.) [de

  10. Utilization of the BARC critical facility for ADS related experiments

    Indian Academy of Sciences (India)

    The paper discusses the basic design of the critical facility, whose main pur- ... systems. In addition, it will have a flux mapping system based on 25 fission ... neutron source leads to peaked flux distribution exciting other higher harmonic.

  11. Precautions for preventing criticality at plutonium fuel treatment facilities

    International Nuclear Information System (INIS)

    Deworm, J.P.; Fieuw, G.; Cank, H. de

    1976-01-01

    Four criticality accidents took place between 1958 and 1964 at fuel processing plants using wet methods. So far accident of this type has taken place at production units where fissionable material is used. The prevention of criticality is one of the major concerns of the officials in charge of the plutonium fuel research laboratories operated at the Mol Nuclear Energy Study Centre by the SCK/CEN-Belgonucleaire Association. The means of preventing such an accident are of three types: introducing different types of treatment in well-defined work units; thorough analysis of planned experiments or fabrication programmes to determine the sub-criticality factors; application of technical and administrative procedures which ensure that the facilities are always sub-critical during the treatment and storage of fissionable materials. The installation includes a detection and warning system and provision is made for the immediate evacuation of staff should a crticality incident occur. The effects of a critical excursion on the building have been assessed. (author)

  12. Developing guidance in the nuclear criticality safety assessment for fuel cycle facilities

    International Nuclear Information System (INIS)

    Galet, C.; Evo, S.

    2012-01-01

    In this poster IRSN (Institute for radiation protection and nuclear safety) presents its safety guides whose purpose is to transmit the safety assessment know-how to any 'junior' staff or even to give a view of the safety approach on the overall risks to any staff member. IRSN has written a first version of such a safety guide for fuel cycle facilities and laboratories. It is organized into several chapters: some refer to types of assessments, others concern the types of risks. Currently, this guide contains 13 chapters and each chapter consists of three parts. In parallel to the development of criticality chapter of this guide, the IRSN criticality department has developed a nuclear criticality safety guide. It follows the structure of the three parts fore-mentioned, but it presents a more detailed first part and integrates, in the third part, the experience feedback collected on nuclear facilities. The nuclear criticality safety guide is online on the IRSN's web site

  13. Criticality Safety Lessons Learned in a Deactivation and Decommissioning Environment [A Guide for Facility and Project Managers

    Energy Technology Data Exchange (ETDEWEB)

    Nirider, L. Tom

    2003-08-06

    This document was designed as a reference and a primer for facility and project managers responsible for Deactivation and Decommissioning (D&D) processes in facilities containing significant inventories of fissionable materials. The document contains lessons learned and guidance for the development and management of criticality safety programs. It also contains information gleaned from occurrence reports, assessment reports, facility operations and management, NDA program reviews, criticality safety experts, and criticality safety evaluations. This information is designed to assist in the planning process and operational activities. Sufficient details are provided to allow the reader to understand the events, the lessons learned, and how to apply the information to present or planned D&D processes. Information is also provided on general lessons learned including criticality safety evaluations and criticality safety program requirements during D&D activities. The document also explores recent and past criticality accidents in operating facilities, and it extracts lessons learned pertinent to D&D activities. A reference section is included to provide additional information. This document does not address D&D lessons learned that are not pertinent to criticality safety.

  14. Low-power critical facilities: their role in the nuclear renaissance

    International Nuclear Information System (INIS)

    Didsbury, R.

    2011-01-01

    This paper discusses the role of low power critical facilities and their role in the nuclear renaissance. It outline the role of human capital in some detail. sufficient conditions for the renaissance are that nuclear power is safe, sustainable, economical and proliferation resistant.

  15. Technical specifications for the Pajarito Site Critical Experiments Facility

    International Nuclear Information System (INIS)

    Malenfant, R.E.; Paxton, H.C.

    1980-12-01

    This document is to satisfy the requirement for technical specifications spelled out in DOE Manual Chapter 0540, Safety of DOE-Owned Reactors. Technical specifications are defined in Sec. 0540-048, and the requirement for them appears in Sec. 0540-015. The following technical specifications update the document, Technical Specifications for the Pajarito Site Critical Experiments Facility

  16. The Potential for Criticality Following Disposal of Uranium at Low-Level-Waste Facilities. Containerized Disposal

    International Nuclear Information System (INIS)

    Colten-Bradley, V.A.; Hopper, C.M.; Parks, C.V.; Toran, L.E.

    1999-01-01

    The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop and test some reasonable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM) and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team's approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some possible scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increase in uranium concentration over disposal limits. The analysis of SNM was restricted to 235 U in the present scope of work. The work documented in this report indicates that the potential for a criticality safety concern to arise in an LLW facility is extremely remote, but not impossible. Theoretically, conditions that lead to a potential criticality safety concern might arise. However, study of the hydrogeochemical mechanisms, the associated time frames, and the factors required for an actual criticality event indicate that proper emplacement of the SNM at the site can eliminate practical concerns relative to the occurrence and possible consequences of a criticality event

  17. Criticality Safety Lessons Learned in a Deactivation and Decommissioning Environment [A Guide for Facility and Project Managers

    International Nuclear Information System (INIS)

    NIRIDER, L.T.

    2003-01-01

    This document was designed as a reference and a primer for facility and project managers responsible for Deactivation and Decommissioning (D and D) processes in facilities containing significant inventories of fissionable materials. The document contains lessons learned and guidance for the development and management of criticality safety programs. It also contains information gleaned from occurrence reports, assessment reports, facility operations and management, NDA program reviews, criticality safety experts, and criticality safety evaluations. This information is designed to assist in the planning process and operational activities. Sufficient details are provided to allow the reader to understand the events, the lessons learned, and how to apply the information to present or planned D and D processes. Information is also provided on general lessons learned including criticality safety evaluations and criticality safety program requirements during D and D activities. The document also explores recent and past criticality accidents in operating facilities, and it extracts lessons learned pertinent to D and D activities. A reference section is included to provide additional information. This document does not address D and D lessons learned that are not pertinent to criticality safety

  18. A new facility for the determination of critical heat flux in nuclear fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Fortman, R A; Hadaller, G I; Hamilton, R C; Hayes, R C; Shin, K S; Stern, F [Stern Laboratories Inc., Hamilton, ON (Canada)

    1993-11-01

    A facility for the determination of critical heat flux in simulated reactor fuel assemblies has been constructed at Stern Laboratories for CANDU Owners` Group. This paper describes the facility and method of testing. 9 figs.

  19. Criticality safety evaluation report for the cold vacuum drying facility's process water handling system

    International Nuclear Information System (INIS)

    NELSON, J.V.

    1999-01-01

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified

  20. Optically-based Sensor System for Critical Nuclear Facilities Post-Event Seismic Structural Assessment

    Energy Technology Data Exchange (ETDEWEB)

    McCallen, David [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Petrone, Floriana [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Buckle, Ian [Univ. of Nevada, Reno, NV (United States); Wu, Suiwen [Univ. of Nevada, Reno, NV (United States); Coates, Jason [California State Univ., Chico, CA (United States)

    2017-09-30

    The U.S. Department of Energy (DOE) has ownership and operational responsibility for a large enterprise of nuclear facilities that provide essential functions to DOE missions ranging from national security to discovery science and energy research. These facilities support a number of DOE programs and offices including the National Nuclear Security Administration, Office of Science, and Office of Environmental Management. With many unique and “one of a kind” functions, these facilities represent a tremendous national investment, and assuring their safety and integrity is fundamental to the success of a breadth of DOE programs. Many DOE critical facilities are located in regions with significant natural phenomenon hazards including major earthquakes and DOE has been a leader in developing standards for the seismic analysis of nuclear facilities. Attaining and sustaining excellence in nuclear facility design and management must be a core competency of the DOE. An important part of nuclear facility management is the ability to monitor facilities and rapidly assess the response and integrity of the facilities after any major upset event. Experience in the western U.S. has shown that understanding facility integrity after a major earthquake is a significant challenge which, lacking key data, can require extensive effort and significant time. In the work described in the attached report, a transformational approach to earthquake monitoring of facilities is described and demonstrated. An entirely new type of optically-based sensor that can directly and accurately measure the earthquake-induced deformations of a critical facility has been developed and tested. This report summarizes large-scale shake table testing of the sensor concept on a representative steel frame building structure, and provides quantitative data on the accuracy of the sensor measurements.

  1. Development of inherent technologies for advanced PWR core - A study on the current status and the construction feasibility of critical facilities

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Won Sik; Yang, Hyun Seok [Chosun University, Kwangju (Korea); Kim, Chang Hyo; Shim, Hyung Jin [Seoul National University, Seoul (Korea)

    1999-03-01

    The objective of this study is to examine the appropriateness of constructing critical facilities in our country and to decide a course of constructing them if necessary by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. We investigated the status and the utilization of foreign critical facilities through literature survey and personal visitation. In our judgement, critical facilities are necessary for developing the advanced reactors and fuels which are being studied as parts of the Nuclear R and D Program by MOST. Considering the construction cost and the current state of domestic economy, however, it is unjustifiable to build three different types of critical facilities (the light water, the heavy water, and the fast critical facility). It appears to be reasonable to build a light water critical, considering the construction cost, degree of utilization, and other constraints. (author). 89 refs., 134 figs., 64 tabs.

  2. Data base of reactor physics experimental results in Kyoto University critical assembly experimental facilities

    International Nuclear Information System (INIS)

    Ichihara, Chihiro; Fujine, Shigenori; Hayashi, Masatoshi

    1986-01-01

    The Kyoto University critical assembly experimental facilities belong to the Kyoto University Research Reactor Institute, and are the versatile critical assembly constructed for experimentally studying reactor physics and reactor engineering. The facilities are those for common utilization by universities in whole Japan. During more than ten years since the initial criticality in 1974, various experiments on reactor physics and reactor engineering have been carried out using many experimental facilities such as two solidmoderated cores, a light water-moderated core and a neutron generator. The kinds of the experiment carried out were diverse, and to find out the required data from them is very troublesome, accordingly it has become necessary to make a data base which can be processed by a computer with the data accumulated during the past more than ten years. The outline of the data base, the data base CAEX using personal computers, the data base supported by a large computer and so on are reported. (Kako, I.)

  3. A cost effective approach for criticality accident analysis of a DOE SNF storage facility

    International Nuclear Information System (INIS)

    Garrett, R.L.; Couture, G.F.; Gough, S.T.

    1997-01-01

    This paper presents the methodologies used to derive criticality accident analyses for a spent nuclear fuel receipt, storage, handling, and shipping facility. Two criticality events are considered: process-induced and Natural Phenomena Hazards (NPH)-induced. The criticality analyses required the development of: (1) the frequency at which each sceanario occurred, (2) the estimated number of fissions for each scenario, and (3) the consequences associated with each criticality scenario. A fault tree analysis was performed to quantify the frequency of criticality due to process-induced events. For the frequency analysis of NPH-induced criticality, a probabilistic approach was employed. To estimate the consequences of a criticality event, the resulting fission yield was determined using a probabilistic approach. For estimating the source term, a 95% amount of overall conservatism was targeted. This methodology applied to the facility criticality scenarios indicated that: (1) the 95th percentile yield levels from the historical yield distributions are approximately 5 x 10 17 fissions and 5 x 10 18 fissions for internal event and NPH-induced criticality event, respectively; and (2) using probabilistic Latin Hypercube Sampling, the downwind 95th percentile dose to a receptor at the US DOE reservation boundary is 2.2 mrem. This estimate is compared to the bounding dose of 1.4 rem. 4 refs., 1 fig

  4. Review of design criteria for Criticality Accident Alarm System (CAAS) used in Fuel Reprocessing Facility

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Basu, Pew; Sivasubramaniyan, K.; Venkatraman, B.

    2016-01-01

    Though fuel cycle facilities handling fissile materials are designed with careful criticality safety analysis, the criticality accident cannot be ruled out completely. Criticality Accident Alarm System (CAAS) is being installed as part of criticality safety management in fuel cycle facilities. CAAS system being used in India, is ECIL make, ionization chamber based gamma detector, which houses three identical detectors and works on 2/3 logic. As per ISO 7753 and ANSI/ANS-8.3, the CAAS must be designed to be capable of detecting any minimum accident occurs which could be of concern. Based on this, alarm limit used in CAAS is: 4 R/h (fast transient excursion) and 3 mR in 0.5 sec (slow excursion). In case of reprocessing facilities wherein process tanks located in heavy shielding, identification of CAAS installation locations require detailed radiation transport calculations. A study has been taken to estimate the gamma dose rate from thick concrete hot cells in order to determine the locations of CAAS to meet the present design criteria of alarm limit

  5. SRTC criticality safety technical review: Nuclear criticality safety evaluation 94-02, uranium solidification facility pencil tank module spacing

    International Nuclear Information System (INIS)

    Rathbun, R.

    1994-01-01

    Review of NMP-NCS-94-0087, ''Nuclear Criticality Safety Evaluation 94-02: Uranium Solidification Facility Pencil Tank Module Spacing (U), April 18, 1994,'' was requested of the SRTC Applied Physics Group. The NCSE is a criticality assessment to show that the USF process module spacing, as given in Non-Conformance Report SHM-0045, remains safe for operation. The NCSE under review concludes that the module spacing as given in Non-Conformance Report SHM-0045 remains in a critically safe configuration for all normal and single credible abnormal conditions. After a thorough review of the NCSE, this reviewer agrees with that conclusion

  6. Conceptual design of the test facility for the two-phase critical flow with non-condensable gas

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Chung, Chang Hwan

    2000-12-01

    The two-phase critical flow test with non-condensible gas is for the simulation of the critical flow phenomena which can be occurred during SB-LOCA on SMART reactor. The requirements of the critical flow test are 7{approx}20mm pipe break dia., 7{approx}12MPa stagnation pressure, 0{approx}60 deg C subcooling degree and 0{approx}0.5kg/s N2 gas flow rate. For the satisfaction of these requirements on the test facility, critical flow rates were calculated with various models. With the selected reference pressure vessel(1.3m{sup 3}), the conceptual design of the test facility was performed. The important components of the test facility are the pressure vessel which has main circulation line, the test section attached to the bottom of the pressure vessel, suppression tank, the N2 gas supply tanks for maintaining the system pressure and N2 gas flow rate at test section and the auxiliary N2 gas converting system. For the measurements of the critical flow rate, flowmeter and level gauge is installed at the upstream of the test section and the pressure vessel, respectively. The realtime pressure control system is installed at the entrance of the pressure vessel for maintaining the system pressure and the N2 gas flow regulating system is also installed at the upstream of the test section. The design of the control and monitoring system for the operation of the test facility and the DAS for acquiring the test data were also performed. The conceptual operating process of the test facility was determined.

  7. Conceptual design of the test facility for the two-phase critical flow with non-condensable gas

    International Nuclear Information System (INIS)

    Chang, Seok Kyu; Chung, Chang Hwan

    2000-12-01

    The two-phase critical flow test with non-condensible gas is for the simulation of the critical flow phenomena which can be occurred during SB-LOCA on SMART reactor. The requirements of the critical flow test are 7∼20mm pipe break dia., 7∼12MPa stagnation pressure, 0∼60 deg C subcooling degree and 0∼0.5kg/s N2 gas flow rate. For the satisfaction of these requirements on the test facility, critical flow rates were calculated with various models. With the selected reference pressure vessel(1.3m 3 ), the conceptual design of the test facility was performed. The important components of the test facility are the pressure vessel which has main circulation line, the test section attached to the bottom of the pressure vessel, suppression tank, the N2 gas supply tanks for maintaining the system pressure and N2 gas flow rate at test section and the auxiliary N2 gas converting system. For the measurements of the critical flow rate, flowmeter and level gauge is installed at the upstream of the test section and the pressure vessel, respectively. The realtime pressure control system is installed at the entrance of the pressure vessel for maintaining the system pressure and the N2 gas flow regulating system is also installed at the upstream of the test section. The design of the control and monitoring system for the operation of the test facility and the DAS for acquiring the test data were also performed. The conceptual operating process of the test facility was determined

  8. Nuclear criticality safety analysis summary report: The S-area defense waste processing facility

    International Nuclear Information System (INIS)

    Ha, B.C.

    1994-01-01

    The S-Area Defense Waste Processing Facility (DWPF) can process all of the high level radioactive wastes currently stored at the Savannah River Site with negligible risk of nuclear criticality. The characteristics which make the DWPF critically safe are: (1) abundance of neutron absorbers in the waste feeds; (2) and low concentration of fissionable material. This report documents the criticality safety arguments for the S-Area DWPF process as required by DOE orders to characterize and to justify the low potential for criticality. It documents that the nature of the waste feeds and the nature of the DWPF process chemistry preclude criticality

  9. Criticality Safety Evaluation Report for the Cold Vacuum Drying (CVD) Facility's Process Water Handling System

    International Nuclear Information System (INIS)

    KESSLER, S.F.

    2000-01-01

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility. The controls and limitations on equipment design and operations to control potential criticality occurrences are identified

  10. Operation databook of the fuel treatment system of the Static Experiment Critical Facility (STACY) and the Transient Experiment Critical Facility (TRACY). JFY 2004 to JFY 2008

    International Nuclear Information System (INIS)

    Kokusen, Junya; Sumiya, Masato; Seki, Masakazu; Kobayashi, Fuyumi; Ishii, Junichi; Umeda, Miki

    2013-02-01

    Uranyl nitrate solution fuel used in the Static Experiment Critical Facility (STACY) and the Transient Experiment Critical Facility (TRACY) is adjusted in the Fuel Treatment System, in which such parameters are varied as concentration of uranium, free nitric acid, soluble neutron poison, and so on. Operations for concentration and denitration of the solution fuel were carried out with an evaporator from JFY 2004 to JFY 2008 in order to adjust the fuel to the experimental condition of the STACY and the TRACY. In parallel, the solution fuel in which some kinds of soluble neutron poison were doped was also adjusted in JFY 2005 and JFY 2006 for the purpose of the STACY experiments to determine neutron absorption effects brought by fission products, etc. After these experiments in the STACY, a part of the solution fuel including the soluble neutron poison was purified by the solvent extraction method with mixer-settlers in JFY 2006 and JFY 2007. This report summarizes operation data of the Fuel Treatment System from JFY 2004 to JFY 2008. (author)

  11. Operating manual for the critical experiments facility

    International Nuclear Information System (INIS)

    1986-01-01

    The operation of the Critical Experiments Facility (CEF) requires careful attention to procedures in order that all safety precautions are observed. Since an accident could release large amounts of radioactivity, careful operation and strict enforcement of procedures are necessary. To provide for safe operation, detailed procedures have been written for all phases of the operation of this facility. The CEF operating procedures are not to be construed to constitute a part ofthe Technical Specifications. In the event of any discrepancy between the information given herein and the Technical Specifications, limits set forth in the Technical Specifications apply. All normal and most emergency operation conditions are covered by procedures presented in this manual. These procedures are designed to be followed by the operating personnel. Strict adherence to these procedures is expected for the following reasons. (1) To provide a standard, safe method of performing all operations, the procedures were written by reactor engineers experienced in supervising the operation of reactors and were reviewed by an organization with over 30 years of reactor operating experience. (2) To have an up-to-date description of operating techniques available at all times for reference and review, it is necessary that the procedures be written

  12. Operating manual for the critical experiments facility

    Energy Technology Data Exchange (ETDEWEB)

    1986-01-01

    The operation of the Critical Experiments Facility (CEF) requires careful attention to procedures in order that all safety precautions are observed. Since an accident could release large amounts of radioactivity, careful operation and strict enforcement of procedures are necessary. To provide for safe operation, detailed procedures have been written for all phases of the operation of this facility. The CEF operating procedures are not to be construed to constitute a part ofthe Technical Specifications. In the event of any discrepancy between the information given herein and the Technical Specifications, limits set forth in the Technical Specifications apply. All normal and most emergency operation conditions are covered by procedures presented in this manual. These procedures are designed to be followed by the operating personnel. Strict adherence to these procedures is expected for the following reasons. (1) To provide a standard, safe method of performing all operations, the procedures were written by reactor engineers experienced in supervising the operation of reactors and were reviewed by an organization with over 30 years of reactor operating experience. (2) To have an up-to-date description of operating techniques available at all times for reference and review, it is necessary that the procedures be written.

  13. Critical/non-critical system methodology report

    International Nuclear Information System (INIS)

    1989-01-01

    The method used to determine how the waste Isolation Pilot Plant (WIPP) facilities/systems were classified as critical or non-critical to the receipt of CH waste is described within this report. All WIPP critical facilities/systems are listed in the Operational Readiness Review Dictionary. Using the Final Safety Analysis Report (FSAR) as a guide to define the boundaries of the facilities/systems, a direct correlation of the ORR Dictionary to the FSAR can be obtained. The critical facilities/systems are those which are directly related to or have a critical support role in the receipt of CH waste. The facility/systems must meet one of the following requirements to be considered critical: (a) confinement or measure of the release of radioactive materials; (b) continued receipt and/or storage of transuranic waste (TRU) without an interruption greater than one month according to the shipping plan schedule; (c) the environmental and occupational safety of personnel meets the established site programs; and (d) the physical security of the WIPP facilities

  14. Nuclear criticality safety assessment of the Consolidated Edison Uranium-Solidification Program Facility

    International Nuclear Information System (INIS)

    Thomas, J.T.

    1984-01-01

    A nuclear criticality assessment of the Consolidated Edison Uranium-Solidification Program facility confirms that all operations involved in the process may be conducted with an acceptable margin of subcriticality. Normal operation presents no concern since subcriticality is maintained by design. Several recommendations are presented to prevent, or mitigate the consequences of, any abnormal events that might occur in the various portions of the process. These measures would also serve to reduce to a minimum the administrative controls required to prevent criticality

  15. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    International Nuclear Information System (INIS)

    Brogli, R.; Bucher, K.H.; Chawla, R.; Foskolos, K.; Luchsinger, H.; Mathews, D.; Sarlos, G.; Seiler, R.

    1990-01-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  16. LEU-HTR critical experiment program for the PROTEUS facility in Switzerland

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R; Bucher, K H; Chawla, R; Foskolos, K; Luchsinger, H; Mathews, D; Sarlos, G; Seiler, R [Paul Scherrer Institute, Laboratory for Reactor Physics and System Technology Wuerenlingen and Villigen, Villigen PSI (Switzerland)

    1990-07-01

    New critical experiments in the framework of an IAEA Coordinated Research Program on 'Validation of Safety Related Reactor Physics Calculations for Low Enriched HTRs' are planned at the PSI PROTEUS facility. The experiments are designed to supplement the experimental data base and reduce the design and licensing uncertainties for small- and medium-sized helium-cooled reactors using low-enriched uranium (LEU) and graphite high temperature fuel. The main objectives of the new experiments are to provide first-of-a-kind high quality experimental data on: 1) The criticality of simple, easy to interpret, single core region LEU HTR systems for several moderator-to-fuel ratios and several lattice geometries; 2) the changes in reactivity, neutron balance components and control rod effectiveness caused by water ingress into this type of reactor, and 3) the effects of the boron and/or hafnium absorbers that are used to modify the reactivity and the power distributions in typical HTR systems. Work on the design and licensing of the modified PROTEUS critical facility is now in progress with the HTR experiments scheduled to begin early in 1991. Several international partners will be involved in the planning, execution and analysis of these experiments in order to insure that they are relevant and cost effective with respect to the various gas cooled reactor national programs. (author)

  17. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1993-11-01

    This document contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE non-reactor nuclear facilities. Adherence to these guidelines will provide consistency and uniformity in criticality safety evaluations (CSEs) across the complex and will document compliance with the requirements of DOE Order 5480.24

  18. Analysis of RA-8 critical facility core in some configurations

    International Nuclear Information System (INIS)

    Abbate, Maximo J.; Sbaffoni, Maria M.

    2000-01-01

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  19. Monitoring critical facilities by using advanced RF devices

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, Hanchung; Liu, Yung Y. [Argonne National Laboratory, Argonne, IL (United States); Shuler, James [U.S. Department of Energy, Washington, D.C. (United States)

    2013-07-01

    developed by Argonne National Laboratory. ARG-US RAMM, powered by on-board battery, can sustain extended autonomous surveillance operation during and following an incident. The benefits could be invaluable to such critical facilities as nuclear power plants, research and test reactors, fuel cycle manufacturing centers, spent-fuel dry-cask storage facilities, and other nuclear installations. (authors)

  20. Monitoring critical facilities by using advanced RF devices

    International Nuclear Information System (INIS)

    Tsai, Hanchung; Liu, Yung Y.; Shuler, James

    2013-01-01

    developed by Argonne National Laboratory. ARG-US RAMM, powered by on-board battery, can sustain extended autonomous surveillance operation during and following an incident. The benefits could be invaluable to such critical facilities as nuclear power plants, research and test reactors, fuel cycle manufacturing centers, spent-fuel dry-cask storage facilities, and other nuclear installations. (authors)

  1. YALINA facility a sub-critical Accelerator- Driven System (ADS) for nuclear energy research facility description and an overview of the research program (1997-2008).

    Energy Technology Data Exchange (ETDEWEB)

    Gohar, Y.; Smith, D. L.; Nuclear Engineering Division

    2010-04-28

    The YALINA facility is a zero-power, sub-critical assembly driven by a conventional neutron generator. It was conceived, constructed, and put into operation at the Radiation Physics and Chemistry Problems Institute of the National Academy of Sciences of Belarus located in Minsk-Sosny, Belarus. This facility was conceived for the purpose of investigating the static and dynamic neutronics properties of accelerator driven sub-critical systems, and to serve as a neutron source for investigating the properties of nuclear reactions, in particular transmutation reactions involving minor-actinide nuclei. This report provides a detailed description of this facility and documents the progress of research carried out there during a period of approximately a decade since the facility was conceived and built until the end of 2008. During its history of development and operation to date (1997-2008), the YALINA facility has hosted several foreign groups that worked with the resident staff as collaborators. The participation of Argonne National Laboratory in the YALINA research programs commenced in 2005. For obvious reasons, special emphasis is placed in this report on the work at YALINA facility that has involved Argonne's participation. Attention is given here to the experimental program at YALINA facility as well as to analytical investigations aimed at validating codes and computational procedures and at providing a better understanding of the physics and operational behavior of the YALINA facility in particular, and ADS systems in general, during the period 1997-2008.

  2. Critical Protection Item classification for a waste processing facility at Savannah River Site

    International Nuclear Information System (INIS)

    Ades, M.J.; Garrett, R.J.

    1993-01-01

    This paper describes the methodology for Critical Protection Item (CPI) classification and its application to the Structures, Systems and Components (SSC) of a waste processing facility at the Savannah River Site (SRS). The WSRC methodology for CPI classification includes the evaluation of the radiological and non-radiological consequences resulting from postulated accidents at the waste processing facility and comparison of these consequences with allowable limits. The types of accidents considered include explosions and fire in the facility and postulated accidents due to natural phenomena, including earthquakes, tornadoes, and high velocity straight winds. The radiological analysis results indicate that CPIs are not required at the waste processing facility to mitigate the consequences of radiological release. The non-radiological analysis, however, shows that the Waste Storage Tank (WST) and the dike spill containment structures around the formic acid tanks in the cold chemical feed area and waste treatment area of the facility should be identified as CPIs. Accident mitigation options are provided and discussed

  3. Onsite and Electric Backup Capabilities at Critical Infrastructure Facilities in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Julia A. [Argonne National Lab. (ANL), Argonne, IL (United States); Wallace, Kelly E. [Argonne National Lab. (ANL), Argonne, IL (United States); Kudo, Terence Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Eto, Joseph H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2016-04-01

    The following analysis, conducted by Argonne National Laboratory’s (Argonne’s) Risk and Infrastructure Science Center (RISC), details an analysis of electric power backup of national critical infrastructure as captured through the Department of Homeland Security’s (DHS’s) Enhanced Critical Infrastructure Program (ECIP) Initiative. Between January 1, 2011, and September 2014, 3,174 ECIP facility surveys have been conducted. This study focused first on backup capabilities by infrastructure type and then expanded to infrastructure type by census region.

  4. Fire criticality probability analysis for 300 Area N Reactor fuel fabrication and storage facility. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.E.

    1995-02-08

    Uranium fuel assemblies and other uranium associated with the shutdown N Reactor are stored in the 300 Area N Reactor Fuel Fabrication and Storage Facility (Facility). The 3712 Building, where the majority of the fuel assemblies and other uranium is stored, is where there could be a potential for a criticality bounding case. The purpose of this study is to evaluate the probability of potential fires in the Facility 3712 Building that could lead to criticality. This study has been done to support the criticality update. For criticality to occur, the wooden fuel assembly containers would have to burn such that the fuel inside would slump into a critical geometry configuration, a sufficient number of containers burn to form an infinite wide configuration, and sufficient water (about a 17 inch depth) be placed onto the slump. To obtain the appropriate geometric configuration, enough fuel assembly containers to form an infinite array on the floor would have to be stacked at least three high. Administrative controls require the stacks to be limited to two high for 1.25 wt% enriched finished fuel. This is not sufficient to allow for a critical mass regardless of the fire and accompanying water moderation. It should be noted that 0.95 wt% enriched fuel and billets are stacked higher than only two high. In this analysis, two initiating events will be considered. The first is a random fire that is hot enough and sufficiently long enough to burn away the containers and fuel separators such that the fuel can establish a critical mass. The second is a seismically induced fire with the same results.

  5. Survey to identify depth of penetration of critical incident reporting systems in Austrian healthcare facilities.

    Science.gov (United States)

    Sendlhofer, Gerald; Eder, Harald; Leitgeb, Karina; Gorges, Roland; Jakse, Heidelinde; Raiger, Marianne; Türk, Silvia; Petschnig, Walter; Pregartner, Gudrun; Kamolz, Lars-Peter; Brunner, Gernot

    2018-01-01

    Incident reporting systems or so-called critical incident reporting systems (CIRS) were first recommended for use in health care more than 15 years ago. The uses of these CIRS are highly variable among countries, ranging from being used to report critical incidents, falls, or sentinel events resulting in death. In Austria, CIRS have only been introduced to the health care sector relatively recently. The goal of this work, therefore, was to determine whether and specifically how CIRS are used in Austria. A working group from the Austrian Society for Quality and Safety in Healthcare (ASQS) developed a survey on the topic of CIRS to collect information on penetration of CIRS in general and on how CIRS reports are used to increase patient safety. Three hundred seventy-one health care professionals from 274 health care facilities were contacted via e-mail. Seventy-eight respondents (21.0%) completed the online survey, thereof 66 from hospitals and 12 from other facilities (outpatient clinics, nursing homes). In all, 64.1% of the respondents indicated that CIRS were used in the entire health care facility; 20.6% had not yet introduced CIRS and 15.4% used CIRS only in particular areas. Most often, critical incidents without any harm to patients were reported (76.9%); however, some health care facilities also use their CIRS to report patient falls (16.7%), needle stick injuries (17.9%), technical problems (51.3%), or critical incidents involving health care professionals. CIRS are not yet extensively or homogeneously used in Austria. Inconsistencies exist with respect to which events are reported as well as how they are followed up and reported to health care professionals. Further recommendations for general use are needed to support the dissemination in Austrian health care environments.

  6. Installation places of criticality accident detectors in the plutonium conversion development facility

    International Nuclear Information System (INIS)

    Sanada, Yukihisa; Tsujimura, Norio; Shimizu, Yoshio; Izaki, Kenji; Furuta, Sadaaki

    2008-01-01

    At the Plutonium Conversion Development Facility (PCDF) in the Nuclear Fuel Cycle Engineering Laboratories, the co-conversion technologies to purify the mixed plutonium and uranium nitrate solution discharged from a reprocessing plant have been developed. The probability of a criticality accident in PCDF is extremely low. However, the criticality accident alarm system (CAAS) has been in place since 1982 to reduce the radiation dose to workers in case of such a rare criticality accident. The CAAS contains criticality accident detector units (CADs), one unit consisting of three plastic scintillation detectors, and using the 2 out of 3 voting system for the purpose of high reliability. Currently, eight CADs are installed in PCDF evaluating the dose using a simple equation allowing for a safety margin. The purpose of this study is to show the determination procedures for the adequate relocation of the CADs which adequately ensures safety in PCDF. (author)

  7. Criticality safety evaluation of the fuel cycle facility electrorefiner

    International Nuclear Information System (INIS)

    Lell, R.M.; Mariani, R.D.; Fujita, E.K.; Benedict, R.W.; Turski, R.B.

    1993-01-01

    The integral Fast Reactor (IFR) being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal cooled reactors and a closed-loop fuel cycle. Some of the primary advantages are passive safety for the reactor and resistance to diversion for the heavy metal in the fuel cycle. in addition, the IFR pyroprocess recycles all the long-lived actinide activation products for casting into new fuel pins so that they may be burned in the reactor. A key component in the Fuel Cycle Facility (FCF) recycling process is the electrorefiner (ER) in which the actinides are separated from the fission products. In the process, the metal fuel is electrochemically dissolved into a high-temperature molten salt, and electrorefined uranium or uranium/plutonium products are deposited at cathodes. This report addresses the new and innovative aspects of the criticality analysis ensuing from processing metallic fuel, rather than metal oxide fuel, and from processing the spent fuel in batch operations. in particular, the criticality analysis employed a mechanistic approach as opposed to a probabilistic one. A probabilistic approach was unsuitable because of a lack of operational experience with some of the processes, rendering the estimation of accident event risk factors difficult. The criticality analysis also incorporated the uncertainties in heavy metal content attending the process items by defining normal operations envelopes (NOES) for key process parameters. The goal was to show that reasonable process uncertainties would be demonstrably safe toward criticality for continuous batch operations provided the key process parameters stayed within their NOES. Consequently the NOEs became the point of departure for accident events in the criticality analysis

  8. Criticality safety studies for plutonium–uranium metal fuel pin fabrication facility

    International Nuclear Information System (INIS)

    Stephen, Neethu Hanna; Reddy, C.P.

    2013-01-01

    Highlights: ► Criticality safety limits for PUMP-F facility is identified. ► The fissile mass which can be handled safely during alloy preparation is 10.5 kg. ► The number of fuel slugs which can be handled safely during injection casting is 53. ► The number of fuel slugs which can be handled safely after fuel fabrication is 71. - Abstract: This study focuses on the criticality safety during the fabrication of fast reactor metal fuel pins comprising of the fuel type U–15Pu, U–19Pu and U–19Pu–6Zr in the Plutonium–Uranium Metal fuel Pin fabrication Facility (PUMP-F). Maximum amount of fissile mass which can be handled safely during master alloy preparation, Injection casting and fuel slug preparation following fuel pin fabrication were identified and fixed based on this study. In the induction melting furnace, the fissile mass can be limited to 10.5 kg. During fuel slug preparation and fuel pin fabrication, fuel slugs and pins were arranged in hexagonal and square lattices to identify the most reactive configuration. The number of fuel slugs which can be handled safely after injection casting can be fixed to be 53, whereas after fuel fabrication it is 71

  9. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  10. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    International Nuclear Information System (INIS)

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs

  11. The potential for criticality following disposal of uranium at low-level waste facilities: Uranium blended with soil

    Energy Technology Data Exchange (ETDEWEB)

    Toran, L.E.; Hopper, C.M.; Naney, M.T. [and others

    1997-06-01

    The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop achievable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM), and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team`s approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some achievable scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via sorption or precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits. The analysis of SNM was restricted to {sup 235}U in the present scope of work. The outcome of the work indicates that criticality is possible given established regulatory limits on SNM disposal. However, a review based on actual disposal records of an existing site operation indicates that the potential for criticality is not a concern under current burial practices.

  12. The potential for criticality following disposal of uranium at low-level waste facilities: Uranium blended with soil

    International Nuclear Information System (INIS)

    Toran, L.E.; Hopper, C.M.; Naney, M.T.

    1997-06-01

    The purpose of this study was to evaluate whether or not fissile uranium in low-level-waste (LLW) facilities can be concentrated by hydrogeochemical processes to permit nuclear criticality. A team of experts in hydrology, geology, geochemistry, soil chemistry, and criticality safety was formed to develop achievable scenarios for hydrogeochemical increases in concentration of special nuclear material (SNM), and to use these scenarios to aid in evaluating the potential for nuclear criticality. The team's approach was to perform simultaneous hydrogeochemical and nuclear criticality studies to (1) identify some achievable scenarios for uranium migration and concentration increase at LLW disposal facilities, (2) model groundwater transport and subsequent concentration increase via sorption or precipitation of uranium, and (3) evaluate the potential for nuclear criticality resulting from potential increases in uranium concentration over disposal limits. The analysis of SNM was restricted to 235 U in the present scope of work. The outcome of the work indicates that criticality is possible given established regulatory limits on SNM disposal. However, a review based on actual disposal records of an existing site operation indicates that the potential for criticality is not a concern under current burial practices

  13. Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

    International Nuclear Information System (INIS)

    Maudlin, P.J.; Maerker, R.E.

    1982-01-01

    A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility

  14. Supplementary neutron-flux calculations for the ORNL Pool Critical Assembly Pressure Vessel Facility

    Energy Technology Data Exchange (ETDEWEB)

    Maudlin, P.J.; Maerker, R.E.

    1982-01-01

    A three-dimensional Monte Carlo calculation using the MORSE code was performed to validate a procedure previously adopted in the ORNL discrete ordinate analysis of measurements made in the ORNL Pool Critical Assembly Pressure Vessel Facility. The results of these flux calculations agree, within statistical undertainties of about 5%, with those obtained from a discrete ordinate analysis employing the same procedure. This study therefore concludes that the procedure for combining several one- and two-dimensional discrete ordinate calculations into a three-dimensional flux is sufficiently accurate that it does not account for the existing discrepancies observed between calculations and measurements in this facility.

  15. Dosimetry at the Los Alamos Critical Experiments Facility: Past, present, and future

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1993-10-01

    Although the primary reason for the existence of the Los Alamos Critical Experiments Facility is to provide basic data on the physics of systems of fissile material, the physical arrangements and ability to provide sources of radiation have led to applications for all types of radiation dosimetry. In the broad definition of radiation phenomena, the facility has provided sources to evaluate biological effects, radiation shielding and transport, and measurements of basic parameters such as the evaluation of delayed neutron parameters. Within the last 15 years, many of the radiation measurements have been directed to calibration and intercomparison of dosimetry related to nuclear criticality safety. Future plans include (1) the new applications of Godiva IV, a bare-metal pulse assembly, for dosimetry (including an evaluation of neutron and gamma-ray room return); (2) a proposal to relocate the Health Physics Research Reactor from the Oak Ridge National Laboratory to Los Alamos, which will provide the opportunity to continue the application of a primary benchmark source to radiation dosimetry; and (3) a proposal to employ SHEBA, a low-enrichment solution assembly, for accident dosimetry and evaluation

  16. Introduction to the nuclear criticality safety evaluation of facility X-705, Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    Sheaffer, M.K.; Keeton, S.C.

    1993-01-01

    This report is the first in a series of documents that will evaluate nuclear criticality safety in the Decontamination and Recovery Facility, X-705, Portsmouth Gaseous Diffusion Plant. It provides an overview of the facility, categorizes its functions for future analysis, reviews existing NCS documentation, and explains the follow-on effort planned for X-705. A detailed breakdown of systems, subsystems, and operational areas is presented and cross-referenced to existing NCS documentation

  17. The National Criticality Experiments Research Center at the Device Assembly Facility, Nevada National Security Site: Status and Capabilities, Summary Report

    International Nuclear Information System (INIS)

    Bragg-Sitton, S.; Bess, J.; Werner, J.

    2011-01-01

    The National Criticality Experiments Research Center (NCERC) was officially opened on August 29, 2011. Located within the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS), the NCERC has become a consolidation facility within the United States for critical configuration testing, particularly those involving highly enriched uranium (HEU). The DAF is a Department of Energy (DOE) owned facility that is operated by the National Nuclear Security Agency/Nevada Site Office (NNSA/NSO). User laboratories include the Lawrence Livermore National Laboratory (LLNL) and Los Alamos National Laboratory (LANL). Personnel bring their home lab qualifications and procedures with them to the DAF, such that non-site specific training need not be repeated to conduct work at DAF. The NNSS Management and Operating contractor is National Security Technologies, LLC (NSTec) and the NNSS Safeguards and Security contractor is Wackenhut Services. The complete report provides an overview and status of the available laboratories and test bays at NCERC, available test materials and test support configurations, and test requirements and limitations for performing sub-critical and critical tests. The current summary provides a brief summary of the facility status and the method by which experiments may be introduced to NCERC.

  18. Isotopic dilution requirements for 233U criticality safety in processing and disposal facilities

    International Nuclear Information System (INIS)

    Elam, K.R.; Forsberg, C.W.; Hopper, C.M.; Wright, R.Q.

    1997-11-01

    The disposal of excess 233 U as waste is being considered. Because 233 U is a fissile material, one of the key requirements for processing 233 U to a final waste form and disposing of it is to avoid nuclear criticality. For many processing and disposal options, isotopic dilution is the most feasible and preferred option to avoid nuclear criticality. Isotopic dilution is dilution of fissile 233 U with nonfissile 238 U. The use of isotopic dilution removes any need to control nuclear criticality in process or disposal facilities through geometry or chemical composition. Isotopic dilution allows the use of existing waste management facilities, that are not designed for significant quantities of fissile materials, to be used for processing and disposing of 233 U. The amount of isotopic dilution required to reduce criticality concerns to reasonable levels was determined in this study to be ∼ 0.66 wt% 233 U. The numerical calculations used to define this limit consisted of a homogeneous system of silicon dioxide (SiO 2 ), water (H 2 O), 233 U, and depleted uranium (DU) in which the ratio of each component was varied to determine the conditions of maximum nuclear reactivity. About 188 parts of DU (0.2 wt% 235 U) are required to dilute 1 part of 233 U to this limit in a water-moderated system with no SiO 2 present. Thus, for the US inventory of 233 U, several hundred metric tons of DU would be required for isotopic dilution

  19. Initial tank calibration at NUCEF critical facility. 1. Measurement procedure and its result

    International Nuclear Information System (INIS)

    Yanagisawa, Hiroshi; Mineo, Hideaki; Tonoike, Kotaro; Takeshita, Isao; Hoshi, Katsuya; Hagiwara, Hiroyuki.

    1994-07-01

    Initial tank calibrations were carried out prior to hot operation of critical facilities in NUCEF: Nuclear Fuel Cycle Safety Engineering Research Facility, for the purpose of the nuclear material accountancy and control for the facility. Raw calibration data were collected from single run per one tank by measuring differential pressure with dip-tube systems, weight of calibration liquid (demineralized water) poured into the tank, temperature in the tank and so on, without operation of tank ventilation system. Volume and level data were obtained by applying density and buoyancy corrections to the raw data. As a result, the evaluated measurement errors of volume and level were small enough, e.g. within 0.2 lit. and 1.0 mm, respectively, for Pu accountancy tanks. This paper summarizes the above-mentioned measurement procedures, collected data, data correction procedures and evaluated measurement errors. (author)

  20. Safety analysis of the Los Alamos critical experiments facility: burst operation of Skua

    International Nuclear Information System (INIS)

    Orndoff, J.D.; Paxton, H.C.; Wimett, T.F.

    1979-05-01

    A detailed consideration of the Skua burst assembly is presented, thereby supplementing the facility safety analysis report covering the operation of other critical assemblies at Los Alamos. As with these assemblies the small fission-product inventory, ambient pressure, and moderate temperatures in Skua are amenable to straightforward measures to ensure the protection of the public

  1. Safety analysis of the Los Alamos critical experiments facility: burst operation of Skua

    International Nuclear Information System (INIS)

    Orndoff, J.D.; Paxton, H.C.; Wimett, T.F.

    1980-12-01

    Detailed consideration of the Skua burst assembly is provided, thereby supplementing the facility Safety Analysis Report covering the operation of other critical assemblies at the Los Alamos Scientific Laboratory. As with these assemblies the small fission-product inventory, ambient pressure, and moderate temperatures in Skua are amenable to straightforward measures to ensure the protection of the public

  2. Questionnaire survey report about the criticality accident at a nuclear fuel processing facility

    International Nuclear Information System (INIS)

    2000-01-01

    The Radiation Protection Section of the Japanese Society of Radiological Technology conducted a questionnaire survey on the criticality accident at the nuclear fuel processing facility in Tokai village on September 30, 1999 in order to identify factors related to the accident and consider countermeasures to deal with such accidents. The questionnaire was distributed to 347 members (122 facilities) of the Japanese Society of Radiological Technology who were working or living in Ibaraki Prefecture, and replies were obtained from 104 members (75 facilities). Questions to elicit the opinions of individuals were as following: method of obtaining information about the accident, knowledge about radiation, opinions about the accident, and requests directed to the Society. Questions regarding facilities concerned the following: communication after the accident, requests for dispatch to the accident site, and possession of radiometry devices. In regard to acquisition of information, 91 of the 104 members (87.5%) answered 'television or radios' followed by newspapers. Forty-five of 101 members were questioned about radiation exposure and radiation effects by the public. There were many opinions that accurate news should be provided rapidly, by the mass media. Many members (75%) felt that they lacked knowledge about radiation, reconfirming the importance of education and instruction concerning radiation. Dispatch was requested of 36 of the 75 facilities (48%), and 44 of 83 facilities (53%) owned radiometry instruments. (K.H.)

  3. Criticality assessment of initial operations at the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Ha, B.C.; Williamson, T.G.

    1993-01-01

    At the Savannah River Site (SRS), high level radioactive wastes will be immobilized into borosilicate glass for long term storage and eventual disposal. Since the waste feed streams contain uranium and plutonium, the Defense Waste Processing Facility (DWPF) process has been evaluated to ensure that a subcritical condition is maintained. It was determined that the risk of nuclear criticality in the DWPF during initial, sludge-only operations is minimal due to the dilute concentration of fissile material in the sludge combined with excess neutron absorbers

  4. Integrated network for structural integrity monitoring of critical components in nuclear facilities, RIMIS

    International Nuclear Information System (INIS)

    Roth, Maria; Constantinescu, Dan Mihai; Brad, Sebastian; Ducu, Catalin; Malinovschi, Viorel

    2008-01-01

    The round table aims to join specialists working in the research area of the Romanian R and D Institutes and Universities involved in structural integrity assessment of materials, especially those working in the nuclear field, together with the representatives of the end user, the Cernavoda NPP. This scientific event will offer the opportunity to disseminate the theoretical, experimental and modelling activities, carried out to date, in the framework of the National Program 'Research of Excellence', Module I 2006-2008, managed by the National Authority for Scientific Research. Entitled 'Integrated Network for Structural Integrity Monitoring of Critical Components in Nuclear Facilities, RIMIS, the project has two main objectives: 1. - to elaborate a procedure applicable to the structural integrity assessment of critical components used in Romanian nuclear facilities (CANDU type Reactor, Hydrogen Isotopes Separation installations); 2. - to integrate the national networking into a similar one of European level, and to enhance the scientific significance of Romanian R and D organisations as well as to increase the contribution in solving major issues of the nuclear field. The topics of the round table will be focused on: 1. Development of a Structural Integrity Assessment Methodology applicable to the nuclear facilities components; 2. Experimental investigation methods and procedures; 3. Numeric simulation of nuclear components behaviour; 4. Further activities to finalize the assessment procedure. Also participations and contributions to sustain the activity in the European Network NULIFE, FP6 will be discussed. (authors)

  5. Criticality control during conditioning of spent nuclear fuel in the Fuel Cycle Facility

    International Nuclear Information System (INIS)

    Lell, R.M.; Khalil, H.S.

    1994-01-01

    Spent nuclear fuel may be unacceptable for direct repository storage because of composition, enrichment, form, physical condition, or the presence of undesirable materials such as sodium. Fuel types which are not acceptable for direct storage must be processed or conditioned to produce physical forms which can safely be stored in a repository. One possible approach to conditioning is the pyroprocess implemented in the Fuel Cycle Facility (FCF) at Argonne National Laboratory-West. Conditioning of binary (U-Zr) and ternary (U-Pu-Zr) metallic fuels from the EBR-2 reactor is used to demonstrate the process. Criticality safety considerations limit batch sizes during the conditioning steps and provide one constraint on the final form of conditioned material. Criticality safety during conditioning is assured by the integration of criticality safety analysis, equipment design, process development, a measurement program, accountability procedures, and a computerized Mass Tracking System. Criticality issues related to storage and shipment of conditioned material have been examined

  6. Critical need for MFE: the Alcator DX advanced divertor test facility

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

  7. Final report, November 1958--June 30, 1974

    International Nuclear Information System (INIS)

    1974-01-01

    A summary of research work at Rensselaer Polytechnic Institute undertaken in the period November 1958 to June 1974 is reported. Also included are a description of the facility, listings of research projects and publications, and identification of personnel

  8. Risk management for operations of the LANL Critical Experiments Facility

    International Nuclear Information System (INIS)

    Paternoster, R.; Butterfield, K.

    1998-01-01

    The Los Alamos Critical Experiments Facility (LACEF) currently operates two burst reactors (Godiva-IV and Skua), one solution assembly [the Solution High-Energy Burst Assembly (SHEBA)], two fast-spectrum benchmark assemblies (Flattop and Big Ten), and five general-purpose remote assembly machines that may be configured with nuclear materials and assembled by remote control. Special nuclear materials storage vaults support these and other operations at the site. With this diverse set of operations, several approaches are possible in the analysis and management of risk. The most conservative approach would be to write a safety analysis report (SAR) for each assembly and experiment. A more cost-effective approach is to analyze the probability and consequences of several classes of operations representative of operations on each critical assembly machine and envelope the bounding case accidents. Although the neutron physics of these machines varies widely, the operations performed at LACEF fall into four operational modes: steady-state mode, approach-to-critical mode, prompt burst mode, and nuclear material operations, which can include critical assembly fuel loading. The operational sequences of each mode are very nearly identical, whether operated on one assembly machine or another. The use of an envelope approach to accident analysis is facilitated by the use of classes of operations and the use of bounding case consequence analysis. A simple fault tree analysis of operational modes helps resolve which operations are sensitive to human error and which are initiated by hardware of software failures. Where possible, these errors and failures are blocked by TSR LCOs. Future work will determine the probability of accidents with various initiators

  9. NENIMF: Northeast National Ion Microprobe Facility - A Multi-User Facility for SIMS Microanalysis

    Science.gov (United States)

    Layne, G. D.; Shimizu, N.

    2002-12-01

    The MIT-Brown-Harvard Regional Ion Microprobe Facility was one of the earliest multi-user facilities enabled by Dan Weill's Instrumentation and Facilities Program - and began with the delivery of a Cameca IMS 3f ion microprobe to MIT in 1978. The Northeast National Ion Microprobe Facility (NENIMF) is the direct descendant of this original facility. Now housed at WHOI, the facility incorporates both the original IMS 3f, and a new generation, high transmission-high resolution instrument - the Cameca IMS 1270. Purchased with support from NSF, and from a consortium of academic institutions in the Northeast (The American Museum of Natural History, Brown University, The Lamont-Doherty Earth Observatory, MIT, Rensselaer Polytechnic Institute, WHOI) - this latest instrument was delivered and installed during 1996. NENIMF continues to be supported by NSF EAR I&F as a multi-user facility for geochemical research. Work at NENIMF has extended the original design strength of the IMS 1270 for microanalytical U-Pb zircon geochronology to a wide variety of novel and improved techniques for geochemical research. Isotope microanalysis for studies in volcanology and petrology is currently the largest single component of facility activity. This includes the direct measurement of Pb isotopes in melt inclusions, an application developed at NENIMF, which is making an increasingly significant contribution to our understanding of basalt petrogenesis. This same technique has also been extended to the determination of Pb isotopes in detrital feldspar grains, for the study of sedimentary provenance and tectonics of the Himalayas and other terrains. The determination of δ11B in volcanic melt inclusions has also proven to be a powerful tool in the modeling of subduction-related magmatism. The recent development of δ34S and δ37Cl determination in glasses is being applied to studies of the behavior of these volatile elements in both natural and experimental systems. Other recent undertakings

  10. Epidemiological studies of leukaemia in children and young adults around nuclear facilities: a critical review

    International Nuclear Information System (INIS)

    2008-01-01

    An epidemiological study published in late 2007 described an increased risk of leukaemia in children under 5 living within 5 kilometres of German nuclear power plants. A great deal of research has been carried out on this subject since the early 1980's. The aim of this report was to provide a synthesis and critical analysis of results related to the risk of leukaemia in children and young adults aged under 25 living close to nuclear facilities. The report is structured in three sections: - a reminder of the main characteristics of childhood leukaemia and a description of the methods used to conduct epidemiological studies; - the most exhaustive review possible of epidemiological studies published in the international literature describing the frequency of leukaemia close to nuclear facilities in different countries around the world. A critical analysis is made of the published results. Some results from studies not focused on nuclear facilities are also presented. The methodological limitations associated with descriptive studies are explained and discussed; - the last section discusses the possible causes of childhood leukaemia and the main hypotheses explored to explain certain clusters of cases observed locally close to some nuclear sites. Appendices at the end of the document provide additional explanations of the concepts and methods used in epidemiology and statistics, and of the classification of malignant hemopathies. (authors)

  11. Status of criticality safety research at NUCEF

    Energy Technology Data Exchange (ETDEWEB)

    Nakajima, Ken [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Two critical facilities, named STACY (Static Experiment Critical Facility) and TRACY (Transient Experiment Critical Facility), at the Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) started their hot operations in 1995. Since then, basic experimental data for criticality safety research have been accumulated using STACY, and supercritical experiments for the study of criticality accident in a reprocessing plant have been performed using TRACY. In this paper, the outline of those critical facilities and the main results of TRACY experiments are presented. (author)

  12. Risk management for operations of the Los Alamos critical experiments facility

    International Nuclear Information System (INIS)

    Paternoster, R.; Butterfield, K.

    1998-01-01

    The Los Alamos Critical Experiments Facility (LACEF) currently operates two burst reactors (Godiva-IV and Skua), one solution assembly (SHEBA 2--Solution high-Energy Burst Assembly), two fast-spectrum benchmark assemblies (Flattop and Big Ten), and five general-purpose remote assembly machines which may be configured with nuclear materials and assembled by remote control. SNM storage vaults support these and other operations at the site. With this diverse set of operations, several approaches are possible in the analysis and management of risk. The most conservative approach would be to write a safety analysis report (SAR) for each assembly and experiment. A more cost-effective approach is to analyze the probability and consequences of several classes of operations representative of operations on each critical assembly machine and envelope the bounding case accidents. Although the neutron physics of these machines varies widely, the operations performed at LACEF fall into four operational modes: steady-state mode, approach-to-critical mode, prompt burst mode, and nuclear material operations which can include critical assembly fuel loading. The operational sequences of each mode are very nearly the same, whether operated on one assembly machine or another. The use of an envelope approach to accident analysis is facilitated by the use of classes of operations and the use of bounding case consequence analysis. A simple fault tree analysis of operational modes helps resolve which operations are sensitive to human error and which are initiated by hardware of software failures. Where possible, these errors and failures are blocked by TSR LCOs

  13. A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

    Energy Technology Data Exchange (ETDEWEB)

    Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Moroz, N. P.; Fomichenko, P. A.; Timoshinov, A. V. [National Research Center Kurchatov Institute (Russian Federation); Volkov, Yu. N. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

  14. A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

    International Nuclear Information System (INIS)

    Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A.; Moroz, N. P.; Fomichenko, P. A.; Timoshinov, A. V.; Volkov, Yu. N.

    2016-01-01

    The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

  15. Experimental Fuels Facility Re-categorization Based on Facility Segmentation

    Energy Technology Data Exchange (ETDEWEB)

    Reiss, Troy P.; Andrus, Jason

    2016-07-01

    The Experimental Fuels Facility (EFF) (MFC-794) at the Materials and Fuels Complex (MFC) located on the Idaho National Laboratory (INL) Site was originally constructed to provide controlled-access, indoor storage for radiological contaminated equipment. Use of the facility was expanded to provide a controlled environment for repairing contaminated equipment and characterizing, repackaging, and treating waste. The EFF facility is also used for research and development services, including fuel fabrication. EFF was originally categorized as a LTHC-3 radiological facility based on facility operations and facility radiological inventories. Newly planned program activities identified the need to receive quantities of fissionable materials in excess of the single parameter subcritical limit in ANSI/ANS-8.1, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors” (identified as “criticality list” quantities in DOE-STD-1027-92, “Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports,” Attachment 1, Table A.1). Since the proposed inventory of fissionable materials inside EFF may be greater than the single parameter sub-critical limit of 700 g of U-235 equivalent, the initial re-categorization is Hazard Category (HC) 2 based upon a potential criticality hazard. This paper details the facility hazard categorization performed for the EFF. The categorization was necessary to determine (a) the need for further safety analysis in accordance with LWP-10802, “INL Facility Categorization,” and (b) compliance with 10 Code of Federal Regulations (CFR) 830, Subpart B, “Safety Basis Requirements.” Based on the segmentation argument presented in this paper, the final hazard categorization for the facility is LTHC-3. Department of Energy Idaho (DOE-ID) approval of the final hazard categorization determined by this hazard assessment document (HAD) was required per the

  16. Long-term criticality control in radioactive waste disposal facilities using depleted uranium

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1997-01-01

    Plant photosynthesis has created a unique planetary-wide geochemistry - an oxidizing atmosphere with oxidizing surface waters on a planetary body with chemically reducing conditions near or at some distance below the surface. Uranium is four orders of magnitude more soluble under chemically oxidizing conditions than it is under chemically reducing conditions. Thus, uranium tends to leach from surface rock and disposal sites, move with groundwater, and concentrate where chemically reducing conditions appear. Earth's geochemistry concentrates uranium and can separate uranium from all other elements except oxygen, hydrogen (in water), and silicon (silicates, etc). Fissile isotopes include 235 U, 233 U, and many higher actinides that eventually decay to one of these two uranium isotopes. The potential for nuclear criticality exists if the precipitated uranium from disposal sites has a significant fissile enrichment, mass, and volume. The earth's geochemistry suggests that isotopic dilution of fissile materials in waste with 238 U is a preferred strategy to prevent long-term nuclear criticality in and beyond the boundaries of waste disposal facilities because the 238 U does not separate from the fissile uranium isotopes. Geological, laboratory, and theoretical data indicate that the potential for nuclear criticality can be minimized by diluting fissile materials with- 238 U to 1 wt % 235 U equivalent

  17. Outline of criticality safety research project

    International Nuclear Information System (INIS)

    Kobayashi, Iwao; Tachimori, Shoichi; Suzaki, Takenori; Takeshita, Isao; Miyoshi, Yoshinori; Nakajima, Ken; Sakurai, Satoshi; Yanagisawa, Hiroshi

    1987-01-01

    As the power generation capacity of LWRs in Japan increased, the establishment and development of nuclear fuel cycle have become the important subject. Conforming to the safety research project of the nation, the Japan Atomic Energy Research Institute has advanced the project of constructing a new research facility, that is, Nuclear Fuel Cycle Engineering Research Facility (NUCEF). In this facility, it is planned to carry out the research on criticality safety, upgraded reprocessing techniques, and the treatment and disposal of transuranium element wastes. In this paper, the subjects of criticality safety research and the research carried out with a criticality safety experiment facility which is expected to be installed in the NUCEF are briefly reported. The experimental data obtained from the criticality safety handbooks and published literatures in foreign countries are short of the data on the mixture of low enriched uranium and plutonium which is treated in the reprocessing of spent fuel from LWRs. The acquisition of the criticality data for various forms of fuel, the elucidation of the scenario of criticality accidents, and the soundness of the confinement system for gaseous fission products and plutonium are the main subjects. The Static Criticality Safety Facility, Transient Criticality Safety Facility and pulse column system are the main facilities. (Kako, I.)

  18. Smart Power Supply Systems for Mission Critical Facilities

    Science.gov (United States)

    Hirose, Keiichi; Babasaki, Tadatoshi

    To develop the advanced and rich life, and the also economy and social activity continuously, various types of energy are necessary. At the same time, to protect the global environment and to prevent the depletion of natural resources, the effective and moreover efficient use of energy is becoming important. Electric power is one of the most important forms of energy for our life and society. This paper describes topics and survey results of technical trends regarding the electric power supply systems which are playing a core role as the important infrastructure to support the emergence of information-oriented society. Specifically, the power supply systems that enhance high power quality and reliability (PQR) are important for the steady growth of information and communication services. The direct current (DC) power, which has been used for telecommunications power systems and information and communications technologies (ICT), enables existing utilities' grid and distributed energy resources to keep a balance between supply and demand of small-scaled power systems or microgirds. These techniques are expected to be part of smartgrid technologies and facilitate the installation of distributed generators in mission critical facilities.

  19. Methods for developing seismic and extreme wind-hazard models for evaluating critical structures and equipment at US Department of Energy facilities and commercial plutonium facilities in the United States

    International Nuclear Information System (INIS)

    Coats, D.W.; Murray, R.C.; Bernreuter, D.L.

    1981-01-01

    Lawrence Livermore National Laboratory (LLNL) is developing seismic and wind hazard models for the US Department of Energy (DOE). The work is part of a three-phase effort to establish building design criteria developed with a uniform methodology for seismic and wind hazards at the various DOE sites throughout the United States. In Phase 1, LLNL gathered information on the sites and their critical facilities, including nuclear reactors, fuel-reprocessing plants, high-level waste storage and treatment facilities, and special nuclear material facilities. Phase 2 - development of seismic and wind hazard models - is discussed in this paper, which summarizes the methodologies used by seismic and extreme-wind experts and gives sample hazard curves for the first sites to be modeled. These hazard models express the annual probability that the site will experience an earthquake (or windspeed) greater than some specified magnitude. In the final phase, the DOE will use the hazards models and LLNL-recommended uniform design criteria to evaluate critical facilities. The methodology presented in this paper also was used for a related LLNL study - involving the seismic assessment of six commercial plutonium fabrication plants licensed by the US Nuclear Regulatory Commission (NRC). Details and results of this reassessment are documented in reference

  20. An Interactive Graphics Program for Assistance in Learning Convolution.

    Science.gov (United States)

    Frederick, Dean K.; Waag, Gary L.

    1980-01-01

    A program has been written for the interactive computer graphics facility at Rensselaer Polytechnic Institute that is designed to assist the user in learning the mathematical technique of convolving two functions. Because convolution can be represented graphically by a sequence of steps involving folding, shifting, multiplying, and integration, it…

  1. RELAP5 simulations of critical break experiments in the RD-14 test facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, I-G; Cho, Y-J; Lee, S [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1996-12-31

    RELAP5/MOD3 simulations of critical break tests in the RD-14 facility, modelling a loss of coolant in a CANDU reactor, were compared to the experimental results, and to CATHENA simulations of the early stage of the test. The RELAP5/MOD3 predicted thermal hydraulic behaviour reasonably well, but some discrepancies were observed after emergency cooling injection (ECI). Pressure differences between headers govern flow through the heated sections, particularly after ECI, and there is much uncertainty in the header pressures; further work is therefore recommended. 6 refs., 3 figs.

  2. Impact of Distributed Energy Resources on the Reliability of a Critical Telecommunications Facility

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, D.; Atcitty, C.; Zuffranieri, J.; Arent, D.

    2006-03-01

    Telecommunications has been identified by the Department of Homeland Security as a critical infrastructure to the United States. Failures in the power systems supporting major telecommunications service nodes are a main contributor to major telecommunications outages, as documented by analyses of Federal Communications Commission (FCC) outage reports by the National Reliability Steering Committee (under auspices of the Alliance for Telecommunications Industry Solutions). There are two major issues that are having increasing impact on the sensitivity of the power distribution to telecommunication facilities: deregulation of the power industry, and changing weather patterns. A logical approach to improve the robustness of telecommunication facilities would be to increase the depth and breadth of technologies available to restore power in the face of power outages. Distributed energy resources such as fuel cells and gas turbines could provide one more onsite electric power source to provide backup power, if batteries and diesel generators fail. But does the diversity in power sources actually increase the reliability of offered power to the office equipment, or does the complexity of installing and managing the extended power system induce more potential faults and higher failure rates? This report analyzes a system involving a telecommunications facility consisting of two switch-bays and a satellite reception system.

  3. Use of critical pathway models and log-normal frequency distributions for siting nuclear facilities

    International Nuclear Information System (INIS)

    Waite, D.A.; Denham, D.H.

    1975-01-01

    The advantages and disadvantages of potential sites for nuclear facilities are evaluated through the use of environmental pathway and log-normal distribution analysis. Environmental considerations of nuclear facility siting are necessarily geared to the identification of media believed to be sifnificant in terms of dose to man or to be potential centres for long-term accumulation of contaminants. To aid in meeting the scope and purpose of this identification, an exposure pathway diagram must be developed. This type of diagram helps to locate pertinent environmental media, points of expected long-term contaminant accumulation, and points of population/contaminant interface for both radioactive and non-radioactive contaminants. Confirmation of facility siting conclusions drawn from pathway considerations must usually be derived from an investigatory environmental surveillance programme. Battelle's experience with environmental surveillance data interpretation using log-normal techniques indicates that this distribution has much to offer in the planning, execution and analysis phases of such a programme. How these basic principles apply to the actual siting of a nuclear facility is demonstrated for a centrifuge-type uranium enrichment facility as an example. A model facility is examined to the extent of available data in terms of potential contaminants and facility general environmental needs. A critical exposure pathway diagram is developed to the point of prescribing the characteristics of an optimum site for such a facility. Possible necessary deviations from climatic constraints are reviewed and reconciled with conclusions drawn from the exposure pathway analysis. Details of log-normal distribution analysis techniques are presented, with examples of environmental surveillance data to illustrate data manipulation techniques and interpretation procedures as they affect the investigatory environmental surveillance programme. Appropriate consideration is given these

  4. Identification and categorisation of critical digital assets of I and C systems at nuclear facilities: implementation guide - TAFICS/IG/1

    International Nuclear Information System (INIS)

    2015-06-01

    This document is the first in a series of documents being developed by TAFICS for protecting computer-based I and C systems of Indian nuclear facilities from cyber attacks. This document identifies the Indian nuclear facilities and the types of computer systems within facilities - called Critical Digital Assets (CDA) - that are to be covered by security program. It also describes the process for identification and categorisation of CDA. The document covers operational facilities - such as reactors - as well as development facilities - such as I and C design organisations. The CDA identification and categorisation would help to implement a robust security program in a graded manner - as stipulated by international standards such as that of IAEA. It is recommended that all applicable Indian nuclear facilities should implement the process described in this document to generate a list of CD As of the respective facility. (author)

  5. FLUX SENSOR EVALUATIONS AT THE ATR CRITICAL FACILITY

    International Nuclear Information System (INIS)

    Unruh, Troy; Rempe, Joy; Nigg, David; Imel, George; Harris, Jason; Bonebrake, Eric

    2010-01-01

    The Advanced Test Reactor (ATR) and the ATR Critical (ATRC) facilities lack real-time methods for detecting thermal neutron flux and fission reaction rates for irradiation capsules. Direct measurements of the actual power deposited into a test are now possible without resorting to complicated correction factors. In addition, it is possible to directly measure minor actinide fission reaction rates and to provide time-dependent monitoring of the fission reaction rate or fast/thermal flux during transient testing. A joint Idaho State University/Idaho National Laboratory ATR National Scientific User Facility (ATR NSUF) project was recently initiated to evaluate new real-time state-of-the-art in-pile flux detection sensors. Initially, the project is comparing the accuracy, response time, and long duration performance of French Atomic Energy Commission (CEA)-developed miniature fission chambers, specialized self-powered neutron detectors (SPNDs) by the Argentinean National Energy Commission (CNEA), specially developed commercial SPNDs, and back-to-back fission (BTB) chambers developed by Argonne National Laboratory (ANL). As discussed in this paper, specialized fixturing and software was developed by INL to facilitate these joint ISU/INL evaluations. Calculations were performed by ISU to assess the performance of and reduce uncertainties in flux detection sensors and compare data obtained from these sensors with existing integral methods employed at the ATRC. Ultimately, project results will be used to select the detector that can provide the best online regional ATRC power measurement. It is anticipated that project results may offer the potential to increase the ATRC's current power limit and its ability to perform low-level irradiation experiments. In addition, results from this effort will provide insights about the viability of using these detectors in the ATR. Hence, this effort complements current activities to improve ATR software tools, computational protocols

  6. Reliability assessment of shut-off rod drive mechanism for TAPP - 3 and 4 and critical facility through life cycle testing

    International Nuclear Information System (INIS)

    Singh, Manjit; Badodkar, D.N.; Singh, N.K.; Dalal, N.S.; Mishra, M.K.; Veda Vyas, G.; Kothari, C.B.; Rao, V.V.S.S.; Saraf, R.K.

    2006-01-01

    Shut-off rod drive mechanism forms a safety critical system of a nuclear reactor. It is the space constraints for the given reactor layout, which makes design of shut-off rod drive mechanism (SRDM) a custom built design. Design of SRDM adopts fail-safe, replaceability and the simplicity criterion ensuring very high reliability of its operation. Shut-off rod drive mechanism for TAPP-3 and 4 and 'Critical Facility' have been recently designed and developed at Division of Remote Handling and Robotics (DRHR), BARC. These are designed with a number of advanced features and these are significantly different than those used in Dhruva and 220 MWe PHWRs. Design of SRDM is qualified through proto typing and life cycle testing on a full-scale test set-up. This paper gives details of qualification and life cycle test data for prototype SRDM for TAPP-3 and 4 and 'Critical Facility' and reliability assessment. (author)

  7. Pool critical assembly pressure vessel facility benchmark

    International Nuclear Information System (INIS)

    Remec, I.; Kam, F.B.K.

    1997-07-01

    This pool critical assembly (PCA) pressure vessel wall facility benchmark (PCA benchmark) is described and analyzed in this report. Analysis of the PCA benchmark can be used for partial fulfillment of the requirements for the qualification of the methodology for pressure vessel neutron fluence calculations, as required by the US Nuclear Regulatory Commission regulatory guide DG-1053. Section 1 of this report describes the PCA benchmark and provides all data necessary for the benchmark analysis. The measured quantities, to be compared with the calculated values, are the equivalent fission fluxes. In Section 2 the analysis of the PCA benchmark is described. Calculations with the computer code DORT, based on the discrete-ordinates method, were performed for three ENDF/B-VI-based multigroup libraries: BUGLE-93, SAILOR-95, and BUGLE-96. An excellent agreement of the calculated (C) and measures (M) equivalent fission fluxes was obtained. The arithmetic average C/M for all the dosimeters (total of 31) was 0.93 ± 0.03 and 0.92 ± 0.03 for the SAILOR-95 and BUGLE-96 libraries, respectively. The average C/M ratio, obtained with the BUGLE-93 library, for the 28 measurements was 0.93 ± 0.03 (the neptunium measurements in the water and air regions were overpredicted and excluded from the average). No systematic decrease in the C/M ratios with increasing distance from the core was observed for any of the libraries used

  8. 2010 Criticality Accident Alarm System Benchmark Experiments At The CEA Valduc SILENE Facility

    International Nuclear Information System (INIS)

    Miller, Thomas Martin; Dunn, Michael E.; Wagner, John C.; McMahan, Kimberly L.; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Piot, Jerome; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Masse, Veronique; Trama, Jean-Christophe; Gagnier, Emmanuel; Naury, Sylvie; Lenain, Richard; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

    2011-01-01

    Several experiments were performed at the CEA Valduc SILENE reactor facility, which are intended to be published as evaluated benchmark experiments in the ICSBEP Handbook. These evaluated benchmarks will be useful for the verification and validation of radiation transport codes and evaluated nuclear data, particularly those that are used in the analysis of CAASs. During these experiments SILENE was operated in pulsed mode in order to be representative of a criticality accident, which is rare among shielding benchmarks. Measurements of the neutron flux were made with neutron activation foils and measurements of photon doses were made with TLDs. Also unique to these experiments was the presence of several detectors used in actual CAASs, which allowed for the observation of their behavior during an actual critical pulse. This paper presents the preliminary measurement data currently available from these experiments. Also presented are comparisons of preliminary computational results with Scale and TRIPOLI-4 to the preliminary measurement data.

  9. Operation and utilization of low power research reactor critical facility for Advanced Heavy Water Reactor (AHWR)

    International Nuclear Information System (INIS)

    De, S.K.; Karhadkar, C.G.

    2017-01-01

    An Advanced Heavy Water Reactor (AHWR) has been designed and developed for maximum power generation from thorium considering large reserves of thorium. The design envisages using 54 pin MOX cluster with different enrichment of "2"3"3U and Pu in Thoria fuel pins. Theoretical models developed to neutron transport and the geometrical details of the reactor including all reactivity devices involve approximations in modelling, resulting in uncertainties. With a view to minimize these uncertainties, a low power research reactor Critical Facility was built in which cold clean fuel can be arranged in a desired and precise geometry. Different experiments conducted in this facility greatly contribute to understand and validate the physics design parameters

  10. Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

    Energy Technology Data Exchange (ETDEWEB)

    Maerker, R.E.; Maudlin, P.J.

    1981-02-01

    A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment.

  11. Supplementary neutron flux calculations for the ORNL pool critical assembly pressure vessel facility

    International Nuclear Information System (INIS)

    Maerker, R.E.; Maudlin, P.J.

    1981-02-01

    A three-dimensional Monte Carlo calculation was performed to estimate the neutron flux in the 8/7 configuration of the ORNL Pool Critical Assembly Pressure Vessel Facility. The calculational tool was the multigroup transport code MORSE operated in the adjoint mode. The MORSE flux results compared well with those using a previously adopted procedure for constructing a three-dimensional flux from one- and two-dimensional discrete ordinates calculations using the DOT-IV code. This study concluded that use of these discrete ordinates constructions in previous calculations is sufficiently accurate and does not account for the existing discrepancies between calculation and experiment

  12. Outline of NUCEF facility

    International Nuclear Information System (INIS)

    Takeshita, Isao

    1996-01-01

    NUCEF is a multipurpose research facility in the field of safety and advanced technology of nuclear fuel cycle back-end. Various experiment facilities and its supporting installations, in which nuclear fuel materials, radio isotopes and TRU elements can be handled, are arranged in more than one hundred rooms of two experiment buildings. Its construction was completed in middle of 1994 and hot experiments have been started since then. NUCEF is located on the site (30,000 m 2 ) of southeastern part in the Tokai Research Establishment of JAERI facing to the Pacific Ocean. The base of Experiment Buildings A and B was directly founded on the rock existing at 10-15 m below ground level taking the aseismatic design into consideration. Each building is almost same sized and composed of one basement and three floors of which area is 17,500 m 2 in total. In the basement, there are exhaust facilities of ventilation system, treatment system of solution fuel and radioactive waste solution and storage tanks of them. Major experiment facilities are located on the first or the second floors in each building. An air-inlet facility of ventilation system for each building is equipped on the third floor. Most of experiment facilities for criticality safety research including two critical facilities: Static Experiment Critical Facility (STACY) and Transient Experiment Critical Facility (TRACY) are installed in Experiment Building A. Experiment equipments for research on advanced fuel reprocessing process and on TRU waste management, which are named BECKY (Back End Fuel Cycle Key Elements Research Facility), are installed in laboratories and a-g cells in Experiment Building B. (J.P.N.)

  13. International safeguards for fast critical facilities

    International Nuclear Information System (INIS)

    Gunderson, D.O.; Todd, J.L.

    1978-12-01

    It was concluded that practical routine inventory verification techniques can be effective in detecting protracted diversion but will not meet the seven-day timeliness criteria either for protracted or large one-time diversions. An effective international safeguards system requires a method of continuously monitoring facility activities either with instrumentation, inspectors, or a combination thereof. It was also concluded that a resident inspector is required at this type of facility because of the many nonroutine operations. However, a single inspector cannot adequately monitor all activities to assure that no diversion is taking place. The use of existing structural features and unattended monitoring at portals as well as surveillance by a resident inspector can provide an effective detection capability. A rapid special inventory verification is required following detection to verify any suspected diversion

  14. Impact of distributed energy resources on the reliability of a critical telecommunications facility.

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, David; Zuffranieri, Jason V.; Atcitty, Christopher B.; Arent, Douglas (National Renewable Energy Laboratory, Golden, CO)

    2006-03-01

    This report documents a probabilistic risk assessment of an existing power supply system at a large telecommunications office. The focus is on characterizing the increase in the reliability of power supply through the use of two alternative power configurations. Telecommunications has been identified by the Department of Homeland Security as a critical infrastructure to the United States. Failures in the power systems supporting major telecommunications service nodes are a main contributor to major telecommunications outages. A logical approach to improve the robustness of telecommunication facilities would be to increase the depth and breadth of technologies available to restore power in the face of power outages. Distributed energy resources such as fuel cells and gas turbines could provide one more onsite electric power source to provide backup power, if batteries and diesel generators fail. The analysis is based on a hierarchical Bayesian approach and focuses on the failure probability associated with each of three possible facility configurations, along with assessment of the uncertainty or confidence level in the probability of failure. A risk-based characterization of final best configuration is presented.

  15. The nuclear criticality information system's project to archive unpublished critical experiment data

    International Nuclear Information System (INIS)

    Koponen, B.L.; Doherty, A.L.; Clayton, E.D.

    1991-01-01

    Critical experiment facilities produced a large amount of important data during the past forty-five years. However, much useful data remains unpublished. The unpublished material exists in the form of experimenters' logbooks, notes, photographs, material descriptions, etc. These data could be important for computer code validation, understanding the physics of criticality, facility design, or for setting process limits. In the past, criticality specialists have been able to obtain unpublished details by direct contact with the experimenters. The closure of facilities and the loss of personnel is likely to lead to the loss of the facility records unless an effort is made to ensure that the records are preserved. It has been recognized for some time that the unpublished records of critical experiment facilities comprise a valuable resource, thus the Nuclear Criticality Information System (NCIS) is working to ensure that the records are preserved and made available via NCIS. As a first step in the archiving project, we identified criteria to help judge which series of experiments should be considered for archiving. Data that are used for validating calculations or the basis for subcritical limits in standards, handbooks, and guides are of particular importance. In this paper we will discuss the criteria for archiving, the priority list of experiments for archiving, and progress in developing an NCIS image database using current CD-ROM technology. (Author)

  16. CSER-98-002: Criticality analysis for the storage of special nuclear material sources and standards in the WRAP Facility

    International Nuclear Information System (INIS)

    Goldberg, H.J.

    1998-01-01

    The Waste Receiving and Processing (WRAP) Facility will store uranium and transuranic (TRU) sources and standards for certification that WRAP meets the requirements of the Quality Assurance Program Plan (QAPP) for the Waste Isolation Pilot Plant (WIPP). In addition, WRAP must meet internal requirements for testing and validation of measuring instruments for nondestructive assay (NDA). In order to be certified for WIPP, WRAP will participate in the NDA Performance Demonstration Program (PDP). This program is a blind test of the NDA capabilities for TRU waste. It is intended to ensure that the NDA capabilities of this facility satisfy the requirements of the quality assurance program plan for the WIPP. The PDP standards have been provided by the Los Alamos National Laboratory (LANL) for this program. These standards will be used in the WRAP facility. To internally check the accuracy and sensitivity of the NDA instruments, a further set of sources and standards will also be used by the facility. Each sealed source or standard will be referred to herein as a unit. Various combinations of these units will be placed in test drums and/or boxes which will be subject to their own limits until unloaded. There will be two sealed test drums with five grams of weapons grade plutonium loaded in them. These drums will be appropriately marked and will be subject to the unit limits rather than the drum limits. This analysis shows that the storage and use of special nuclear material sources and standards within the limited control facility of WRAP (Rooms 101 and 104) is safe from a criticality standpoint. With the form, geometry, and masses involved with this evaluation, a criticality is not possible. The limits given in Section 2 should be imposed on facility operations

  17. Physical, Structural and Operational Vulnerability of Critical Facilities in Valle de Chalco Solidaridad, Estado de Mexico, Mexico. Case of study: Avándaro, San Isidro and El Triunfo

    Science.gov (United States)

    Garcia Payne, D. G.; Novelo-Casanova, D. A.; Ponce-Pacheco, A. B.; Espinosa-Campos, O.; Huerta-Parra, M.; Reyes-Pimentel, T.; Rodriguez, F.; Benitez-Olivares, I.

    2010-12-01

    Valle de Chalco Solidaridad is located in Mexico City Metropolitan Area in Estado de Mexico, Mexico. In this town there is a sewage canal called “La Compañía”. A wall of this canal collapsed on February 5, 2010 due to heavy rains creating the flooding of four surrounding communities. It is important to point out that this area is frequently exposed to floods. In this work, we consider a critical facility as an essential structure for performance, health care and welfare within a community or/and as a place that can be used as shelter in case of emergency or disaster. Global vulnerability (the sum of the three measured vulnerabilities) of the 25 critical facilities identified in the locations of Avándaro, San Isidro and El Triunfo was assessed using the Community Vulnerability Assessment Tool developed by the National Oceanic and Atmospheric Administration (NOAA). For each critical facility we determined its operational, structural and physical vulnerabilities. For our analysis, we considered the four main natural hazards to which Valle de Chalco is exposed: earthquakes, floods, landslides and sinking. We considered five levels of vulnerability using a scale from 1 to 5, where values range from very low to very high vulnerability, respectively. A critical facilities database was generated by collecting general information for three categories: schools, government and church. Each facility was evaluated considering its location in relation to identified high-risk areas. Our results indicate that in average, the global vulnerability of all facilities is low, however, there are particular cases in which this global vulnerability is high. The average operational vulnerability of the three communities is moderate. The global structural vulnerability (sum of the structural vulnerability for the four analyzed hazards) is moderate. In particular, the structural vulnerability to earthquakes is low, to landslides is very low, to flooding is moderate and to sinking is

  18. Potential criticality accident at the General Electric Nuclear Fuel and Component Manufacturing Facility, May 29, 1991

    International Nuclear Information System (INIS)

    1991-08-01

    At the General Electric Nuclear Fuel and Component Manufacturing facility, located near Wilmington, North Carolina, on May 28 and 29, 1991, approximately 150 kilograms of uranium were inadvertently transferred from safe process tanks to an unsafe tank located at the waste treatment facility, thus creating the potential for a localized criticality safety problem. The excess uranium was ultimately safely recovered when the tank contents were centrifuged to remove the uranium-bearing material. Subsequently, the US Nuclear Regulatory Commission dispatched an Incident Investigation Team to determine what happened, to identify probable causes, and to make appropriate findings and conclusions. This report describes the incident, the methodology used by the team in its investigation, and presents the team's findings and conclusions. 48 figs., 8 tabs

  19. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    International Nuclear Information System (INIS)

    Watanabe, Norio; Tamaki, Hitoshi

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  20. Review and compilation of criticality accidents in nuclear fuel processing facilities outside of Japan

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Norio [Planning and Analysis Division, Nuclear Safety Research Center, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Tamaki, Hitoshi [Department of Safety Research Technical Support, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2000-04-01

    On September 30, 1999, a criticality accident occurred at the Tokai-mura uranium processing plant operated by JCO Co., Ltd., which resulted in the first nuclear accident involving a fatality, in Japan, and forced the residents in the vicinity of the site to be evacuated and be sheltered indoors. There have now been 21 criticality accidents reported in nuclear fuel processing facilities in foreign countries: seven in the United States, one in the United Kingdom and thirteen in Russia. Most of them occurred during the period from mid-1950's to mid-1960's, but one criticality accident tool place in Russian in 1997. This report reviews and compiles the published information on these accidents, including the latest information, focusing on the event sequence, the consequence of accident, and the cause of accident. The observations from the reviews are summarized as follows: Twenty of the 21 accidents occurred with the fissile material in a liquid. Twenty of the 21 accidents occurred in vessels/tanks with unfavorable geometry but one occurred in the vessel with favorable geometry. There were seven fatalities that were involved in five accidents. Three accidents involved a re-criticality condition caused by inadequate operator actions and two of them led to the death of the operators. One accident reached a re-criticality condition several hours after the first excursion was terminated by injecting borated water into the affected vessel. This accident implies the possibility that the borated water injection might not be effective to the criticality termination due to solubility of boric acid. Mechanisms of the criticality termination vary as follows: ejection or splashing of the solution at the time of power excursion, boiling or evaporation, addition of neutron poisons, or manual draining of solutions. (author)

  1. Accelerator-driven sub-critical research facility with low-enriched fuel in lead matrix: Neutron flux calculation

    Directory of Open Access Journals (Sweden)

    Avramović Ivana

    2007-01-01

    Full Text Available The H5B is a concept of an accelerator-driven sub-critical research facility (ADSRF being developed over the last couple of years at the Vinča Institute of Nuclear Sciences, Belgrade, Serbia. Using well-known computer codes, the MCNPX and MCNP, this paper deals with the results of a tar get study and neutron flux calculations in the sub-critical core. The neutron source is generated by an interaction of a proton or deuteron beam with the target placed inside the sub-critical core. The results of the total neutron flux density escaping the target and calculations of neutron yields for different target materials are also given here. Neutrons escaping the target volume with the group spectra (first step are used to specify a neutron source for further numerical simulations of the neutron flux density in the sub-critical core (second step. The results of the calculations of the neutron effective multiplication factor keff and neutron generation time L for the ADSRF model have also been presented. Neutron spectra calculations for an ADSRF with an uranium tar get (highest values of the neutron yield for the selected sub-critical core cells for both beams have also been presented in this paper.

  2. Engineering design guidelines for nuclear criticality safety

    International Nuclear Information System (INIS)

    Waltz, W.R.

    1988-08-01

    This document provides general engineering design guidelines specific to nuclear criticality safety for a facility where the potential for a criticality accident exists. The guide is applicable to the design of new SRP/SRL facilities and to major modifications Of existing facilities. The document is intended an: A guide for persons actively engaged in the design process. A resource document for persons charged with design review for adequacy relative to criticality safety. A resource document for facility operating personnel. The guide defines six basic criticality safety design objectives and provides information to assist in accomplishing each objective. The guide in intended to supplement the design requirements relating to criticality safety contained in applicable Department of Energy (DOE) documents. The scope of the guide is limited to engineering design guidelines associated with criticality safety and does not include other areas of the design process, such as: criticality safety analytical methods and modeling, nor requirements for control of the design process

  3. Criticality safety strategy for the Fuel Cycle Facility electrorefiner at Argonne National Laboratory, West

    International Nuclear Information System (INIS)

    Mariani, R.D.; Benedict, R.W.; Lell, R.M.; Turski, R.B.; Fujita, E.K.

    1993-01-01

    The Integral Fast Reactor being developed by Argonne National Laboratory (ANL) combines the advantages of metal-fueled, liquid-metal-cooled reactors and a closed fuel cycle. Presently, the Fuel Cycle Facility (FCF) at ANL-West in Idaho Falls, Idaho is being modified to recycle spent metallic fuel from Experimental Breeder Reactor II as part of a demonstration project sponsored by the Department of Energy. A key component of the FCF is the electrorefiner (ER) in which the actinides are separated from the fission products. In the electrorefining process, the metal fuel is anodically dissolved into a high-temperature molten salt and refined uranium or uranium/plutonium products are deposited at cathodes. In this report, the criticality safety strategy for the FCF ER is summarized. FCF ER operations and processes formed the basis for evaluating criticality safety and control during actinide metal fuel refining. In order to show criticality safety for the FCF ER, the reference operating conditions for the ER had to be defined. Normal operating envelopes (NOES) were then defined to bracket the important operating conditions. To keep the operating conditions within their NOES, process controls were identified that can be used to regulate the actinide forms and content within the ER. A series of operational checks were developed for each operation that wig verify the extent or success of an operation. The criticality analysis considered the ER operating conditions at their NOE values as the point of departure for credible and incredible failure modes. As a result of the analysis, FCF ER operations were found to be safe with respect to criticality

  4. Critical enrichment and critical density of infinite systems for nuclear criticality safety evaluation

    International Nuclear Information System (INIS)

    Naito, Yoshitaka; Koyama, Takashi; Komuro, Yuichi

    1986-03-01

    Critical enrichment and critical density of homogenous infinite systems, such as U-H 2 O, UO 2 -H 2 O, UO 2 F 2 aqueous solution, UO 2 (NO 3 ) 2 aqueous solution, Pu-H 2 O, PuO 2 -H 2 O, Pu(NO 3 ) 4 aqueous solution and PuO 2 ·UO 2 -H 2 O, were calculated with the criticality safety evaluation computer code system JACS for nuclear criticality safety evaluation on fuel facilities. The computed results were compared with the data described in European and American criticality handbooks and showed good agreement with each other. (author)

  5. Basic design of the test facility for the two-phase critical flow with non-condensable gas

    International Nuclear Information System (INIS)

    Chang, Seok Kyu; Kim, Chang Hwe; Chung, Chang Hwan

    2000-12-01

    The two-phase critical flow test with non-condensible gas is for the simulation of the critical flow phenomena which can be occurred during SB-LOCA on SMART reactor. The basic design of the test facility for the actual installation is performed from the basis of the previous conceptual design according to the test requirements. The 1.3m 3 pressure vessel has the circulation pipeline which contains pump(5m 3 /hr), main heater(150KW) and cooler for heating the working fluid to the test temperature within 6 hours. The N2 gas, water supply line are attached to the upper part and test section, flowmeter and various sensors are installed at the lower part of the pressure vessel. The suppression tank is for the storage and cooling of the discharged water. The N2 gas storage tank provides the system pressure to the pressure vessel during the test. The 0.7m 3 N2 gas injection tank supplies the required N2 gas to the entrance of the test section. Since these N2 supply systems require much amount of gas during short period, multistage valve systems and optimal control logics are needed and applied. For the filling of the N2 gas to the N2 storage tank, 5m 3 LN2 tank and related gas converting system were designed. The operating mode of the test facility can be classified to the starting, steady, main test and cooling modes and the proper monitoring and control logics are developed for each operating mode. The operation of the test facility is performed through the PLC and the acquisition of the test data is done with DAS

  6. DOE Lab-to-Lab MPC ampersand A workshop for cooperative tasks with Russian institutes: Focus on critical assemblies and item facilities

    International Nuclear Information System (INIS)

    Bieber, A.M. Jr.; Fishbone, L.G.; Kato, W.Y.; Lazareth, O.W.; Suda, S.C.; Garcia, D.; Haga, R.

    1995-01-01

    Seventeen Russian scientists and engineers representing five different institutes participated in a Workshop on material control and accounting as part of the US-Russian Lab-to-Lab Cooperative Program in Nuclear Materials Protection, Control, and Accounting (MPC ampersand A). In addition to presentations and discussions, the Workshop included an exercise at Brookhaven National Laboratory (BNL) and demonstrations at the Zero Power Physics Reactor (critical-assembly facility) of Argonne National Laboratory-West (ANL-W). The Workshop particularly emphasized procedures for physical inventory-taking at critical assemblies and item facilities, with associated supporting techniques and methods. By learning these topics and applying the methods and experience at their own institutes, the Russian scientists and engineers will be able to determine and verify nuclear material inventories based on sound procedures, including measurements. This will constitute a significant enhancement to MPC ampersand A at the Russian institutes

  7. Design philosophy for the new harwell α,β/γ ILW facility and associated NDA instrumentation with regard to criticality safety

    International Nuclear Information System (INIS)

    Chard, P.M.J.; Cooper, T.J.; Croft, S.; Lambert, K.P.; Syme, D.B.; Wilkins, C.G.

    1995-01-01

    A new Alpha Beta/Gamma Waste Facility is currently being commissioned at Harwell. The facility provides for the assay and re-packing of existing Intermediate Level Waste (ILW) cans and future ILW arisings into 400 litre drums which are then stored in an integral vault prior to ultimate disposal. Paramount to the design philosophy for the plant was the safe retrieval and movement of radioactive material throughout the treatment process and the avoidance of criticality and other hazards. This required sound managerial controls underpinned by state-of-the-art non destructive assay (NDA) measurements. These consist primarily of a gamma spectrometer and a passive/active neutron interrogator. Their prime role is to confirm can identity against plant records and enable a fissile inventory to be developed for each can for criticality assessment. An expert system aids interpretation of assay results and the reconciliation of discrepancies. This paper describes the design philosophy with emphasis on the control measures used and the operation of the expert system. (author)

  8. Outline of a fuel treatment facility in NUCEF

    International Nuclear Information System (INIS)

    Sugikawa, Susumu; Umeda, Miki; Kokusen, Junya

    1997-03-01

    This report presents outline of the nuclear fuel treatment facility for the purpose of preparing solution fuel used in Static Experiment Critical Facility (STACY) and Transient Experiment Critical Facility (TRACY) in Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF), including descriptions of process conditions and dimensions of major process equipments on dissolution system of oxide fuel, chemical adjustment system, purification system, acid recovery system, solution fuel storage system, and descriptions of safety design philosophy such as safety considerations of criticality, solvent fire, explosion of hydrogen and red-oil and so on. (author)

  9. A Framework for Seismic Design of Items in Safety-Critical Facilities for Implementing a Risk-Informed Defense-in-Depth-Based Concept

    Directory of Open Access Journals (Sweden)

    Tatsuya Itoi

    2017-05-01

    Full Text Available Recently, especially after the 2011 off the Pacific coast of Tohoku earthquake and the Fukushima Daiichi nuclear power plant accident, the need for treating residual risks and cliff-edge effects in safety-critical facilities has been widely recognized as an extremely important issue. In this article, the sophistication of seismic designs in safety-critical facilities is discussed from the viewpoint of mitigating the consequences of accidents, such as the avoidance of cliff-edge effects. For this purpose, the implementation of a risk-informed defense-in-depth-based framework is proposed in this study. A basic framework that utilizes diversity in the dynamic characteristics of items and also provides additional seismic margin to items important for safety when needed is proposed to prevent common cause failure and to avoid cliff-edge effects as far as practicable. The proposed method is demonstrated to be effective using an example calculation.

  10. Studies of neutron methods for process control and criticality surveillance of fissile material processing facilities

    International Nuclear Information System (INIS)

    Zoltowski, T.

    1988-01-01

    The development of radiochemical processes for fissile material processing and spent fuel handling need new control procedures enabling an improvement of plant throughput. This is strictly related to the implementation of continuous criticality control policy and developing reliable methods for monitoring the reactivity of radiochemical plant operations in presence of the process perturbations. Neutron methods seem to be applicable for fissile material control in some technological facilities. The measurement of epithermal neutron source multiplication with heuristic evaluation of measured data enables surveillance of anomalous reactivity enhancement leading to unsafe states. 80 refs., 47 figs., 33 tabs. (author)

  11. Program of nuclear criticality safety experiment at JAERI

    International Nuclear Information System (INIS)

    Kobayashi, Iwao; Tachimori, Shoichi; Takeshita, Isao; Suzaki, Takenori; Ohnishi, Nobuaki

    1983-11-01

    JAERI is promoting the nuclear criticality safety research program, in which a new facility for criticality safety experiments (Criticality Safety Experimental Facility : CSEF) is to be built for the experiments with solution fuel. One of the experimental researches is to measure, collect and evaluate the experimental data needed for evaluation of criticality safety of the nuclear fuel cycle facilities. Another research area is a study of the phenomena themselves which are incidental to postulated critical accidents. Investigation of the scale and characteristics of the influences caused by the accident is also included in this research. The result of the conceptual design of CSEF is summarized in this report. (author)

  12. Nuclear plant fire incident data file

    International Nuclear Information System (INIS)

    Sideris, A.G.; Hockenbury, R.W.; Yeater, M.L.; Vesely, W.E.

    1979-01-01

    A computerized nuclear plant fire incident data file was developed by American Nuclear Insurers and was further analyzed by Rensselaer Polytechnic Institute with technical and monetary support provided by the Nuclear Regulatory Commission. Data on 214 fires that occurred at nuclear facilities have been entered in the file. A computer program has been developed to sort the fire incidents according to various parameters. The parametric sorts that are presented in this article are significant since they are the most comprehensive statistics presently available on fires that have occurred at nuclear facilities

  13. The evaluation of set of criticality parameters using scale system

    International Nuclear Information System (INIS)

    Abe, Alfredo; Sanchez, Andrea; Yamaguchi, Mistuo

    2009-01-01

    In evaluating the criticality safety of the nuclear fuel facility, it is important to apply a consistent methodology, which consider every aspects concerning various types of criticality parameters. Usually, the critical parameters are compiled and arranged into handbooks, and these handbooks are based on experience with nuclear facilities, experimental data from criticality safety research facilities, and theoretical studies performed using numerical simulations. Most of criticality safety evaluation can be addressed using the criticality parameters data directly from handbook, but some critical parameters for a specific chemical mixtures and/or enrichment are not be available. Consequently, not available parameters has to be evaluated. This work present the methodology to evaluate a set of critical parameters using SCALE system for various types of mixtures present at nuclear fuel cycle facilities for two different level of enrichment, the results are verified in the independent calculation using MCNP Monte Carlo Code. (author)

  14. Prevention of criticality accidents

    International Nuclear Information System (INIS)

    Canavese, S.I.

    1982-01-01

    These notes used in the postgraduate course on Radiological Protection and Nuclear Safety discuss macro-and microscopic nuclear constants for fissile materials systems. Critical systems: their definition; criteria to analyze the critical state; determination of the critical size; analysis of practical problems about prevention of criticality. Safety of isolated units and of sets of units. Application of standards. Conception of facilities from the criticality control view point. (author) [es

  15. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  16. Understanding the application of knowledge management to the safety critical facilities

    International Nuclear Information System (INIS)

    Ilina, Elena

    2010-01-01

    Challenges to the operating nuclear power plants and transport infrastructures are outlined. It is concluded that most aggravating factors are related to knowledge. Thus, of necessity, effective knowledge management is required. Knowledge management theories are reviewed in their historical perspective as a natural extension and unification of information theories and theories about learning. The first line is identified with names as Wiener, Ashby, Shannon, Jaynes, Dretske, Harkevich. The second line - with Vygotsky, Engestroem, Carayannis. The recent developments of knowledge management theorists as Davenport, Prusak, Drew, Wiig, Zack are considered stressing learning, retaining of knowledge, approaching the state awareness of awareness, and alignment of knowledge management with the strategy of the concerned organizations. Further, some of the details and results are presented of what is achieved so far. More specifically, knowledge management tools are applied to the practical work activities as event reporting, data collection, condition assessment, verification of safety functions and incident investigation. Obstacles are identified and improvements are proposed. Finally, it is advised to continue to implement and further develop knowledge management tools in the organizations involved in various aspects of safety critical facilities

  17. Technical criteria for the prevention of criticality

    Energy Technology Data Exchange (ETDEWEB)

    Tomlinson, R E

    1971-04-15

    Atlantic Richfield Hanford Company (ARHCO) Policy Guide 1.6.6, 'Criticality Prevention' and Operating Instruction 1.6.6.2, 'Criticality Prevention in Process Facilities', present the policy of the Chemical Processing Division with respect to the control of criticality hazards, and delegate the responsibility for specifying safe limits for the design and operation of process facilities to the Manager, Research and Development Department. The purpose of this document is to define the technical criteria to be used in developing the limits within which CPD facilities are to be designed and operated. These criteria are based on the operating experience accumulated from the processing of fissile materials since the year 1944. The mere existence of a fissile material in quantities greater than a minimum critical mass creates some finite risk that criticality will occur. This risk of criticality can be held to a very low value by imposing restrictions on the manner in which the fissile material is stored or handled. Such controls are to be imposed as needed.

  18. Technical criteria for the prevention of criticality

    International Nuclear Information System (INIS)

    Tomlinson, R.E.

    1971-01-01

    Atlantic Richfield Hanford Company (ARHCO) Policy Guide 1.6.6, 'Criticality Prevention' and Operating Instruction 1.6.6.2, 'Criticality Prevention in Process Facilities', present the policy of the Chemical Processing Division with respect to the control of criticality hazards, and delegate the responsibility for specifying safe limits for the design and operation of process facilities to the Manager, Research and Development Department. The purpose of this document is to define the technical criteria to be used in developing the limits within which CPD facilities are to be designed and operated. These criteria are based on the operating experience accumulated from the processing of fissile materials since the year 1944. The mere existence of a fissile material in quantities greater than a minimum critical mass creates some finite risk that criticality will occur. This risk of criticality can be held to a very low value by imposing restrictions on the manner in which the fissile material is stored or handled. Such controls are to be imposed as needed

  19. Safety analysis of the critical facility for AHWR and 500 MWe PHWR

    International Nuclear Information System (INIS)

    Pushpam, Neelima Prasad; Arvind Kumar; Srivenkatesan, R.

    2002-01-01

    Full text: The initiating event for the design basis reactivity accident is the uncontrolled moderator pump up at criticality. This uncontrolled pump up transient is considered to be the enveloping scenario and has been analysed for the reference core with AHWR fuel using point kinetics model. It is the most reactive core among the three with a small b value (core average b= 5.96 mk). The maximum pump rate in critical facility is limited to 300 litres/min which corresponds to the maximum rate of reactivity addition about 0.1 mk/sec. On any reactor trip 6 shut-off rods are inserted into the core along with partial moderator dumps. The reactor is provided with independent safety and regulating channels (SC and RC) to monitor reactor neutronic power and initiate trip at different power levels. After the reactor trips five of the six fast acting shut-off rods (maximum worth rod is unavailable) fall under gravity and at the same time moderator dump is initiated. We have considered shut off rods and moderator dump as two independent shutdown systems. The analysis shows that even if the reactor trips at the high power at 550 watt ignoring the earlier trips, the fuel temperature does not rise beyond 50 degC and the total energy released is less than 20 kW. We also analysed the transients due to uncontrolled withdrawal of absorber rod. In this case also we found that the fuel temperature became ∼54 degC and the total energy release was about 25 kW. The fuel can withstand this temperature. This shows that reactor is safe

  20. Research on consequence analysis method for probabilistic safety assessment of nuclear fuel facilities (5). Evaluation method and trial evaluation of criticality accident

    International Nuclear Information System (INIS)

    Yamane, Yuichi; Abe, Hitoshi; Nakajima, Ken; Hayashi, Yoshiaki; Arisawa, Jun; Hayami, Satoru

    2010-01-01

    A special committee of 'Research on the analysis methods for accident consequence of nuclear fuel facilities (NFFs)' was organized by the Atomic Energy Society of Japan (AESJ) under the entrustment of Japan Atomic Energy Agency (JAEA). The committee aims to research on the state-of-the-art consequence analysis method for the Probabilistic Safety Assessment (PSA) of NFFs, such as fuel reprocessing and fuel fabrication facilities. The objectives of this research are to obtain information useful for establishing quantitative performance objectives and to demonstrate risk-informed regulation through qualifying issues needed to be resolved for applying PSA to NFFs. The research activities of the committee were mainly focused on the consequence analysis method for postulated accidents with potentially large consequences in NFFs, e.g., events of criticality, spill of molten glass, hydrogen explosion, boiling of radioactive solution and fire (including the rapid decomposition of TBP complexes), resulting in the release of radioactive materials to the environment. The results of the research were summarized in a series of six reports, which consist of a review report and five technical ones. In this report, the evaluation methods of criticality accident, such as simplified methods, one-point reactor kinetics codes and quasi-static method, were investigated and their features were summarized to provide information useful for the safety evaluation of NFFs. In addition, several trial evaluations were performed for a hypothetical scenario of criticality accident using the investigated methods, and their results were compared. The release fraction of volatile fission products in a criticality accident was also investigated. (author)

  1. Criticality experiments with fast flux test facility fuel pins

    International Nuclear Information System (INIS)

    Bierman, S.R.

    1990-11-01

    A United States Department of Energy program was initiated during the early seventies at the Hanford Critical Mass Laboratory to obtain experimental criticality data in support of the Liquid Metal Fast Breeder Reactor Program. The criticality experiments program was to provide basic physics data for clean well defined conditions expected to be encountered in the handling of plutonium-uranium fuel mixtures outside reactors. One task of this criticality experiments program was concerned with obtaining data on PuO 2 -UO 2 fuel rods containing 20--30 wt % plutonium. To obtain this data a series of experiments were performed over a period of about twelve years. The experimental data obtained during this time are summarized and the associated experimental assemblies are described. 8 refs., 7 figs

  2. New approach to managing nuclear criticality risk at Nuclear Fuel Services, Inc

    International Nuclear Information System (INIS)

    Green, R.; Droke, R.; Paine, D.

    1992-01-01

    The negative aspects of having a nuclear criticality accident at a fuel fabrication facility have substantially increased in recent years. Although ensuring that the facility is designed and operating in a critically safe manner is a high management priority, practices of managing the risk of a criticality accident have not significantly changed. The method of evaluating risk based on quantitative analysis can enable management to adapt to the increased consequences of a nuclear criticality accident. Additional controls may be placed on high-risk areas within a facility to ensure safe operation of the plant. Areas where controls are in place that impede the productivity of the facility and have negligible impact on criticality safety may be removed or replaced. Management can also streamline the safety analysis efforts applied to facility upgrades by demonstrating that proposed design changes do not compromise criticality safety. Future expansion of quantitative analysis techniques will also allow risk-based management decisions on industrial, radiological, and environmental safety

  3. 10 CFR 70.64 - Requirements for new facilities or new processes at existing facilities.

    Science.gov (United States)

    2010-01-01

    ... postulated accidents that could lead to loss of safety functions. (5) Chemical protection. The design must... 10 Energy 2 2010-01-01 2010-01-01 false Requirements for new facilities or new processes at... Critical Mass of Special Nuclear Material § 70.64 Requirements for new facilities or new processes at...

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

    International Nuclear Information System (INIS)

    Mitake, Susumu

    2003-01-01

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

  5. Criticality safety evaluation report for the Cold Vacuum Drying Facility's process water handling system

    International Nuclear Information System (INIS)

    Roblyer, S.D.

    1998-01-01

    This report addresses the criticality concerns associated with process water handling in the Cold Vacuum Drying Facility (CVDF). The controls and limitations on equipment design and operations to control potential criticality occurrences are identified. The effectiveness of equipment design and operation controls in preventing criticality occurrences during normal and abnormal conditions is evaluated and documented in this report. Spent nuclear fuel (SNF) is removed from existing canisters in both the K East and K West Basins and loaded into a multicanister overpack (MCO) in the K Basin pool. The MCO is housed in a shipping cask surrounded by clean water in the annulus between the exterior of the MCO and the interior of the shipping cask. The fuel consists of spent N Reactor and some single pass reactor fuel. The MCO is transported to the CVDF near the K Basins to remove process water from the MCO interior and from the shipping cask annulus. After the bulk water is removed from the MCO, any remaining free liquid is removed by drawing a vacuum on the MCO's interior. After cold vacuum drying is completed, the MCO is filled with an inert cover gas, the lid is replaced on the shipping cask, and the MCO is transported to the Canister Storage Building. The process water removed from the MCO contains fissionable materials from metallic uranium corrosion. The process water from the MCO is first collected in a geometrically safe process water conditioning receiver tank. The process water in the process water conditioning receiver tank is tested, then filtered, demineralized, and collected in the storage tank. The process water is finally removed from the storage tank and transported from the CVDF by truck

  6. Risk assessment methodology for extreme wind and missile effects on critical facilities

    International Nuclear Information System (INIS)

    Twisdale, L.A.; Dunn, W.L.

    1985-01-01

    The TORMIS methodology has been applied to a number of probabilistic risk assessments of critical facilities in the continental United States. These analyses have centered on the estimation of tornado missile impact and damage risks to individual targets as well as to groups of targets at specific plants. A number of advancements and generalizations in the approach have recently been made. These include: (1) generalization of windfield options to include straight winds (WINMIS) and hurricanes (HURMIS); (2) generalization of the scoring to enable analysis of Boolean system expressions for damage probabilities on compound series and parallel safety trains; (3) generalization of the failure criteria to include wind pressure as well as missile impact; (4) generalization of the plant modeling capability to enable more detailed treatment of targets partially or fully enclosed by vulnerable cladding and to allow tracking of missiles inside such enclosures; and (5) incorporation of windspeed criteria for structural failure and subsequent production of potential missiles. This paper will present some of the basic theory and key results of recent TORMIS, WINMIS, and HURMIS applications. The influence of uncertainties in the estimation process and the data needed for plant-specific risk assessments will also be discussed

  7. Preclosure Criticality Analysis Process Report

    International Nuclear Information System (INIS)

    Thomas, D.A.

    1999-01-01

    The design approach for criticality of the disposal container and waste package will be dictated by existing regulatory requirements. This conclusion is based on the fact that preclosure operations and facilities have significant similarities to existing facilities and operations currently regulated by the NRC. The major difference would be the use of a risk-informed approach with burnup credit. This approach could reduce licensing delays and costs of the repository. The probability of success for this proposed seamless licensing strategy is increased, since there is precedence of regulation (10 CFR Part 63 and NUREG 1520) and commercial precedence for allowing burnup credit at sites similar to Yucca Mountain during preclosure. While NUREG 1520 is not directly applicable to a facility for handling spent nuclear fuel, the risk-informed approach to criticality analysis in NUREG 1520 is considered indicative of how the NRC will approach risk-informed criticality analysis at spent fuel facilities in the future. The types of design basis events which must be considered during the criticality safety analysis portion of the Integrated Safety Analysis (ISA) are those events which result in unanticipated moderation, loss of neutron absorber, geometric changes in the critical system, or administrative errors in waste form placement (loading) of the disposal container. The specific events to be considered must be based on the review of the system's design, as discussed in Section 3.2. A transition of licensing approach (e.g., deterministic versus risk-informed, performance-based) is not obvious and will require analysis. For commercial spent nuclear fuel, the probability of interspersed moderation may be low enough to allow nearly the same Critical Limit for both preclosure and postclosure, though an administrative margin will be applied to preclosure and possibly not to postclosure. Similarly the Design Basis Events for the waste package may be incredible and therefore not

  8. Preclosure Criticality Analysis Process Report

    International Nuclear Information System (INIS)

    Thomas, D.A.

    1999-01-01

    The design approach for criticality of the disposal container and waste package will be dictated by existing regulatory requirements. This conclusion is based on the fact that preclosure operations and facilities have significant similarities to existing facilities and operations currently regulated by the NRC. The major difference would be the use of a risk-informed approach with burnup credit. This approach could reduce licensing delays and costs of the repository. The probability of success for this proposed seamless licensing strategy is increased, since there is precedence of regulation (10 CFR Part 63 and NUREG 1520) and commercial precedence for allowing burnup credit at sites similar to Yucca Mountain during preclosure. While NUREG 1520 is not directly applicable to a facility for handling spent nuclear fuel, the risk-informed approach to criticality analysis in NUREG 1520 is considered indicative of how the NRC will approach risk-informed criticality analysis at spent fuel facilities in the future. The types of design basis events which must be considered during the criticality safety analysis portion of the Integrated Safety Analysis (ISA) are those events which result in unanticipated moderation, loss of neutron absorber, geometric changes in the critical system, or administrative errors in waste form placement (loading) of the disposal container. The specific events to be considered must be based on the review of the system's design, as discussed in Section 3.2. A transition of licensing approach (e.g., deterministic versus risk-informed, performance-based) is not obvious and will require analysis. For commercial spent nuclear fuel, the probability of interspersed moderation may be low enough to allow nearly the same Critical Limit for both preclosure and postclosure, though an administrative margin will be applied to preclosure and possibly not to postclosure. Similarly the Design Basis Events for the waste package may be incredible and therefore not

  9. PRECLOSURE CRITICALITY ANALYSIS PROCESS REPORT

    International Nuclear Information System (INIS)

    Danise, A.E.

    2004-01-01

    This report describes a process for performing preclosure criticality analyses for a repository at Yucca Mountain, Nevada. These analyses will be performed from the time of receipt of fissile material until permanent closure of the repository (preclosure period). The process describes how criticality safety analyses will be performed for various configurations of waste in or out of waste packages that could occur during preclosure as a result of normal operations or event sequences. The criticality safety analysis considers those event sequences resulting in unanticipated moderation, loss of neutron absorber, geometric changes, or administrative errors in waste form placement (loading) of the waste package. The report proposes a criticality analyses process for preclosure to allow a consistent transition from preclosure to postclosure, thereby possibly reducing potential cost increases and delays in licensing of Yucca Mountain. The proposed approach provides the advantage of using a parallel regulatory framework for evaluation of preclosure and postclosure performance and is consistent with the U.S. Nuclear Regulatory Commission's approach of supporting risk-informed, performance-based regulation for fuel cycle facilities, ''Yucca Mountain Review Plan, Final Report'', and 10 CFR Part 63. The criticality-related criteria for ensuring subcriticality are also described as well as which guidance documents will be utilized. Preclosure operations and facilities have significant similarities to existing facilities and operations currently regulated by the U.S. Nuclear Regulatory Commission; therefore, the design approach for preclosure criticality safety will be dictated by existing regulatory requirements while using a risk-informed approach with burnup credit for in-package operations

  10. Consensus standards utilized and implemented for nuclear criticality safety in Japan

    International Nuclear Information System (INIS)

    Nomura, Yasushi; Okuno, Hiroshi; Naito, Yoshitaka

    1996-01-01

    The fundamental framework for the criticality safety of nuclear fuel facilities regulations is, in many advanced countries, generally formulated so that technical standards or handbook data are utilized to support the licensing safety review and to implement its guidelines. In Japan also, adequacy of the safety design of nuclear fuel facilities is checked and reviewed on the basis of licensing safety review guides. These guides are, first, open-quotes The Basic Guides for Licensing Safety Review of Nuclear Fuel Facilities,close quotes and as its subsidiaries, open-quotes The Uranium Fuel Fabrication Facility Licensing Safety Review Guidesclose quotes and open-quotes The Reprocessing Facility Licensing Safety Review Guides.close quotes The open-quotes Nuclear Criticality Safety Handbook close-quote of Japan and the Technical Data Collection are published and utilized to supply related data and information for the licensing safety review, such as for the Rokkasho reprocessing plant. The well-established technical standards and data abroad such as those by the American Nuclear Society and the American National Standards Institute are also utilized to complement the standards in Japan. The basic principles of criticality safety control for nuclear fuel facilities in Japan are duly stipulated in the aforementioned basic guides as follows: 1. Guide 10: Criticality control for a single unit; 2. Guide 11: Criticality control for multiple units; 3. Guide 12: Consideration for a criticality accident

  11. PREPP criticality control

    International Nuclear Information System (INIS)

    Cargo, C.H.; Becker, G.K.; Berglund, G.R.; O'Connor, N.J.

    1989-03-01

    There is a general lack of data available with regards to criticality control in the use of incineration for full scale processing of TRU waste. In addition to the new analytical tools, such as PRA, are only now starting to be applied to criticality control. These factors have made the development of the PREPP Criticality Control Plan somewhat unique. It is hoped that the PREPP experiences gained in these efforts will provide new insight which will be useful to future facilities processing waste material containing fissile material. 1 fig

  12. Application of international safeguards to fast critical assembly facilities. FY 1980 summary report

    International Nuclear Information System (INIS)

    1980-12-01

    Nuclear materials inventory-verification techniques for large split-table fast critical assemblies are being studied in this program. Emphasis has been given to techniques that minimize fuel handling in order to reduce facility downtime and radiation exposure to the inventory team. The techniques studied include drawer seals, autoradiography, and spectral index measurements. Two-drawer sealing techniques have been studied, and the relative strengths and weaknesses are pointed out. The rod-type locking mechanism would not disrupt the reactor cooling air flow or interfere with autoradiography but is more expensive to implement. Passive autoradiography was used in a ZPPR inventory to verify to a 93% confidence level that less than 8-kg Pu was missing. The inventory was completed in four days by a five-member team with radiation exposures well within acceptable limits. Two autoradiographic film packages were developed to distinguish HEU from a DU matrix. The 30-mil pack requires an exposure between 4 and 16 hours and fits into most of the drawers. The 40-mil pack requires only a two-hour exposure but fits into less than half the drawers

  13. Insight from a Critical Review on the Safety Analysis of Nuclear Fuel Cycle Facility for Domestic Regulatory System

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Chung, Young Wook; Jeong, Seung Young

    2010-01-01

    Korea has 20 nuclear power plants in operation, and 10,761 ton of spent fuel deposited in plant sites. The capacity of reservoir for spent fuel in plant sites is to begin to be full in 2016. The light water reactors of 16 units generate around 320 ton/year and the heavy water reactors of 4 units around 380 ton/year in Korea. And the electricity generated by nuclear power plants is planned to increase up to 59% share by 2030. Spent fuel classified as high level radioactive waste in law is characterized by high level radiation, high heat generation, and high radiological toxicity. In the contrary, it is also a very useful domestic energy source. Thus, the safe management of spent fuel is very important confronting job in nuclear industry. Advanced fuel cycle (AFC) using pyro-process is an innovative technology, by which environmental load is drastically relieved because the extracted long-lived fission products are burn in fast breeder reactors. Domestic nuclear industry also has a perspective road map for the construction of AFC facilities. However, there is not a sufficiently detailed licensing regulatory system yet. Moreover, there is no systematic frame for the safety evaluation. This paper reviews the safety analysis system of foreign fuel cycle facilities. Critical review leads to the insight for setting-up safety analysis system of domestic AFC facilities

  14. Utilization of AHWR critical facility for research and development work on large sample NAA

    International Nuclear Information System (INIS)

    Acharya, R.; Dasari, K.B.; Pujari, P.K.; Swain, K.K.; Reddy, A.V.R.; Verma, S.K.; De, S.K.

    2014-01-01

    The graphite reflector position of AHWR critical facility (CF) was utilized for analysis of large size (g-kg scale) samples using internal mono standard neutron activation analysis (IM-NAA). The reactor position was characterized by cadmium ratio method using In monitor for total flux and sub cadmium to epithermal flux ratio (f). Large sample neutron activation analysis (LSNAA) work was carried out for samples of stainless steel, ancient and new clay potteries and dross. Large as well as non-standard geometry samples (1 g - 0.5 kg) were irradiated. Radioactive assay was carried out using high resolution gamma ray spectrometry. Concentration ratios obtained by IM-NAA were used for provenance study of 30 clay potteries, obtained from excavated Buddhist sites of AP, India. Concentrations of Au and Ag were determined in not so homogeneous three large size samples of dross. An X-Z rotary scanning unit has been installed for counting large and not so homogeneous samples. (author)

  15. Critical Point Facility (CPE) Group in the Spacelab Payload Operations Control Center (SL POCC)

    Science.gov (United States)

    1992-01-01

    The primary payload for Space Shuttle Mission STS-42, launched January 22, 1992, was the International Microgravity Laboratory-1 (IML-1), a pressurized manned Spacelab module. The goal of IML-1 was to explore in depth the complex effects of weightlessness of living organisms and materials processing. Around-the-clock research was performed on the human nervous system's adaptation to low gravity and effects of microgravity on other life forms such as shrimp eggs, lentil seedlings, fruit fly eggs, and bacteria. Materials processing experiments were also conducted, including crystal growth from a variety of substances such as enzymes, mercury iodide, and a virus. The Huntsville Operations Support Center (HOSC) Spacelab Payload Operations Control Center (SL POCC) at the Marshall Space Flight Center (MSFC) was the air/ground communication channel used between the astronauts and ground control teams during the Spacelab missions. Featured is the Critical Point Facility (CPE) group in the SL POCC during STS-42, IML-1 mission.

  16. Nuclear criticality safety training: guidelines for DOE contractors

    International Nuclear Information System (INIS)

    Crowell, M.R.

    1983-09-01

    The DOE Order 5480.1A, Chapter V, Safety of Nuclear Facilities, establishes safety procedures and requirements for DOE nuclear facilities. This guide has been developed as an aid to implementing the Chapter V requirements pertaining to nuclear criticality safety training. The guide outlines relevant conceptual knowledge and demonstrated good practices in job performance. It addresses training program operations requirements in the areas of employee evaluations, employee training records, training program evaluations, and training program records. It also suggests appropriate feedback mechanisms for criticality safety training program improvement. The emphasis is on academic rather than hands-on training. This allows a decoupling of these guidelines from specific facilities. It would be unrealistic to dictate a universal program of training because of the wide variation of operations, levels of experience, and work environments among DOE contractors and facilities. Hence, these guidelines do not address the actual implementation of a nuclear criticality safety training program, but rather they outline the general characteristics that should be included

  17. Results of 16 years' experiments at the critical facility of Valduc

    International Nuclear Information System (INIS)

    Houelle, M.; Mangin, D.; Maubert, L.

    After briefly recalling the sub-critical approach procedure, the fields of experimental studies on criticality explored at the Valduc Criticality Station since 1963 are listed. This was the year in which the ''appareillage B'' went into service as the first installation of sub-critical experiments of the Section [fr

  18. SPI Project Criticality Task Force initial review and assessment

    International Nuclear Information System (INIS)

    McKinley, K.B.; Cannon, J.W.; Marsden, R.S.; Worle, H.A.

    1980-03-01

    The Slagging Pyrolysis Incinerator (SPI) Facility is being developed to process transuranic waste stored and buried at the Idaho National Engineering Laboratory (INEL) into a chemically inert, physically stable, basalt-like residue acceptable for a Federal Repository. A task force was established by the SPI Project Division to review and assess all aspects of criticality safety for the SPI Facility. This document presents the initial review, evaluations, and recommendations of the task force and includes the following: background information on waste characterization, and criticality control approaches and philosophies, a description of the SPI Facility Waste Processing Building, a review and assessment of potentially relevant codes and regulations; a review and assessment of the present state of criticality and assaying/monitoring studies, and recommendations for changes in and additions to these studies. The review and assessment of potentially relevant codes and regulations indicate that ERDAM 0530, Nuclear Criticality Safety should be the controlling document for criticality safety for the SPI Project. In general, the criticality control approaches and philosophies for the SPI Project comply with this document

  19. Nuclear Criticality Experimental Research Center (NCERC) Overview

    Energy Technology Data Exchange (ETDEWEB)

    Goda, Joetta Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Grove, Travis Justin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hayes, David Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, William L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Sanchez, Rene Gerardo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-03-03

    The mission of the National Criticality Experiments Research Center (NCERC) at the Device Assembly Facility (DAF) is to conduct experiments and training with critical assemblies and fissionable material at or near criticality in order to explore reactivity phenomena, and to operate the assemblies in the regions from subcritical through delayed critical. One critical assembly, Godiva-IV, is designed to operate above prompt critical. The Nuclear Criticality Experimental Research Center (NCERC) is our nation’s only general-purpose critical experiments facility and is only one of a few that remain operational throughout the world. This presentation discusses the history of NCERC, the general activities that makeup work at NCERC, and the various government programs and missions that NCERC supports. Recent activities at NCERC will be reviewed, with a focus on demonstrating how NCERC meets national security mission goals using engineering fundamentals. In particular, there will be a focus on engineering theory and design and applications of engineering fundamentals at NCERC. NCERC activities that relate to engineering education will also be examined.

  20. Approaches to nontraditional delivery of nuclear engineering education

    International Nuclear Information System (INIS)

    Malaviya, B.K.

    1991-01-01

    At Rensselaer Polytechnic Institute, the faculty of the nuclear engineering and engineering physics department have, over the years, been involved in a variety of such approaches in response to the changing needs of nuclear industry personnel. A number of different types of short course and workshop programs have been developed and implemented both on and off campus in such areas as basic nuclear technology, reactor design computer codes and applications, nuclear power plant design and maintenance, reactor operations, health physics, modern developments in boiling heat transfer and two-phase flow, and probabilistic risk assessment. Customized coursed tailored to meet the particular needs of personnel in specialized areas can also be offered on specific industrial site locations, generally resulting in substantial savings of time as well as costs associated with tuition, travel, lodging. The Rensselaer Satellite Video Program (RSVP) brings the latest technological aids to the nontraditional delivery of courses and provides the facilities and opportunities for off-campus students and professional personnel to participate in regular academic programs and courses without leaving their industrial sites

  1. The Department of Energy nuclear criticality safety program

    International Nuclear Information System (INIS)

    Felty, J.R.

    2004-01-01

    This paper broadly covers key events and activities from which the Department of Energy Nuclear Criticality Safety Program (NCSP) evolved. The NCSP maintains fundamental infrastructure that supports operational criticality safety programs. This infrastructure includes continued development and maintenance of key calculational tools, differential and integral data measurements, benchmark compilation, development of training resources, hands-on training, and web-based systems to enhance information preservation and dissemination. The NCSP was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 97-2, Criticality Safety, and evolved from a predecessor program, the Nuclear Criticality Predictability Program, that was initiated in response to Defense Nuclear Facilities Safety Board Recommendation 93-2, The Need for Critical Experiment Capability. This paper also discusses the role Dr. Sol Pearlstein played in helping the Department of Energy lay the foundation for a robust and enduring criticality safety infrastructure.

  2. An updated nuclear criticality slide rule

    International Nuclear Information System (INIS)

    Hopper, C.M.; Broadhead, B.L.

    1998-04-01

    This Volume 2 contains the functional version of the updated nuclear criticality slide rule (more accurately, sliding graphs) that is referenced in An Updated Nuclear Criticality Slide Rule: Technical Basis, NUREG/CR-6504, Vol. 1 (ORNL/TM-13322/V1). This functional slide rule provides a readily usable open-quotes in-handclose quotes method for estimating pertinent nuclear criticality accident information from sliding graphs, thereby permitting (1) the rapid estimation of pertinent criticality accident information without laborious or sophisticated calculations in a nuclear criticality emergency situation, (2) the appraisal of potential fission yields and external personnel radiation exposures for facility safety analyses, and (3) a technical basis for emergency preparedness and training programs at nonreactor nuclear facilities. The slide rule permits the estimation of neutron and gamma dose rates and integrated doses based upon estimated fission yields, distance from the fission source, and time-after criticality accidents for five different critical systems. Another sliding graph permits the estimation of critical solution fission yields based upon fissile material concentration, critical vessel geometry, and solution addition rate. Another graph provides neutron and gamma dose-reduction factors for water, steel, and concrete. Graphs from historic documents are provided as references for estimating critical parameters of various fissile material systems. Conversion factors for various English and metric units are provided for quick reference

  3. Technical guide to criticality alarm system design

    International Nuclear Information System (INIS)

    Greenfield, B.

    2009-01-01

    An instructional manual was created to guide criticality safety engineers through the technical aspects of designing a criticality alarm system (CAS) for Dept. of Energy (DOE) hazard class 1 and 2 facilities. The manual was structured such that it can be used by engineers designing completely new systems and by those who are working with existing facilities. Major design tasks are thoroughly analyzed to provide concise direction for how to complete the analysis. Regulatory and technical performance requirements were both addressed. (authors)

  4. Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

    International Nuclear Information System (INIS)

    Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

    2011-01-01

    Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

  5. Expected performance and benefits of an advanced containment and surveillance system at the fast critical facility fca of jaeri

    International Nuclear Information System (INIS)

    Mukaiyama, T.; Kuroi, H.

    1981-01-01

    This paper reports on the development and test of a personnel portal monitor for IAEA safeguards at the fast critical facility FCA. The main components of the portal are 1) the walk-through metal detector, 2) the visual surveillance system with CCTV, 3) the tamper indication system using multi sensors and 4) the remote monitoring capability through RECOVER system. The metal detector developed can detect a single coupon of metallic nuclear fuel plate (2 in. *2 in. *1/16 in.) regardless of the orientation of a fuel plate relative to the electromagnetic field generated in the metal detector. 3 refs

  6. Subcriticality calculations for the FFTF reverse approach to critical experiment

    International Nuclear Information System (INIS)

    Selby, D.L.; Flanagan, G.F.

    1975-01-01

    The reverse approach to critical (RAC) experiments were performed in the ZPR-IX critical facility at Argonne National Laboratory. One of the major objectives of this project is to determine the adequacy of the low-level flux monitor (LLFM) detectors for initial loading of the Fast Flux Test Facility (FFTF). 5 references

  7. Radioactive material inventory control at a waste characterization facility

    International Nuclear Information System (INIS)

    Yong, L.K.; Chapman, J.A.; Schultz, F.J.

    1996-01-01

    Due to the recent introduction of more stringent Department of Energy (DOE) regulations and requirements pertaining to nuclear and criticality safety, the control of radioactive material inventory has emerged as an important facet of operations at DOE nuclear facilities. In order to comply with nuclear safety regulations and nuclear criticality requirements, radioactive material inventories at each nuclear facility have to be maintained below limits specified for the facility in its safety authorization basis documentation. Exceeding these radioactive material limits constitutes a breach of the facility's nuclear and criticality safety envelope and could potentially result in an accident, cause a shut-down of the facility, and bring about imminent regulatory repercussions. The practice of maintaining control of radioactive material, especially sealed and unsealed sources, is commonplace and widely implemented; however, the requirement to track the entire radioactivity inventory at each nuclear facility for the purpose of ensuring nuclear safety is a new development. To meet the new requirements, the Applied Radiation Measurements Department at Oak Ridge National Laboratory (ORNL) has developed an information system, called the open-quotes Radioactive Material Inventory Systemclose quotes (RMIS), to track the radioactive material inventory at an ORNL facility, the Waste Examination and Assay Facility (WEAF). The operations at WEAF, which revolve around the nondestructive assay and nondestructive examination of waste and related research and development activities, results in an ever-changing radioactive material inventory. Waste packages and radioactive sources are constantly being brought in or taken out of the facility; hence, use of the RMIS is necessary to ensure that the radioactive material inventory limits are not exceeded

  8. Analysis of RA-8 critical facility core in some configurations; Analisis del nucleo de la facilidad critica RA-8 en distintas configuraciones

    Energy Technology Data Exchange (ETDEWEB)

    Abbate, Maximo J [Instituto de Investigaciones Cientificas y Tecnicas de las Fuerzas Armadas (CITEFA), Villa Martelli (Argentina); Sbaffoni, Maria M [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Gerencia de Tecnologia

    2000-07-01

    The RA-8 critical facility was designated and built to be used in the experimental plan of the 'CAREM' Project but is, in itself, very versatile and adequate to perform many types of other experiments. The present paper includes calculated estimates of some critical configurations and comparisons with experimental results obtained during its start up. Results for Core 1 with homogeneous arrangement of rods containing 1.8 % enriched uranium, showed very good agreement. In fact, an experimentally critical configuration was reached with 1.300 rods and calculated values were: 1.310 using the WIMS code and 1.148 from the CONDOR code. Moreover, it was verified that the estimated number of 3.4% enriched uranium rods to be fabricated is enough to build a heterogeneous core or even a homogeneous core with this enrichment. The replacement of 3.4 % enriched uranium by 3.6 % will not present problems related with the original plan. (author)

  9. Nuclear criticality safety aspects of emergency response at the Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Baker, J.S.

    2003-01-01

    Emergency response at Los Alamos National Laboratory (LANL) is handled through a graded approach depending on the specific emergency situation . LANL maintains a comprehensive capability to respond to events ranging from minor facility events (alerts) through major community events (general emergencies), including criticality accidents . Criticality safety and emergency response apply to all activities involving significant quantities of fissile material at LANL, primarily at Technical Area 18 (TA-18, the Los Alamos Critical Experiments Facility) and Technical Area 55 (TA-55, the Plutonium Facility). This discussion focuses on response to a criticality accident at TA-55; the approach at TA-18 is comparable .

  10. Investigation for Low-Cost Permanent Magnets.

    Science.gov (United States)

    1981-09-01

    Code RRM I | -. .. _ _ _ - -,’ .-- -- nnwi-- l aI l l ai Professor G.A. Ansell Professor N.S. Stoloff Rensselaer Polytechnic Institute Rensselaer...Polytechnic Institute Department of Metallurgical Engineering School of Engineering Troy, NY 02181 Troy, NY 12181 Dr. C. Adam Professor A. Lawley Pratt

  11. Request from nuclear fuel cycle and criticality safety design

    International Nuclear Information System (INIS)

    Hamasaki, Manabu; Sakashita, Kiichiro; Natsume, Toshihiro

    2005-01-01

    The quality and reliability of criticality safety design of nuclear fuel cycle systems such as fuel fabrication facilities, fuel reprocessing facilities, storage systems of various forms of nuclear materials or transportation casks have been largely dependent on the quality of criticality safety analyses using qualified criticality calculation code systems and reliable nuclear data sets. In this report, we summarize the characteristics of the nuclear fuel cycle systems and the perspective of the requirements for the nuclear data, with brief comments on the recent issue about spent fuel disposal. (author)

  12. Criticality Safety Evaluation for Small Sample Preparation and Non-Destructive Assay (NDA) Operations in Wing 7 Basement of the CMR Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kunkle, Paige Elizabeth [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zhang, Ning [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-08-02

    Nuclear Criticality Safety (NCS) has reviewed the fissionable material small sample preparation and NDA operations in Wing 7 Basement of the CMR Facility. This is a Level-1 evaluation conducted in accordance with NCS-AP-004 [Reference 1], formerly NCS-GUIDE-01, and the guidance set forth on use of the Standard Criticality Safety Requirements (SCSRs) [Reference 2]. As stated in Reference 2, the criticality safety evaluation consists of both the SCSR CSED and the SCSR Application CSED. The SCSR CSED is a Level-3 CSED [Reference 3]. This Level-1 CSED is the SCSR Application CSED. This SCSR Application (Level-1) evaluation does not derive controls, it simply applies controls derived from the SCSR CSED (Level-3) for the application of operations conducted here. The controls derived in the SCSR CSED (Level-3) were evaluated via the process described in Section 6.6.5 of SD-130 (also reproduced in Section 4.3.5 of NCS-AP-004 [Reference 1]) and were determined to not meet the requirements for consideration of elevation into the safety basis documentation for CMR. According to the guidance set forth on use of the SCSRs [Reference 2], the SCSR CSED (Level-3) is also applicable to the CMR Facility because the process and the normal and credible abnormal conditions in question are bounded by those that are described in the SCSR CSED. The controls derived in the SCSR CSED include allowances for solid materials and solution operations. Based on the operations conducted at this location, there are less-than-accountable (LTA) amounts of 233U. Based on the evaluation documented herein, the normal and credible abnormal conditions that might arise during the execution of this process will remain subcritical with the following recommended controls.

  13. Criticality alarm device

    International Nuclear Information System (INIS)

    Kasai, Kenji.

    1994-01-01

    The device of the present invention is utilized, for example, to a reprocessing facility for storing and processing nuclear fuels and measures and controls the nuclear fuel assembly system so as not to exceed criticality. That is, a conventional criticality alarm device applies a predetermined processing to neutron fluxes generated from a nuclear fuel assembly system containing nuclear fuels and outputs an alarm. The device of the present invention comprises (1) a neutron flux supply source for increasing and decreasing neutron fluxes periodically and supplying them to nuclear fuel assemblies, (2) a detector for detecting neutron fluxes in the nuclear fuel assemblies, (3) a critical state judging section for judging the critical state of the nuclear fuel assemblies based on the periodically changing signals obtained from the detector (2) and (4) an alarm section for outputting criticality alarms depending on the result of the judgement. The device of the present invention can accurately recognize the critical state of the nuclear fuel assembly system and can forecast reaching of the nuclear fuel assembly to criticality or prompt neutron critical state. (I.S.)

  14. Implications of system usability on intermodal facility design.

    Science.gov (United States)

    2010-08-01

    Ensuring good design of intermodal transportation facilities is critical for effective and : satisfactory operation. Passenger use of the facilities is often hindered by inadequate space, a poor : layout, or lack of signage. This project aims to impr...

  15. Identifying critical success factors (CSFs) of Facilities Management (FM) in non-low cost high-rise residential buildings

    Science.gov (United States)

    Dahlan, F. M.; Zainuddin, A.

    2018-02-01

    Critical success factors (CSFs) are important key areas of activity that must be performed well in any Facilities Management (FM) organisation to achieve its missions, objectives or goals. Before implementing CSFs, an FM organisation must identify the key areas where things must be done properly to enable the business to flourish. Although many performance measurements in FM organisation have been discussed in previous research, not much research has been done on CSFs from the perspective of FM business in non-low cost high-rise residential buildings. The purpose of this study is to develop a methodology in developing the CSFs group and CSFs for FM organisation in non-low cost residential buildings. This research will involve three (3) phases of research strategy to achieve the objective of this research.

  16. Transportable criticality alarm system

    International Nuclear Information System (INIS)

    Clem, W.E.

    1988-09-01

    The Transportable Criticality Alarm System was developed at the Hanford Site in 1982 to comply with the requirements of US Department of Energy Order DOE 5480.1, 12/18/80, and ANSI/ANS-8.3- 1979. The portable unit that it replaced failed to comply with the new requirements in that it did not provide the necessary warning of malfunctions, nor did it provide the Hanford Site standard criticality alarm signal. Modern technology allowed the Transportable Criticality Alarm System to comply with the criticality requirements cited and to incorporate other features that make it more usable, maintainable, and reliable. The Transportable Criticality Alarm System (TCAS) provides temporary criticality coverage in manned areas where the facility criticality alarm system is not operable. This gamma radiation-sensitive system has been in use for the past 6 yr at the Hanford Site. 2 refs., 4 figs., 1 tab

  17. Criticality classification of waste receiving and processing module 2A

    International Nuclear Information System (INIS)

    Boothe, G.F.

    1994-10-01

    The purpose of this document is to evaluate the criticality potential of the Waste Receiving and Processing Module 2A (WRAP 2A) and to demonstrate that the facility is an exempt facility, under the provisions of the Nuclear Criticality Safety Manual. The WRAP 2A maximum potential transuranic (TRU) contents of feedstreams and product inventories are discussed. Total plant fissionable materials are estimated and compared with the fissionable material exempt quantity. The WRAP 2A operations and processes are also described, relative to the potential for concentrating or accumulating fissionable material within the facility

  18. Alternate measurements of benefit to criticality issues at Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Toffer, H.; Cabrilla, D.E.

    1997-06-01

    Measurements in a critical mass facility, such as the Los Alamos Critical Experiments Facility, could be performed on simulated nuclear waste materials that would provide important critical mass information and concurrently provide a calibration of alternate measurement techniques. In addition to criticality information, the measurements could also provide a better assessment of diluent material/neutron interaction cross sections. An Hanford, large quantities of fissionable materials are dispersed in tanks, fuel storage pools, and in solid waste. Although the fissionable materials are well diluted by a variety of neutron-moderating and -absorbing substances, it is difficult to assess the margin of subcriticality. A number of measurement approaches are proposed that will either help determine fissionable material concentrations, distributions, or provide a direct measure of subcriticality. The methods under consideration involve passive neutron counting, active neutron measurements, pulse neutron applications, neutron noise analyses, and cover gas evaluations. Active neutron measurements can also provide insight into the determination of neutron absorber concentrations. Efforts are underway to test some of the methods in actual waste tank environments and geometries. It is important that these methods be tested and calibrated in a critical mass facility.

  19. Alternate measurements of benefit to criticality issues at Hanford

    International Nuclear Information System (INIS)

    Toffer, H.; Cabrilla, D.E.

    1997-01-01

    Measurements in a critical mass facility, such as the Los Alamos Critical Experiments Facility, could be performed on simulated nuclear waste materials that would provide important critical mass information and concurrently provide a calibration of alternate measurement techniques. In addition to criticality information, the measurements could also provide a better assessment of diluent material/neutron interaction cross sections. An Hanford, large quantities of fissionable materials are dispersed in tanks, fuel storage pools, and in solid waste. Although the fissionable materials are well diluted by a variety of neutron-moderating and -absorbing substances, it is difficult to assess the margin of subcriticality. A number of measurement approaches are proposed that will either help determine fissionable material concentrations, distributions, or provide a direct measure of subcriticality. The methods under consideration involve passive neutron counting, active neutron measurements, pulse neutron applications, neutron noise analyses, and cover gas evaluations. Active neutron measurements can also provide insight into the determination of neutron absorber concentrations. Efforts are underway to test some of the methods in actual waste tank environments and geometries. It is important that these methods be tested and calibrated in a critical mass facility

  20. Evolvement of nuclear criticality safety programs

    International Nuclear Information System (INIS)

    Ketzlach, N.

    1992-01-01

    Nuclear criticality safety (NCS) has developed from a discipline requiring the services of personnel with only a background in reactor physics to that involving reactor physics, process engineering, and design as well as administration of the program to ensure all its requirements are implemented. When Oak Ridge National Laboratory (ORNL) was designed and constructed, the physicists at Los Alamos National Laboratory (LANL) were performing the criticality analyses. A physicist who had no chemical process or engineering experience was brought in from LANL to determine whether the facility would be safe. It was only because of his understanding of the reactor physics principles, scientific intuition, and some luck that the design and construction of the facility led to a safe plant. It took a number of years of experience with facility operations and the dedication of personnel for NCS to reach its present status as a recognized discipline

  1. A landscape of sport facilities in the Netherlands

    NARCIS (Netherlands)

    Hoekman, R.H.A.; Kraaykamp, G.L.M.; Breedveld, K.

    2016-01-01

    This article explores the geographical distribution of sport facilities in the Netherlands and evaluates the supply and proximity of such facilities in relation to area level deprivation and urbanity. In today?s context of financial austerity, many municipalities must critically assess their

  2. Criticality assessment of LLRWDF closure

    International Nuclear Information System (INIS)

    Sarrack, A.G.; Weber, J.H.; Woody, N.D.

    1992-01-01

    During the operation of the Low Level Radioactive Waste Disposal Facility (LLRWDF), large amounts (greater than 100 kg) of enriched uranium (EU) were buried. This EU came primarily from the closing and decontamination of the Naval Fuels Facility in the time period from 1987--1989. Waste Management Operations (WMO) procedures were used to keep the EU boxes separated to prevent possible criticality during normal operation. Closure of the LLRWDF is currently being planned, and waste stabilization by Dynamic Compaction (DC) is proposed. Dynamic compaction will crush the containers in the LLRWDF and result in changes in their geometry. Research of the LLRWDF operations and record keeping practices have shown that the EU contents of trenches are known, but details of the arrangement of the contents cannot be proven. Reviews of the trench contents, combined with analysis of potential critical configurations, revealed that some portions of the LLRWDF can be expected to be free of criticality concerns while other sections have credible probabilities for the assembly of a critical mass, even in the uncompacted configuration. This will have an impact on the closure options and which trenches can be compacted

  3. Next generation storage facility

    International Nuclear Information System (INIS)

    Schlesser, J.A.

    1994-01-01

    With diminishing requirements for plutonium, a substantial quantity of this material requires special handling and ultimately, long-term storage. To meet this objective, we at Los Alamos, have been involved in the design of a storage facility with the goal of providing storage capabilities for this and other nuclear materials. This paper presents preliminary basic design data, not for the structure and physical plant, but for the container and arrays which might be configured within the facility, with strong emphasis on criticality safety features

  4. Engineering study for closure of 209E facility

    Energy Technology Data Exchange (ETDEWEB)

    Brevick, C.H.; Heys, W.H.; Johnson, E.D.

    1997-07-07

    This document is an engineering study for evaluating alternatives to determine the most cost effective closure plan for the 209E Facility, Critical Mass Laboratory. This laboratory is located in the 200 East Area of the Hanford Site and contains a Critical Assembly Room and a Mix room were criticality experiments were once performed.

  5. Engineering study for closure of 209E facility

    International Nuclear Information System (INIS)

    Brevick, C.H.; Heys, W.H.; Johnson, E.D.

    1997-01-01

    This document is an engineering study for evaluating alternatives to determine the most cost effective closure plan for the 209E Facility, Critical Mass Laboratory. This laboratory is located in the 200 East Area of the Hanford Site and contains a Critical Assembly Room and a Mix room were criticality experiments were once performed

  6. 78 FR 36242 - Notice of Intent To Repatriate Cultural Items: New York State Museum, Albany, NY

    Science.gov (United States)

    2013-06-17

    ..., Rensselaer County, NY, by Reverend O.C. Auringer of Troy, NY. Museum records indicate that the cultural items..., and Saratoga Counties, NY, by Mr. Dwinel F. Thompson of Troy, NY. Museum records indicate that the... collections. From the former Laureate Grounds in Troy, Rensselaer County, NY, the 90 unassociated funerary...

  7. Experimental assessment of the performance of a proposed lead slowing-down spectrometer at WNR/PSR [Weapons Neutron Research/Proton Storage Ring

    International Nuclear Information System (INIS)

    Moore, M.S.; Koehler, P.E.; Michaudon, A.; Schelberg, A.; Danon, Y.; Block, R.C.; Slovacek, R.E.; Hoff, R.W.; Lougheed, R.W.

    1990-01-01

    In November 1989, we carried out a measurement of the fission cross section of 247 Cm, 250 Cf, and 254 Es on the Rensselaer Intense Neutron Source (RINS) at Rensselaer Polytechnic Institute (RPI). In July 1990, we carried out a second measurement, using the same fission chamber and electronics, in beam geometry at the Los Alamos Neutron Scattering Center (LANSCE) facility. Using the relative count rates observed in the two experiments, and the flux-enhancement factors determined by the RPI group for a lead slowing-down spectrometer compared to beam geometry, we can assess the performance of a spectrometer similar to RINS, driven by the Proton Storage Ring (PSR) at the Los Alamos National Laboratory. With such a spectrometer, we find that is is feasible to make measurements with samples of 1 ng for fission 1 μg for capture, and of isotopes with half-lives of tens of minutes. It is important to note that, while a significant amount of information can be obtained from the low resolution RINS measurement, a definitive determination of average properties, including the level density, requires that the resonance structure be resolved. 12 refs., 5 figs., 3 tabs

  8. Experimental assessment of the performance of a proposed lead slowing-down spectrometer at WNR/PSR (Weapons Neutron Research/Proton Storage Ring)

    Energy Technology Data Exchange (ETDEWEB)

    Moore, M.S.; Koehler, P.E.; Michaudon, A.; Schelberg, A. (Los Alamos National Lab., NM (USA)); Danon, Y.; Block, R.C.; Slovacek, R.E. (Rensselaer Polytechnic Inst., Troy, NY (USA)); Hoff, R.W.; Lougheed, R.W. (Lawrence Livermore National Lab., CA (USA))

    1990-01-01

    In November 1989, we carried out a measurement of the fission cross section of {sup 247}Cm, {sup 250}Cf, and {sup 254}Es on the Rensselaer Intense Neutron Source (RINS) at Rensselaer Polytechnic Institute (RPI). In July 1990, we carried out a second measurement, using the same fission chamber and electronics, in beam geometry at the Los Alamos Neutron Scattering Center (LANSCE) facility. Using the relative count rates observed in the two experiments, and the flux-enhancement factors determined by the RPI group for a lead slowing-down spectrometer compared to beam geometry, we can assess the performance of a spectrometer similar to RINS, driven by the Proton Storage Ring (PSR) at the Los Alamos National Laboratory. With such a spectrometer, we find that is is feasible to make measurements with samples of 1 ng for fission 1 {mu}g for capture, and of isotopes with half-lives of tens of minutes. It is important to note that, while a significant amount of information can be obtained from the low resolution RINS measurement, a definitive determination of average properties, including the level density, requires that the resonance structure be resolved. 12 refs., 5 figs., 3 tabs.

  9. Criticality accident in Argentina

    International Nuclear Information System (INIS)

    Oliveira, A.R. de.

    1984-01-01

    A recent criticality type accident, ocurred in Argetina, is commented. Considerations about the nature of the facility where this accident took place, its genesis, type of operation carried out on the day of the event, and the medical aspects involved are done. (Author) [pt

  10. Tank farms criticality safety manual

    International Nuclear Information System (INIS)

    FORT, L.A.

    2003-01-01

    This document defines the Tank Farms Contractor (TFC) criticality safety program, as required by Title 10 Code of Federal Regulations (CFR-), Subpart 830.204(b)(6), ''Documented Safety Analysis'' (10 CFR- 830.204 (b)(6)), and US Department of Energy (DOE) 0 420.1A, Facility Safety, Section 4.3, ''Criticality Safety.'' In addition, this document contains certain best management practices, adopted by TFC management based on successful Hanford Site facility practices. Requirements in this manual are based on the contractor requirements document (CRD) found in Attachment 2 of DOE 0 420.1A, Section 4.3, ''Nuclear Criticality Safety,'' and the cited revisions of applicable standards published jointly by the American National Standards Institute (ANSI) and the American Nuclear Society (ANS) as listed in Appendix A. As an informational device, requirements directly imposed by the CRD or ANSI/ANS Standards are shown in boldface. Requirements developed as best management practices through experience and maintained consistent with Hanford Site practice are shown in italics. Recommendations and explanatory material are provided in plain type

  11. Estimates of fission yields in nuclear criticality excursions

    International Nuclear Information System (INIS)

    Choi, J.S.; Thompson, J.W.; Reed, R.

    1995-06-01

    There is a need for computer simulation of hypothetical criticality excursions involving significant quantities of fissionable materials, especially in fissile aqueous system. The need arises due to the requirements for the emergency planning of facilities where the fissionable materials are handled, processed, or stored; and the regulatory requirements associated with facility operation or conversion. It is proposed here that a data base of fission yeilds for critical experiments and known accidents (both aqueous and solid) should be generated by using existing or new computer codes. The success in compiling this data base would provide useful source-terms for criticality excursions, realistic estimates of emergency-response boundary, as well as a replacement for the ''rule-of-thumb'' or ''bounding'' method. 10 refs

  12. Critical experiments of JMTRC MEU cores

    International Nuclear Information System (INIS)

    Nagaoka, Y.; Takeda, K.; Shimakawa, S.; Koike, S.; Oyamada, R.

    1984-01-01

    The JMTRC, the critical facility of the Japan Materials Testing Reactor (JMTR), went critical on August 29, 1983, with 14 medium enriched uranium (MEU, 45%) fuel elements. Experiments are now being carried out to measure the change in various reactor characteristics between the previous HEU core and the new MEU fueled core. This paper describes the results obtained thus far on critical mass, excess reactivity, control rod worths and flux distribution, including preliminary neutronics calculations for the experiments using the SRAC code. (author)

  13. Guarding America: Security Guards and U.S. Critical Infrastructure Protection

    National Research Council Canada - National Science Library

    Parfomak, Paul W

    2004-01-01

    The Bush Administration's 2003 National Strategy for the Physical Protection of Critical Infrastructures and Key Assets indicates that security guards are an important source of protection for critical facilities...

  14. SURF II: Characteristics, facilities, and plans

    International Nuclear Information System (INIS)

    Madden, R.P.; Canfield, R.; Furst, M.; Hamilton, A.; Hughey, L.

    1992-01-01

    This facility report describes the Synchrotron Ultraviolet Radiation Facility (SURF II) operated by the National Institute of Standards and Technology, Gaithersburg, Maryland. SURF II is a 300-MeV electron storage ring which provides well characterized continuum radiation from the far infrared to the soft x-ray region with the critical wavelength at 17.4 nm. Brief descriptions are given of the user facilities, the characteristics of the synchrotron radiation, the main storage ring, the injector system and each of the operating beam lines, and associated instruments. Further description is given of expansion plans for additional beam lines

  15. Design of plutonium processing facilities

    International Nuclear Information System (INIS)

    Derbyshire, W.; Sills, R.J.

    1982-01-01

    Five considerations for the design of plutonium processing facilities are identified. These are: Toxicity, Radiation, Criticality, Containment and Remote Operation. They are examined with reference to reprocessing spent nuclear fuel and application is detailed both for liquid and dry processes. (author)

  16. Review of the CRAC and SILENE Criticality Accident Studies

    International Nuclear Information System (INIS)

    Barbry, F.; Fouillaud, P.; Grivot, P.; Reverdy, L.

    2009-01-01

    In 1967, the Commissariat et l'Energie Atomique (French Atomic Energy Agency) performed its first research on criticality accidents for the purpose of limiting their impact on people, the environment, and nuclear facilities themselves. A criticality accident is accompanied by intense neutron and gamma emissions and release of radioactive fission products-gases and aerosols-gene rating risk of irradiation and contamination. This work has supplemented earlier work in criticality safety, which concentrated on critical mass measurements and computations. Understanding of the consequences of criticality accidents was limited. Emergency planning was hampered by lack of data. Information became available from pulsed reactor experiments, but the experiments were restricted to the established reactor configurations. The objectives of research performed at the Valduc criticality laboratory, mainly on aqueous fissile media, using the CRAC and SILENE facilities, by multidisciplinary teams of physicists, dosimetry specialists, and radio-biologists, were to model criticality accident physics, estimate irradiation risks and radioactive releases, detect excursions, and organize emergency response. The results of the Valduc experiments have contributed toward improved understanding of criticality accident phenomenology and better evaluation of the risks associated with such accidents. (authors)

  17. Review of the CRAC and SILENE Criticality Accident Studies

    Energy Technology Data Exchange (ETDEWEB)

    Barbry, F.; Fouillaud, P.; Grivot, P.; Reverdy, L. [CEA Valduc, Serv Rech Neutron and Critcite, 21 - Is-sur-Tille (France)

    2009-02-15

    In 1967, the Commissariat et l'Energie Atomique (French Atomic Energy Agency) performed its first research on criticality accidents for the purpose of limiting their impact on people, the environment, and nuclear facilities themselves. A criticality accident is accompanied by intense neutron and gamma emissions and release of radioactive fission products-gases and aerosols-gene rating risk of irradiation and contamination. This work has supplemented earlier work in criticality safety, which concentrated on critical mass measurements and computations. Understanding of the consequences of criticality accidents was limited. Emergency planning was hampered by lack of data. Information became available from pulsed reactor experiments, but the experiments were restricted to the established reactor configurations. The objectives of research performed at the Valduc criticality laboratory, mainly on aqueous fissile media, using the CRAC and SILENE facilities, by multidisciplinary teams of physicists, dosimetry specialists, and radio-biologists, were to model criticality accident physics, estimate irradiation risks and radioactive releases, detect excursions, and organize emergency response. The results of the Valduc experiments have contributed toward improved understanding of criticality accident phenomenology and better evaluation of the risks associated with such accidents. (authors)

  18. Gas-cooled fast reactors. Motivation and presentation of the ENIGMA program in the MASURCA experimental critical facility

    International Nuclear Information System (INIS)

    Tommasi, Jean; Jacqmin, Robert; Mellier, Frederic

    2005-01-01

    This paper describes a new experimental physics program in support of gas cooled fast reactor (GCFR) design studies, called ENIGMA, to be performed in the MASURCA critical facility at CEA-Cadarache, France. The prospective GCFR design studies at CEA are presented, as well as the specific neutronics features needing an extension of the validation of calculation tools and nuclear data. The relevant existing experiments are briefly reviewed and the need for new experimental data is pointed out. The first phase of the proposed new experiments includes a reference core with a representative spectrum, and a series of central core substitutions involving spectrum shifts, streaming studies, low-grade Pu substitutions, innovative material (Si, Zr) substitutions. Reflector substitution zones will include elements foreseen for the reflectors (Si, Zr, C). Subsequent phases will involve larger amounts of low-grade Pu or innovative materials, and configurations representative of experimental and demonstration GCFRs. (author)

  19. Development of High-Level Safety Requirements for a Pyroprocessing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Seok Jun; Jo, Woo Jin; You, Gil Sung; Choung, Won Myung; Lee, Ho Hee; Kim, Hyun Min; Jeon, Hong Rae; Ku, Jeong Hoe; Lee, Hyo Jik [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    Korea Atomic Energy Research Institute (KAERI) has been developing a pyroproceesing technology to reduce the waste volume and recycle some elements. The pyroprocessing includes several treatment processes which are related with not only radiological and physical but also chemical and electrochemical properties. Thus, it is of importance to establish safety design requirements considering all the aspects of those properties for a reliable pyroprocessing facility. In this study, high-level requirements are presented in terms of not only radiation protection, nuclear criticality, fire protection, and seismic safety but also confinement and chemical safety for the unique characteristics of a pyroprocessing facility. Several high-level safety design requirements such as radiation protection, nuclear criticality, fire protection, seismic, confinement, and chemical processing were presented for a pyroprocessing facility. The requirements must fulfill domestic and international safety technology standards for a nuclear facility. Furthermore, additional requirements should be considered for the unique electrochemical treatments in a pyroprocessing facility.

  20. Handbook on criticality. Vol. 1. Criticality and nuclear safety; Handbuch zur Kritikalitaet. Bd. 1. Kritikalitaet und nukleare Sicherheit

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2015-04-15

    This handbook was prepared primarily with the aim to provide information to experts in industry, authorities or research facilities engaged in criticality-safety-related problems that will allow an adequate and rapid assessment of criticality safety issues already in the planning and preparation of nuclear facilities. However, it is not the intention of the authors of the handbook to offer ready solutions to complex problems of nuclear safety. Such questions have to remain subject to an in-depth analysis and assessment to be carried out by dedicated criticality safety experts. Compared with the previous edition dated December 1998, this handbook has been further revised and supplemented. The proven basic structure of the handbook remains unchanged. The handbook follows in some ways similar criticality handbooks or instructions published in the USA, UK, France, Japan and the former Soviet Union. The expedient use of the information given in this handbook requires a fundamental understanding of criticality and the terminology of nuclear safety. In Vol. 1, ''Criticality and Nuclear Safety'', therefore, first the most important terms and fundamentals are introduced and explained. Subsequently, experimental techniques and calculation methods for evaluating criticality problems are presented. The following chapters of Vol. 1 deal i. a. with the effect of neutron reflectors and absorbers, neutron interaction, measuring methods for criticality, and organisational safety measures and provide an overview of criticality-relevant operational experience and of criticality accidents and their potential hazardous impact. Vol. 2 parts 1 and 2 finally compile criticality parameters in graphical and tabular form. The individual graph sheets are provided with an initially explained set of identifiers, to allow the quick finding of the information of current interest. Part 1 includes criticality parameters for systems with {sup 235}U as fissile material, while part

  1. Analysis of Critical Parts and Materials

    Science.gov (United States)

    1980-12-01

    1 1 1% 1% 1% 1% Large Orders Manual Ordering of Some Critical Parts Order Spares with Original Order Incentives Belter Capital Investment...demand 23 Large orders 24 Long lead procurement funding (including raw materials, facility funding) 25 Manpower analysis and training 26 Manual ... ordering of some critical parts 27 More active role in schedule negotiation 28 Multiple source procurements 29 Multi-year program funding 30 Order

  2. Volatile organic compounds at swine facilities: a critical review.

    Science.gov (United States)

    Ni, Ji-Qin; Robarge, Wayne P; Xiao, Changhe; Heber, Albert J

    2012-10-01

    Volatile organic compounds (VOCs) are regulated aerial pollutants that have environmental and health concerns. Swine operations produce and emit a complex mixture of VOCs with a wide range of molecular weights and a variety of physicochemical properties. Significant progress has been made in this area since the first experiment on VOCs at a swine facility in the early 1960s. A total of 47 research institutions in 15 North American, European, and Asian countries contributed to an increasing number of scientific publications. Nearly half of the research papers were published by U.S. institutions. Investigated major VOC sources included air inside swine barns, in headspaces of manure storages and composts, in open atmosphere above swine wastewater, and surrounding swine farms. They also included liquid swine manure and wastewater, and dusts inside and outside swine barns. Most of the sample analyses have been focusing on identification of VOC compounds and their relationship with odors. More than 500 VOCs have been identified. About 60% and 10% of the studies contributed to the quantification of VOC concentrations and emissions, respectively. The largest numbers of VOC compounds with reported concentrations in a single experimental study were 82 in air, 36 in manure, and 34 in dust samples. The relatively abundant VOC compounds that were quantified in at least two independent studies included acetic acid, butanoic acid (butyric acid), dimethyl disulfide, dimethyl sulfide, iso-valeric, p-cresol, propionic acid, skatole, trimethyl amine, and valeric acid in air. They included acetic acid, p-cresol, iso-butyric acid, butyric acid, indole, phenol, propionic acid, iso-valeric acid, and skatole in manure. In dust samples, they were acetic acid, propionic acid, butyric acid, valeric acid, p-cresol, hexanal, and decanal. Swine facility VOCs were preferentially bound to smaller-size dusts. Identification and quantification of VOCs were restricted by using instruments based on

  3. Safeguards Automated Facility Evaluation (SAFE) methodology

    International Nuclear Information System (INIS)

    Chapman, L.D.; Grady, L.M.; Bennett, H.A.; Sasser, D.W.; Engi, D.

    1978-08-01

    An automated approach to facility safeguards effectiveness evaluation has been developed. This automated process, called Safeguards Automated Facility Evaluation (SAFE), consists of a collection of a continuous stream of operational modules for facility characterization, the selection of critical paths, and the evaluation of safeguards effectiveness along these paths. The technique has been implemented on an interactive computer time-sharing system and makes use of computer graphics for the processing and presentation of information. Using this technique, a comprehensive evaluation of a safeguards system can be provided by systematically varying the parameters that characterize the physical protection components of a facility to reflect the perceived adversary attributes and strategy, environmental conditions, and site operational conditions. The SAFE procedure has broad applications in the nuclear facility safeguards field as well as in the security field in general. Any fixed facility containing valuable materials or components to be protected from theft or sabotage could be analyzed using this same automated evaluation technique

  4. Development of integrated containment and surveillance system for fast critical facility FCA. Portal and penetration monitors

    International Nuclear Information System (INIS)

    Mukaiyama, Takehiko; Ogawa, Hironobu; Yokota, Yasuhiro.

    1998-01-01

    Manpower and radiation exposure problems, accompanied by frequent Non Destructive Assay (NDA) based inspections at the Fast Critical Facility FCA of Japan Atomic Energy Research Institute (JAERI), are a burden for both the inspectorates and the facility operator. In the hope of alleviating these burdens, the development of containment and surveillance measures for the FCA was initiated in 1979. The integrated containment and surveillance system consists of a portal monitor and a penetration monitor. The reactor building provides an ideal containment measure because of its explosion-proof, airtight structure and limited number of penetrations. The function of the portal monitor is to detect undeclared removal of nuclear material from the reactor building through the doorway. The penetration monitor is designed for surveillance of diversion routes through containment boundaries, and of safeguards related activities for bypassing the portal monitor. The combination of monitoring by the penetration monitor of containment boundaries and all their penetrations except for the doorway, and monitoring by the portal monitor, provides complete coverage of realistic diversion routes. The development of the system was completed in 1988 and the field trial test was conducted for the period of twelve running months. The final report on the field trial was concluded on January 1990. The major conclusion of the report was that the system is effective, reliable and efficient. Following this successful conclusion, the International Atomic Energy Agency (IAEA) accepted the system for meeting its safeguards goals at the FCA on condition that an independent IAEA authentication equipment is provided. The development of the authentication equipment is accomplished as an separate Japan Support Programme for Agency Safeguards (JASPAS) task. (author)

  5. Critical safety function guidelines for experimental fusion facilities

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1989-01-01

    As fusion experiments proceed toward deuterium-tritium operation, more attention is being given to public safety. This paper presents the four classes of functions that fusion experiments must provide to assure safe, stable shutdown and retention of radionuclides. These functions are referred to as critical safety functions (CSFs). Selecting CSFs is an important step in probabilistic risk assessment (PRA). An example of CSF selection and usage for the Compact Ignition Tokamak (CIT) is also presented

  6. Test of the core design methods for the THTR 300 with experimental results from the critical facility KAHTER

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, K.; Huebner, A.; Brandes, S.; Krings, F.

    1974-10-15

    At the Kernforschungsanlage Juelich, core physics experiments with core 1 and core 2 of the critical facility for high temperature reactors KAHTER were carried out in 1973. Core 2 corresponds to the THTR initial core in its moderation ratio S = 7500. Selected experimental results on the critical mass, on control rod worths, and reaction rate distributions were used for testing the most important procedures for the THTR core physics design. The zero-dimensional spectrum program MUPO with its cross section library and the and neutron flux calculations in two-dimensional diffusion approximation by CRAM are of central importance. It proved to be important to introduce modifications specific to the KAHTER plant into the standard models. Thus the void effect (void above the pebble bed) was investigated with DOT-2 by transport theory and a correction was introduced for the critical masses calculated by diffusion theory. Another feature already contained in the standard procedure, the increase of the diffusion constants for the hollow spaces between the spheres, results in a correction of 3.8% < delta-k for KAHTER, whereas in the THTR 300 it only amounts to several tenths % delta-k. Critical masses are predicted with accuracies of < 1.5 % or with regard to reactivity < 0.65 % delta-k. The calculated values for the radial neutron flux distributions deviate from the measured values in the core area by approximately lo %. In the case of the axial profiles, deviations are observed at the pebble bed surface which can be explained by the upper void, which cannot be satisfactorily represented by the diffusion theory. Control rod worths are predicted quite well, i.e., to within +/- 5%. An exception is the bank of 4 reflector rods, where the applied model of the "grey curtain" is not accurate because of the large distances between rods. The calculated control rod worths for that case were found to be too low, which does, however, not result in a safety problem.

  7. USNRC licensing process as related to nuclear criticality safety

    International Nuclear Information System (INIS)

    Ketzlach, N.

    1987-01-01

    The U.S. Code of Federal Regulations establishes procedures and criteria for the issuance of licenses to receive title to, own, acquire, deliver, receive, possess, use, and initially transfer special nuclear material; and establishes and provides for the terms and conditions upon which the Nuclear Regulatory Commission (NRC) will issue such licenses. Section 70.22 of the regulations, ''Contents of Applications'', requires that applications for licenses contain proposed procedures to avoid accidental conditions of criticality. These procedures are elements of a nuclear criticality safety program for operations with fissionable materials at fuels and materials facilities (i.e., fuel cycle facilities other than nuclear reactors) in which there exists a potential for criticality accidents. To assist the applicant in providing specific information needed for a nuclear criticality safety program in a license application, the NRC has issued regulatory guides. The NRC requirements for nuclear criticality safety include organizational, administrative, and technical requirements. For purely technical matters on nuclear criticality safety these guides endorse national standards. Others provide guidance on the standard format and content of license applications, guidance on evaluating radiological consequences of criticality accidents, or guidance for dealing with other radiation safety issues. (author)

  8. Critical safety function guidelines for experimental fusion facilities

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1989-01-01

    As fusion experiments proceed toward deuterium-tritium operation, more attention is being given to public safety. This paper presents the four classes of functions that fusion experiments must provide to assure safe, stable shutdown and retention of radionuclides. These functions are referred to as critical safety functions (CSFs). Selecting CSFs is an important step in probabilistic risk assessment (PRA). An example of CSF selection and usage for the Compact Ignition Tokamak (CIT) is also presented. 10 refs., 6 figs

  9. Preparation of data for criticality safety evaluation of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Okuno, Hiroshi; Suyama, Kenya; Yoshiyama, Hiroshi; Tonoike, Kotaro; Miyoshi, Yoshinori

    2005-01-01

    Nuclear Criticality Safety Handbook/Data Collection, Version 2 was submitted to the Ministry of Education, Culture, Sports, Science and Technology (MEXT) of Japan as a contract report. In this presentation paper, its outline and related recent works are presented. After an introduction in Chapter 1, useful information to obtain the atomic number densities was collected in Chapter 2. The nuclear characteristic parameters for 11 nuclear fuels were provided in Chapter 3, and subcriticality judgment graphs were given in Chapter 4. The estimated critical and estimated lower-limit critical values were supplied for the 11 nuclear fuels as results of calculations by using the Japanese Evaluated Nuclear Data Library, JENDL-3.2, and the continuous energy Monte Carlo neutron transport code MVP in Chapter 5. The results of benchmark calculations based on the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook were summarized into six fuel categories in Chapter 6. As for recent works, subcriticality judgment graphs for U-SiO 2 and Pu-SiO 2 were obtained. Benchmark calculations were made with the combination of the latest version of the library JENDL-3.3 and MVP code for a series of STACY experiments and the estimated critical and estimated lower-limit critical values of 10 wt%-enriched uranium nitrate solutions were calculated. (author)

  10. Obstetric critical care services in South Africa

    Directory of Open Access Journals (Sweden)

    Gladness Nethathe

    2015-01-01

    Full Text Available More than half of all global maternal deaths occur in Africa. A large percentage of these deaths are preventable, and lack of access to adequate critical care facilities is a contributing factor. There are limited published data on the clinical and management challenges presented by the critically ill obstetric patient admitted to the intensive care unit in our setting, and more data are required in order to better define the critical care needs of this group of patients.

  11. Analysis of fuel management in the KIPT neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhong Zhaopeng, E-mail: zzhong@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Gohar, Yousry; Talamo, Alberto [Nuclear Engineering Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

    2011-05-15

    Research highlights: > Fuel management of KIPT ADS was analyzed. > Core arrangement was shuffled in stage wise. > New fuel assemblies was added into core periodically. > Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is {approx}360 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

  12. Outreach on a National Scale: The Critical Role of Facilities

    Science.gov (United States)

    Bartel, B. A.; Charlevoix, D. J.

    2015-12-01

    Facilities provide infrastructure for science that would not be feasible at a single institution. Facilities are also a resource for development of outreach products and activities that reach a national audience of diverse stakeholders. UNAVCO manages the NSF geodetic facility GAGE (Geodesy Advancing Geosciences and Earthscope). Staff at UNAVCO with expertise in education, outreach, and communication translate the science and supporting infrastructure into materials consumable by a wide array of users including teachers, students, museum attendees, emergency managers, park interpreters, and members of the general public. UNAVCO has the ability to distribute materials to a national and international audience, thereby greatly increasing the impact of the science and increasing the value of the investment by the National Science Foundation. In 2014 and 2015, UNAVCO produced multiple print products focused on the Plate Boundary Observatory (PBO), the geodetic component of EarthScope. Products include a deck of playing cards featuring PBO GPS stations, a poster featuring GPS velocities of the Western United States, and another poster focused on GPS velocities in Alaska. We are distributing these products to a broad audience, including teachers, station permit holders, and community members. The Tectonics of the Western United States poster was distributed this year in the American Geosciences Institute Earth Science Week kit for teachers, reaching 16,000 educators around the country. These posters and the PBO playing cards (PBO-52) were distributed to more than 100 teachers through workshops led by UNAVCO, the EarthScope National Office, the Southern California Earthquake Center (SCEC), and more. Additionally, these cards serve as a way to engage landowners who host these scientific stations on their property. This presentation will address the strategies for creating nationally relevant materials and the tools used for dissemination of materials to a broad audience. We

  13. CRITICALITY SAFETY LIMIT EVALUATION PROGRAM (CSLEP's) AND QUICK SCREENS: ANSWERS TO EXPEDITED PROCESSING LEGACY CRITICALITY SAFETY LIMITS AND EVALUATIONS

    International Nuclear Information System (INIS)

    TOFFER, H.

    2006-01-01

    Since the end of the cold war, the need for operating weapons production facilities has faded. Criticality Safety Limits and controls supporting production modes in these facilities became outdated and furthermore lacked the procedure based rigor dictated by present day requirements. In the past, in many instances, the formalism of present day criticality safety evaluations was not applied. Some of the safety evaluations amounted to a paragraph in a notebook with no safety basis and questionable arguments with respect to double contingency criteria. When material stabilization, clean out, and deactivation activities commenced, large numbers of these older criticality safety evaluations were uncovered with limits and controls backed up by tenuous arguments. A dilemma developed: on the one hand, cleanup activities were placed on very aggressive schedules; on the other hand, a highly structured approach to limits development was required and applied to the cleanup operations. Some creative approaches were needed to cope with the limits development process

  14. Critical experiments at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Harms, Gary A.; Ford, John T.; Barber, Allison Delo

    2010-01-01

    Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-III is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark

  15. The Development, Content, Design, and Conduct of the 2011 Piloted US DOE Nuclear Criticality Safety Program Criticality Safety Engineering Training and Education Project

    International Nuclear Information System (INIS)

    Hopper, Calvin Mitchell

    2011-01-01

    In May 1973 the University of New Mexico conducted the first nationwide criticality safety training and education week-long short course for nuclear criticality safety engineers. Subsequent to that course, the Los Alamos Critical Experiments Facility (LACEF) developed very successful 'hands-on' subcritical and critical training programs for operators, supervisors, and engineering staff. Since the inception of the US Department of Energy (DOE) Nuclear Criticality Technology and Safety Project (NCT and SP) in 1983, the DOE has stimulated contractor facilities and laboratories to collaborate in the furthering of nuclear criticality as a discipline. That effort included the education and training of nuclear criticality safety engineers (NCSEs). In 1985 a textbook was written that established a path toward formalizing education and training for NCSEs. Though the NCT and SP went through a brief hiatus from 1990 to 1992, other DOE-supported programs were evolving to the benefit of NCSE training and education. In 1993 the DOE established a Nuclear Criticality Safety Program (NCSP) and undertook a comprehensive development effort to expand the extant LACEF 'hands-on' course specifically for the education and training of NCSEs. That successful education and training was interrupted in 2006 for the closing of the LACEF and the accompanying movement of materials and critical experiment machines to the Nevada Test Site. Prior to that closing, the Lawrence Livermore National Laboratory (LLNL) was commissioned by the US DOE NCSP to establish an independent hands-on NCSE subcritical education and training course. The course provided an interim transition for the establishment of a reinvigorated and expanded two-week NCSE education and training program in 2011. The 2011 piloted two-week course was coordinated by the Oak Ridge National Laboratory (ORNL) and jointly conducted by the Los Alamos National Laboratory (LANL) classroom education and facility training, the Sandia National

  16. MCNPTM criticality primer and training experiences

    International Nuclear Information System (INIS)

    Briesmeister, J.; Forster, R.A.; Busch, R.

    1995-01-01

    With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, the analyst may have little experience with the specific codes available at his or her facility. Usually, the codes are quite complex, black boxes capable of analyzing numerous problems with a myriad of input options. Documentation for these codes is designed to cover all the possible configurations and types of analyses but does not give much detail on any particular type of analysis. For criticality calculations, the user of a code is primarily interested in the value of the effective multiplication factor for a system (k eff ). Most codes will provide this, and truckloads of other information that may be less pertinent to criticality calculations. Based on discussions with code users in the nuclear criticality safety community, it was decided that a simple document discussing the ins and outs of criticality calculations with specific codes would be quite useful. The Transport Methods Group, XTM, at Los Alamos National Laboratory (LANL) decided to develop a primer for criticality calculations with their Monte Carlo code, MCNP. This was a joint task between LANL with a knowledge and understanding of the nuances and capabilities of MCNP and the University of New Mexico with a knowledge and understanding of nuclear criticality safety calculations and educating first time users of neutronics calculations. The initial problem was that the MCNP manual just contained too much information. Almost everything one needs to know about MCNP can be found in the manual; the problem is that there is more information than a user requires to do a simple k eff calculation. The basic concept of the primer was to distill the manual to create a document whose only focus was criticality calculations using MCNP

  17. USAEC Controls for Nuclear Criticality Safety

    Energy Technology Data Exchange (ETDEWEB)

    McCluggage, W. C. [Division of Operational Safety, United States Atomic Energy Commission Washington, DC (United States)

    1966-05-15

    This is a paper written to provide a broad general view of the United States Atomic Energy Commission's controls for nuclear criticality safety within its own facilities. Included also is a brief' discussion of the USAEC's methods of obtaining assurance that the controls are being applied. The body of the document contains three sections. The first two describe the functions of the USAEC; the third deals with the contractors. The provisions of the Atomic Energy Act applicable to health and safety are discussed in relation to nuclear criticality safety. The use of United States Atomic Energy Commission manual chapters and Federal regulations is described. The functions of the USAEC Headquarters' offices and the operations offices are briefly outlined. Comments regarding the USAEC's inspection, auditing and appraisal programmes are included. Also briefly mentioned are the basic qualifications which must be met to become a contractor to possess and process or use fissionable materials. On the plant, factory or facility level the duties and responsibilities of industrial management are briefly outlined. The fundamental standards and their origin, together with the principal documents and guides are mentioned. The chief methods of control used by contractors operating large USAEC facilities and plants are described and compared. These include diagrams of how a typical nuclear criticality safety problem is handled from inception, design, construction and finally plant operation. Also included is a brief discussion of the contractors' methods of assuring strict employee compliance with the operating rules and limits. (author)

  18. Review of WHC criticality safety audit findings for 1970-1981

    International Nuclear Information System (INIS)

    Rogers, C.A.; Paglieri, J.N.

    1984-01-01

    At Westinghouse Hanford Company (WHC) all fissionable material handling must meet DOE requirements for safety. This necessitates a program of regular audits by the Safety group to verify compliance with criticality safety limits and controls and to alert facility management to observed discrepancies and potential problems. Audits of fissionable material facilities by Safety are required at least once every 6 months, but in practice are conducted more frequently. This paper summarizes findings from over 400 criticality safety audits conducted by Safety between July 1970 and July 1981 in seven fissionable material facilities to show their types and frequencies of occurrence. All limit violations occurring during this period are summarized, including those found by the operating group. 1 ref., 1 tab

  19. The epidemiological studies of leukemia around nuclear facilities for children and young adults: critical review

    International Nuclear Information System (INIS)

    2008-01-01

    This objective of this report is to make a review of studies relative to the risk of leukemia among children and young adults less than twenty five years old near the nuclear facilities. the nuclear facilities considered in this report are nuclear power plants (electric power generation), the nuclear research centers, the fuel or weapons fabrication plants, reprocessing plants. This report does not describe the risk analysis near the sites of nuclear weapons test, the consequences of accidents on nuclear facilities such Chernobylsk or the the population near the military and industrial site of Mayak (Ural). The same is for the mining sites and the facilities of uranium extraction that are out of this report. (N.C.)

  20. Criticality calculations with MCNP trademark: A primer

    International Nuclear Information System (INIS)

    Harmon, C.D. II; Busch, R.D.; Briesmeister, J.F.; Forster, R.A.

    1994-01-01

    With the closure of many experimental facilities, the nuclear criticality safety analyst increasingly is required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his/her facility. This primer will help you, the analyst, understand and use the MCNP Monte Carlo code for nuclear criticality safety analyses. It assumes that you have a college education in a technical field. There is no assumption of familiarity with Monte Carlo codes in general or with MCNP in particular. Appendix A gives an introduction to Monte Carlo techniques. The primer is designed to teach by example, with each example illustrating two or three features of MCNP that are useful in criticality analyses. Beginning with a Quickstart chapter, the primer gives an overview of the basic requirements for MCNP input and allows you to run a simple criticality problem with MCNP. This chapter is not designed to explain either the input or the MCNP options in detail; but rather it introduces basic concepts that are further explained in following chapters. Each chapter begins with a list of basic objectives that identify the goal of the chapter, and a list of the individual MCNP features that are covered in detail in the unique chapter example problems. It is expected that on completion of the primer you will be comfortable using MCNP in criticality calculations and will be capable of handling 80 to 90 percent of the situations that normally arise in a facility. The primer provides a set of basic input files that you can selectively modify to fit the particular problem at hand

  1. Analysis of Elektrogorsk 108 test facility experimental data

    International Nuclear Information System (INIS)

    Urbonas, R.

    2001-01-01

    In the paper an evaluation of experimental data obtained at Russian Elektrogorsk 108 (E-108) test facility is presented. E-108 facility is a scaled model of Russian RBMK design reactor. An attempt to validate state-of-the-art thermal hydraulic codes on the basis of E-108 test facility was made. Originally these codes were developed and validated for BWRs and PWRs. Since state-of-art thermal hydraulic codes are widely used for simulation of RBMK reactors further codes' implementation and validation is required. The facility was modelled by employing RELAP5 (INEEL, USA) thermal hydraulic system analysis best estimate code. The results show dependence from number of nodes used in the heated channels, frictional and form losses employed. The obtained oscillatory behaviour is resulted by density wave and critical heat flux. It is shown that codes are able to predict thermal hydraulic instability and sudden heat structure temperature excursion, when critical heat flux is approached, well. In addition, an uncertainty analysis of one of the experiments was performed by employing GRS developed System for Uncertainty and Sensitivity Analysis (SUSA). It was one of the first attempts to use this statistic-based methodology in Lithuania.(author)

  2. Criticality accident studies and methodology implemented at the CEA

    International Nuclear Information System (INIS)

    Barbry, Francis; Fouillaud, Patrick; Reverdy, Ludovic; Mijuin, Dominique

    2003-01-01

    Based on the studies and results of experimental programs performed since 1967 in the CRAC, then SILENE facilities, the CEA has devised a methodology for criticality accident studies. This methodology integrates all the main focuses of its approach, from criticality accident phenomenology to emergency planning and response, and thus includes aspects such as criticality alarm detector triggering, airborne releases, and irradiation risk assessment. (author)

  3. French en engineering and operation rules for plutonium facilities

    International Nuclear Information System (INIS)

    Bertolotti, G.; Drain, F.; Dubois, G.; Monnatte, J.; Mathieu, P.

    1998-01-01

    COGEMA is operating large size purifying and conditioning plutonium facilities at LA HAGUE and MOX fuels fabrication plant at Marcoule. A high safety standard is recognised for these facilities. It is mainly based on : - prevention of spreading of radioactive materials to workers and environment by physical barriers ensuring static containment and by a cascade of pressure differentials ensuring dynamic containment, - radiation shielding and remote controlled processes ensuring very low dose to workers, - prevention of criticality accident by criticality control methods and double contingency principle, - prevention of fire risks by control of ignition sources, adequate management of combustible materials, physical separation between zones where there is a risk of fire and the remainder of the facility. The facilities are operated while respecting safety requirements as described in the General Operating Rules. The equipment involved in safety functions are monitored and periodically checked. Continuous improvement by incorporation of feed back of safety experience results in: - effective decrease of exposure to operating staff; - reduction of solid waste, liquid and gaseous effluents; - no significant incident recorded. (author)

  4. Criticality considerations for 233U fuels in an HTGR fuel refabrication facility

    International Nuclear Information System (INIS)

    McNeany, S.R.; Jenkins, J.D.

    1978-01-01

    Eleven 233 U solution critical assemblies spanning an H/ 233 U ratio range of 40 to 2000 and a bare metal 233 U assembly have been calculated with the ENDF/B-IV and Hansen-Roach cross sections. The results from these calculations are compared with the experimental results and with each other. An increasing disagreement between calculations with ENDF/B and Hansen-Roach data with decreasing H/ 233 U ratio was observed, indicative of large differences in their intermediate energy cross sections. The Hansen-Roach cross sections appeared to give reasonably good agreement with experiments over the whole range; whereas the ENDF/B calculations yielded high values for k/sub eff/ on assemblies of low moderation. It is concluded that serious problems exist in the ENDF/B-IV representation of the 233 U cross sections in the intermediate energy range and that further evaluation of this nuclide is warranted. In addition, it is recommended that an experimental program be undertaken to obtain 233 U criticality data at low H/ 233 U ratios for verification of generalized criticality safety guidelines. Part II of this report presents the results of criticality calculations on specific pieces of equipment required for HTGR fuel refabrication. In particular, fuel particle storage hoppers and resin carbonization furnaces are criticality safe up to 22.9 cm (9.0 in.) in diameter providing water or other hydrogenous moderators are excluded. In addition, no criticality problems arise due to accumulation of particles in the off-gas scrubber reservoirs provided reasonable administrative controls are exercised

  5. Test Facilities and Experience on Space Nuclear System Developments at the Kurchatov Institute

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoi, Nikolai N.; Garin, Vladimir P.; Glushkov, Evgeny S.; Kompaniets, George V.; Kukharkin, Nikolai E.; Madeev, Vicktor G.; Papin, Vladimir K.; Polyakov, Dmitry N.; Stepennov, Boris S.; Tchuniyaev, Yevgeny I.; Tikhonov, Lev Ya.; Uksusov, Yevgeny I.

    2004-01-01

    The complexity of space fission systems and rigidity of requirement on minimization of weight and dimension characteristics along with the wish to decrease expenditures on their development demand implementation of experimental works which results shall be used in designing, safety substantiation, and licensing procedures. Experimental facilities are intended to solve the following tasks: obtainment of benchmark data for computer code validations, substantiation of design solutions when computational efforts are too expensive, quality control in a production process, and 'iron' substantiation of criticality safety design solutions for licensing and public relations. The NARCISS and ISKRA critical facilities and unique ORM facility on shielding investigations at the operating OR nuclear research reactor were created in the Kurchatov Institute to solve the mentioned tasks. The range of activities performed at these facilities within the implementation of the previous Russian nuclear power system programs is briefly described in the paper. This experience shall be analyzed in terms of methodological approach to development of future space nuclear systems (this analysis is beyond this paper). Because of the availability of these facilities for experiments, the brief description of their critical assemblies and characteristics is given in this paper

  6. Analysis of fuel management in the KIPT neutron source facility

    International Nuclear Information System (INIS)

    Zhong Zhaopeng; Gohar, Yousry; Talamo, Alberto

    2011-01-01

    Research highlights: → Fuel management of KIPT ADS was analyzed. → Core arrangement was shuffled in stage wise. → New fuel assemblies was added into core periodically. → Beryllium reflector could also be utilized to increase the fuel life. - Abstract: Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an experimental neutron source facility consisting of an electron accelerator driven sub-critical assembly. The neutron source driving the sub-critical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The sub-critical assembly surrounding the target is fueled with low enriched WWR-M2 type hexagonal fuel assemblies. The U-235 enrichment of the fuel material is <20%. The facility will be utilized for basic and applied research, producing medical isotopes, and training young specialists. With the 100 KW electron beam power, the total thermal power of the facility is ∼360 kW including the fission power of ∼260 kW. The burnup of the fissile materials and the buildup of fission products continuously reduce the system reactivity during the operation, decrease the neutron flux level, and consequently impact the facility performance. To preserve the neutron flux level during the operation, the fuel assemblies should be added and shuffled for compensating the lost reactivity caused by burnup. Beryllium reflector could also be utilized to increase the fuel life time in the sub-critical core. This paper studies the fuel cycles and shuffling schemes of the fuel assemblies of the sub-critical assembly to preserve the system reactivity and the neutron flux level during the operation.

  7. Nuclear Criticality Safety Data Book

    Energy Technology Data Exchange (ETDEWEB)

    Hollenbach, D. F. [Y-12 National Security Complex, Oak Ridge, TN (United States)

    2016-11-14

    The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

  8. Nuclear Criticality Safety Data Book

    International Nuclear Information System (INIS)

    Hollenbach, D. F.

    2016-01-01

    The objective of this document is to support the revision of criticality safety process studies (CSPSs) for the Uranium Processing Facility (UPF) at the Y-12 National Security Complex (Y-12). This design analysis and calculation (DAC) document contains development and justification for generic inputs typically used in Nuclear Criticality Safety (NCS) DACs to model both normal and abnormal conditions of processes at UPF to support CSPSs. This will provide consistency between NCS DACs and efficiency in preparation and review of DACs, as frequently used data are provided in one reference source.

  9. Reducing cooling energy consumption in data centres and critical facilities

    Science.gov (United States)

    Cross, Gareth

    Given the rise of our everyday reliance on computers in all walks of life, from checking the train times to paying our credit card bills online, the need for computational power is ever increasing. Other than the ever-increasing performance of home Personal Computers (PC's) this reliance has given rise to a new phenomenon in the last 10 years ago. The data centre. Data centres contain vast arrays of IT cabinets loaded with servers that perform millions of computational equations every second. It is these data centres that allow us to continue with our reliance on the internet and the PC. As more and more data centres become necessary due to the increase in computing processing power required for the everyday activities we all take for granted so the energy consumed by these data centres rises. Not only are more and more data centres being constructed daily, but operators are also looking at ways to squeeze more processing from their existing data centres. This in turn leads to greater heat outputs and therefore requires more cooling. Cooling data centres requires a sizeable energy input, indeed to many megawatts per data centre site. Given the large amounts of money dependant on the successful operation of data centres, in particular for data centres operated by financial institutions, the onus is predominantly on ensuring the data centres operate with no technical glitches rather than in an energy conscious fashion. This report aims to investigate the ways and means of reducing energy consumption within data centres without compromising the technology the data centres are designed to house. As well as discussing the individual merits of the technologies and their implementation technical calculations will be undertaken where necessary to determine the levels of energy saving, if any, from each proposal. To enable comparison between each proposal any design calculations within this report will be undertaken against a notional data facility. This data facility will

  10. Status of the SXFEL Facility

    Directory of Open Access Journals (Sweden)

    Zhentang Zhao

    2017-06-01

    Full Text Available The Shanghai soft X-ray Free-Electron Laser facility (SXFEL is being developed in two steps; the SXFEL test facility (SXFEL-TF, and the SXFEL user facility (SXFEL-UF. The SXFEL-TF is a critical development step towards the construction a soft X-ray FEL user facility in China, and is under commissioning at the Shanghai Synchrotron Radiation Facility (SSRF campus. The test facility is going to generate 8.8 nm FEL radiation using an 840 MeV electron linac passing through the two-stage cascaded HGHG-HGHG or EEHG-HGHG (high-gain harmonic generation, echo-enabled harmonic generation scheme. The construction of the SXFEL-TF started at the end of 2014. Its accelerator tunnel and klystron gallery were ready for equipment installation in April 2016, and the installation of the SXFEL-TF linac and radiator undulators were completed by the end of 2016. In the meantime, the SXFEL-UF, with a designated wavelength in the water window region, began construction in November 2016. This was based on upgrading the linac energy to 1.5 GeV, and the building of a second undulator line and five experimental end-stations. Construction status and the future plans of the SXFEL are reported in this paper.

  11. NMC and A and nuclear criticality safety systems integration: A prospective way for enhancement of the nuclear industry facilities safety

    International Nuclear Information System (INIS)

    Ryazanov, Boris G.; Sviridov, Victor I.; Frolov, Vladimir V.; Shvedov, Maxim O.; Mclaughlin, Thomas P.; Pruvost, Norman L.

    2003-01-01

    A considerable body of data has now been acquired about the principles, parameters and consequences of nuclear (criticality) accidents at facilities of the atomic industry in Russia, the United States, Great Britain and Japan. The total number of such accidents stands at 22. Russian and US specialists have prepared a rather extensive survey and analysis of these accidents. The final and important section of this survey is the lessons implied by the results of analysis of these 22 accidents. Among these lessons is the necessity of unconditional enforcement of control over the movement and transformations of special nuclear materials (SNM), and in particular fissile materials, (those SNMs with criticality accident concerns) during production and processing. Inadequacies in such control have been among the causes of most of the accidents that have occurred. Nuclear materials control and accounting (MC and A) for the purpose of ensuring storage reliability and nonproliferation safeguards is a major task of nuclear facilities in any nation. MC and A systems use the latest techniques and hardware for periodic control of SNM in specifically organized material balance areas. Immediate checking, periodic inventory of SNM, and measurements of the parameters of SNM at key points are the main sources of data for these systems. Data about the presence and sites of location of SNM in material balance areas that are acquired in inventories can be used for objective assessment of the status of nuclear safety. On the other hand, the inventory itself involves performance of operations that are unlike routine process engineering, and require special consideration of nuclear safety. Use of the techniques and hardware of MC and A systems not only for purposes of storage reliability, but also to ensure nuclear safety, will reduce the risk of nuclear accidents. This paper gives a concise overview of nuclear accidents that have occurred due to inadequacies in MC and A, and demonstrates

  12. Critical experiments at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Harms, G.A.; Ford, J.T.; Barber, A.D.

    2011-01-01

    Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-III is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark

  13. Critical experiments at Sandia National Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Harms, G.A.; Ford, J.T.; Barber, A.D., E-mail: gaharms@sandia.gov [Sandia National Laboratories, Albuquerque, NM (United States)

    2011-07-01

    Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-III is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide

  14. The QUASAR facility

    Science.gov (United States)

    Gates, David

    2013-10-01

    The QUAsi-Axisymmetric Research (QUASAR) stellarator is a new facility which can solve two critical problems for fusion, disruptions and steady-state, and which provides new insights into the role of magnetic symmetry in plasma confinement. If constructed it will be the only quasi-axisymmetric stellarator in the world. The innovative principle of quasi-axisymmetry (QA) will be used in QUASAR to study how ``tokamak-like'' systems can be made: 1) Disruption-free, 2) Steady-state with low recirculating power, while preserving or improving upon features of axisymmetric tokamaks, such as 1) Stable at high pressure simultaneous with 2) High confinement (similar to tokamaks), and 3) Scalable to a compact reactor Stellarator research is critical to fusion research in order to establish the physics basis for a magnetic confinement device that can operate efficiently in steady-state, without disruptions at reactor-relevant parameters. The two large stellarator experiments - LHD in Japan and W7-X under construction in Germany are pioneering facilities capable of developing 3D physics understanding at large scale and for very long pulses. The QUASAR design is unique in being QA and optimized for confinement, stability, and moderate aspect ratio (4.5). It projects to a reactor with a major radius of ~8 m similar to advanced tokamak concepts. It is striking that (a) the EU DEMO is a pulsed (~2.5 hour) tokamak with major R ~ 9 m and (b) the ITER physics scenarios do not presume steady-state behavior. Accordingly, QUASAR fills a critical gap in the world stellarator program. This work supported by DoE Contract No. DEAC02-76CH03073.

  15. PROMILLE database as a part of JNC reactor physics analytical system for BFS-2 fast critical facility experiments analysis

    International Nuclear Information System (INIS)

    Bednyakov, Sergey

    2000-12-01

    The PROMILLE database for experimental data from the BFS-2 fast critical facility (Institute of Physics and Power Engineering (IPPE), Russia) has been developed and embedded into the JNC reactor physics analytical system to provide a strict documentation format, a common data source for different analytical tools and a unique export interface with different reactor codes. PROMILLE should be considered not only as a database but also as a bank of interfaces because of its dynamic role in the analytical process. The database currently accepts data from the simulation materials (pellets, tubes and bars) as well as full cores descriptions. A core description involves all different unit cells forming loading elements, all types of the loading elements forming a layout and the layout itself. In fact it is a description of criticality experiments. Export interfaces for the CITATION-FBR code and the SLAROM and CASUP codes have been developed. The PROMILLE software was developed with MS Visual Basic 6.0 and the data is kept in the data tables generated with the MS Access database management system. Data for eight BFS-2 assembly configurations have been incorporated. They include BFS-58-1i1 uranium-free plutonium assembly with inert material included in its fuel matrix and also seven BFS-62 modifications simulating different stages of investigation of MOX fuel based BN-600 core. (author)

  16. Modeling of criticality accidents and their environmental consequences

    International Nuclear Information System (INIS)

    Thomas, W.; Gmal, B.

    1987-01-01

    In the Federal Republic of Germany, potential radiological consequences of accidental nuclear criticality have to be evaluated in the licensing procedure for fuel cycle facilities. A prerequisite to this evaluation is to establish conceivable accident scenarios. First, possibilities for a criticality exceeding the generally applied double contingency principle of safety are identified by screening the equipment and operation of the facility. Identification of undetected accumulations of fissile material or incorrect transfer of fissile solution to unfavorable geometry normally are most important. Second, relevant and credible scenarios causing the most severe consequences are derived from these possibilities. For the identified relevant scenarios, time-dependent fission rates and reasonable numbers for peak power and total fissions must be determined. Experience from real accidents and experiments (KEWB, SPERT, CRAC, SILENE) has been evaluated using empirical formulas. To model the time-dependent behavior of criticality excursions in fissile solutions, a computer program FELIX has been developed

  17. MYRRHA. An innovative and unique research facility

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, Rafaeol; Neerdael, Bernard; Schyns, Marc; Dyck, Steven Van; Michiels, Sidney; Ait Abderrahim, Hamid, E-mail: myrrha@sckcen.be [Belgian Nuclear Research Centre (SCK-CEN), Mol (Belgium)

    2012-03-15

    The MYRRHA project started in 1998 by SCK{center_dot}CEN in collaboration with Ion Beam Applications (IBA, Louvain-la-Neuve), as an upgrade of the ADONIS project. MYRRHA is designed as a multi-purpose irradiation facility in order to support research programmes on fission and fusion reactor structural materials and nuclear fuel development. Applications of these are found in Accelerator Driven Systems (ADS) systems and in present generation as well as in next generation critical reactors. The first objective of MYRRHA however, will be to demonstrate on one hand the ADS concept at a reasonable power level and on the other hand the technological feasibility of transmutation of Minor Actinides (MA) and Long-Lived Fission Products (LLFP) arising from the reprocessing of radioactive waste. MYRRHA will also help the development of the Pb-alloys technology needed for the LFR (Lead Fast Reactor) Gen.IV concept. Transmutation of MA can be completed in an efficient way in fast neutron spectrum facilities. Both critical reactors and sub-critical ADS are potential candidates as dedicated transmutation systems. However, critical reactors, heavily loaded with fuel containing large amounts of MA, pose safety problems caused by unfavourable reactivity coefficients due to the little delayed neutron fraction. A sub-critical ADS operates in a flexible and safe manner even with a core loading containing a high amount of MA leading to achieve a high efficient transmutation. Thus, the sub-criticality is not a virtue but rather a necessity for an efficient and economical burning of the MA. Besides the reduction of the HLW burden, the MYRRHA project will serve the purpose of developing the lead alloys technology as a reactor coolant that can be used in one of the Generation IV reactor concepts namely the Lead Fast Reactor (LFR). Although carrying out the MYRRHA project will lead to the demonstration of the efficient and safe transmutation of MA in ADS systems as the ultimate goal the

  18. Experimental performance and results of the critical pebble bed facility KAHTER

    Energy Technology Data Exchange (ETDEWEB)

    Krings, F. J.; Drueke, V.; Kirch, N.; Neef, R. D.

    1974-10-15

    The paper provides a description and results of critical experiments performed in KAHTER fueled with pebbles containing coated particles of HEU/Th oxide with a ratio of uranium-to-thorium of 1.1:5. Core loadings with varying amounts of fuel and solid graphite pebbles were tested with fuel-to-graphite pebble ratios of 3:1, 1:1, and 1:3. Tests included criticality for various fuel loadings with all control rods removed, control rod worths for reflector-mounted control as single rods and in a bank and control worths for a central control rod, reaction rates by flux wire activations (Dy, Mn, In, Au, and U-235) and detector measurements (BF3 and fission chamber), simulated xenon stability testing using the motions of a Cf-252 source and Cd-absorber observed by an externally-located BF3 detector, and the reactivity worth of a Hf burnable absorber. For calculations of the room-temperature zero-power critical experiment, the values for nitrogen and hydrogen contents of the graphite were taken from previous experiments in CESAR.

  19. Modeling of FREYA fast critical experiments with the Serpent Monte Carlo code

    International Nuclear Information System (INIS)

    Fridman, E.; Kochetkov, A.; Krása, A.

    2017-01-01

    Highlights: • FREYA – the EURATOM project executed to support fast lead-based reactor systems. • Critical experiments in the VENUS-F facility during the FREYA project. • Characterization of the critical VENUS-F cores with Serpent. • Comparison of the numerical Serpent results to the experimental data. - Abstract: The FP7 EURATOM project FREYA has been executed between 2011 and 2016 with the aim of supporting the design of fast lead-cooled reactor systems such as MYRRHA and ALFRED. During the project, a number of critical experiments were conducted in the VENUS-F facility located at SCK·CEN, Mol, Belgium. The Monte Carlo code Serpent was one of the codes applied for the characterization of the critical VENUS-F cores. Four critical configurations were modeled with Serpent, namely the reference critical core, the clean MYRRHA mock-up, the full MYRRHA mock-up, and the critical core with the ALFRED island. This paper briefly presents the VENUS-F facility, provides a detailed description of the aforementioned critical VENUS-F cores, and compares the numerical results calculated by Serpent to the available experimental data. The compared parameters include keff, point kinetics parameters, fission rate ratios of important actinides to that of U235 (spectral indices), axial and radial distribution of fission rates, and lead void reactivity effect. The reported results show generally good agreement between the calculated and experimental values. Nevertheless, the paper also reveals some noteworthy issues requiring further attention. This includes the systematic overprediction of reactivity and systematic underestimation of the U238 to U235 fission rate ratio.

  20. Large sample NAA of a pottery replica utilizing thermal neutron flux at AHWR critical facility and X-Z rotary scanning unit

    International Nuclear Information System (INIS)

    Acharya, R.; Dasari, K.B.; Pujari, P.K.; Swain, K.K.; Shinde, A.D.; Reddy, A.V.R.

    2013-01-01

    Large sample neutron activation analysis (LSNAA) of a clay pottery replica from Peru was carried out using low neutron flux graphite reflector position of Advanced Heavy Water Reactor (AHWR) critical facility. This work was taken up as a part of inter-comparison exercise under IAEA CRP on LSNAA of archaeological objects. Irradiated large size sample, placed on an X-Z rotary scanning unit, was assayed using a 40% relative efficiency HPGe detector. The k 0 -based internal monostandard NAA (IM-NAA) in conjunction with insitu relative detection efficiency was used to calculate concentration ratios of 12 elements with respect to Na. Analyses of both small and large size samples were carried out to check homogeneity and to arrive at absolute concentrations. (author)

  1. NARCISS critical stand experiments for studying the nuclear safety in accident water immersion of highly enriched uranium dioxide fuel elements

    International Nuclear Information System (INIS)

    Ponomarev-Stepnoj, N.N.; Glushkov, E.S.; Bubelev, V.G.

    2005-01-01

    A brief description of the Topaz-2 SNPS designed under scientific supervision of RRC KI in Russia, and of the NARCISS critical facility, is given. At the NARCISS critical facility, neutronic peculiarities and nuclear safety issues of the Topaz-2 system reactor were studied experimentally. This work is devoted to a detailed description of experiments on investigation of criticality safety in accident water immersion og highly enriched uranium dioxide fuel elements, performed at the NARCISS facility. The experiments were carried out at water-moderated critical assemblies with varying height, number, and spacing of fuel elements. The results obtained in the critical experiments, computational models of the investigated critical configurations, and comparison of the computational and experimental results are given [ru

  2. Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

    International Nuclear Information System (INIS)

    Miller, Thomas Martin; Celik, Cihangir; Dunn, Michael E; Wagner, John C; McMahan, Kimberly L; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Trama, Jean-Christophe; Masse, Veronique; Gagnier, Emmanuel; Naury, Sylvie; Blanc-Tranchant, Patrick; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

    2015-01-01

    In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available

  3. Analysis of Department of Defense Organic Depot Maintenance Capacity Management and Facility Utilization Factors

    Science.gov (United States)

    1991-09-01

    System ( CAPMS ) in lieu of using DODI 4151.15H. Facility utilization rate computation is not explicitly defined; it is merely identified as a ratio of...front of a bottleneck buffers the critical resource and protects against disruption of the system. This approach optimizes facility utilization by...run titled BUFFERED BASELINE. Three different levels of inventory were used to evaluate the effect of increasing the inventory level on critical

  4. Radiation protection at nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Endo, K.; Momose, T.; Furuta, S.

    2011-01-01

    Radiation protection methodologies concerning individual monitoring, workplace monitoring and environmental monitoring in nuclear fuel facilities have been developed and applied to facilities in the Nuclear Fuel Cycle Engineering Laboratories (NCL) of Japan Atomic Energy Agency (JAEA) for over 40 y. External exposure to photon, beta ray and neutron and internal exposure to alpha emitter are important issues for radiation protection at these facilities. Monitoring of airborne and surface contamination by alpha and beta/photon emitters at workplace is also essential to avoid internal exposure. A critical accident alarm system developed by JAEA has been proved through application at the facilities for a long time. A centralised area monitoring system is effective for emergency situations. Air and liquid effluents from facilities are monitored by continuous monitors or sampling methods to comply with regulations. Effluent monitoring has been carried out for 40 y to assess the radiological impacts on the public and the environment due to plant operation. (authors)

  5. Physical security technologies for weapons complex reconfiguration facilities

    International Nuclear Information System (INIS)

    Jaeger, C.D.

    1994-01-01

    Sandia National Laboratories was a member of the Weapons Complex Reconfiguration (WCR) Safeguards and Security (S ampersand S) team providing assistance to the Department of Energy's (DOE) Office of Weapons Complex Reconfiguration. The physical security systems in the new and upgraded facilities being considered for the WCR had to meet DOE orders and other requirements set forth in the WCR Programmatic Design Criteria (PDC), incorporate the latest physical security technologies using proven state-of-the-art systems and meet fundamental security principles. The outcome was to avoid costly retrofits and provide effective and comprehensive protection against current and projected threats with minimal impact on operations, costs and schedule. Physical security requirements for WCR facilities include: (1) reducing S ampersand S life-cycle costs, (2) where feasible automating S ampersand S functions to minimize operational costs, access to critical assets and exposure of people to hazardous environments, (3) increasing the amount of delay to outsider adversary attack, (4) compartmentalizing the facility to minimize the number of personnel requiring access to critical areas and (5) having reliable and maintainable systems. To be most effective against threats physical security must be integrated with facility operations, safety and other S ampersand S activities, such as material control and accountability, nuclear measurements and computer and information security. This paper will discuss the S ampersand S issues, requirements, technology opportunities and needs. Physical security technologies and systems considered in the design effort of the Weapons Complex Reconfiguration facilities will be reviewed

  6. Results of Cesar II critical facility with low enriched fuel balls

    Energy Technology Data Exchange (ETDEWEB)

    Langlet, G; Guerange, J; Laponche, B; Morier, F; Neef, R D; Bock, H J; Kring, F J; Scherer, W

    1972-06-15

    The Cesar facility has been transformed to load in its center a pebble bed fuel. This new Cesar assembly is called Cesar II. The program for the measurements with HTR type fuel balls is managed under a cooperation between physicists of CEA/CADARACHE and KFA/JUELICH. A description of the measuring zones of Cesar II and of the experimental results is given.

  7. 75 FR 18850 - National Protection and Programs Directorate; Chemical Facility Anti-Terrorism Standards...

    Science.gov (United States)

    2010-04-13

    ... Directorate; Chemical Facility Anti-Terrorism Standards Personnel Surety Program AGENCY: National Protection...-terrorism Vulnerability Information (CVI), Sensitive Security Information (SSI), or Protected Critical... . SUPPLEMENTARY INFORMATION: Program Description The Chemical Facility Anti-Terrorism Standards (CFATS), 6 CFR...

  8. Nuclear data for criticality safety

    International Nuclear Information System (INIS)

    Westfall, R.M.

    1994-01-01

    A brief overview is presented on emerging requirements for new criticality safety analyses arising from applications involving nuclear waste management, facility remediation, and the storage of nuclear weapons components. A derivation of criticality analyses from the specifications of national consensus standards is given. These analyses, both static and dynamic, define the needs for nuclear data. Integral data, used primarily for analytical validation, and differential data, used in performing the analyses, are listed, along with desirable margins of uncertainty. Examples are given of needs for additional data to address systems having intermediate neutron energy spectra and/or containing nuclides of intermediate mass number

  9. Tank waste remediation system nuclear criticality safety inspection and assessment plan

    International Nuclear Information System (INIS)

    VAIL, T.S.

    1999-01-01

    This plan provides a management approved procedure for inspections and assessments of sufficient depth to validate that the Tank Waste Remediation System (TWRS) facility complies with the requirements of the Project Hanford criticality safety program, NHF-PRO-334, ''Criticality Safety General, Requirements''

  10. Criticality safety evaluations - a {open_quotes}stalking horse{close_quotes} for integrated safety assessment

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R.A. [Westinghouse Electric Corp., Columbia, SC (United States)

    1995-12-31

    The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility`s criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE.

  11. Operation of a low-level waste disposal facility and how to prevent problems in future facilities

    International Nuclear Information System (INIS)

    Di Sibio, R.

    1985-01-01

    Operation of a low-level waste facility is an ever increasing problem nationally, and specifically one that could grow to crisis proportion in Pennsylvania. There have been, nevertheless, a variety of changes over the years in the management of low level radioactive waste, particularly with regard to disposal facilities that can avert a crisis condition. A number of companies have been organized thru possible a broad range of services to the nuclear industry, including those that emphasize solidification of waste materials, engineering services, waste management, and transportation to disposal sites across the United States. This paper addresses one particular site and the problems which evolved at that site from an environmental perspective. It is important that it is clearly understood that, although these problems are resolvable, the lessons learned here are critical for the prevention of problems at future facilities. The focus of this paper is on the Maxey Flats, Kentucky disposal facility which was closed in 1977. It must be understood that the regulations for siting, management, burial techniques, waste classification, and the overall management of disposal sites were limited when this facility was in operation

  12. Development and utilization of USGS ShakeCast for rapid post-earthquake assessment of critical facilities and infrastructure

    Science.gov (United States)

    Wald, David J.; Lin, Kuo-wan; Kircher, C.A.; Jaiswal, Kishor; Luco, Nicolas; Turner, L.; Slosky, Daniel

    2017-01-01

    The ShakeCast system is an openly available, near real-time post-earthquake information management system. ShakeCast is widely used by public and private emergency planners and responders, lifeline utility operators and transportation engineers to automatically receive and process ShakeMap products for situational awareness, inspection priority, or damage assessment of their own infrastructure or building portfolios. The success of ShakeCast to date and its broad, critical-user base mandates improved software usability and functionality, including improved engineering-based damage and loss functions. In order to make the software more accessible to novice users—while still utilizing advanced users’ technical and engineering background—we have developed a “ShakeCast Workbook”, a well documented, Excel spreadsheet-based user interface that allows users to input notification and inventory data and export XML files requisite for operating the ShakeCast system. Users will be able to select structure based on a minimum set of user-specified facility (building location, size, height, use, construction age, etc.). “Expert” users will be able to import user-modified structural response properties into facility inventory associated with the HAZUS Advanced Engineering Building Modules (AEBM). The goal of the ShakeCast system is to provide simplified real-time potential impact and inspection metrics (i.e., green, yellow, orange and red priority ratings) to allow users to institute customized earthquake response protocols. Previously, fragilities were approximated using individual ShakeMap intensity measures (IMs, specifically PGA and 0.3 and 1s spectral accelerations) for each facility but we are now performing capacity-spectrum damage state calculations using a more robust characterization of spectral deamnd.We are also developing methods for the direct import of ShakeMap’s multi-period spectra in lieu of the assumed three-domain design spectrum (at 0.3s for

  13. Criticality Model

    International Nuclear Information System (INIS)

    Alsaed, A.

    2004-01-01

    The ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003) presents the methodology for evaluating potential criticality situations in the monitored geologic repository. As stated in the referenced Topical Report, the detailed methodology for performing the disposal criticality analyses will be documented in model reports. Many of the models developed in support of the Topical Report differ from the definition of models as given in the Office of Civilian Radioactive Waste Management procedure AP-SIII.10Q, ''Models'', in that they are procedural, rather than mathematical. These model reports document the detailed methodology necessary to implement the approach presented in the Disposal Criticality Analysis Methodology Topical Report and provide calculations utilizing the methodology. Thus, the governing procedure for this type of report is AP-3.12Q, ''Design Calculations and Analyses''. The ''Criticality Model'' is of this latter type, providing a process evaluating the criticality potential of in-package and external configurations. The purpose of this analysis is to layout the process for calculating the criticality potential for various in-package and external configurations and to calculate lower-bound tolerance limit (LBTL) values and determine range of applicability (ROA) parameters. The LBTL calculations and the ROA determinations are performed using selected benchmark experiments that are applicable to various waste forms and various in-package and external configurations. The waste forms considered in this calculation are pressurized water reactor (PWR), boiling water reactor (BWR), Fast Flux Test Facility (FFTF), Training Research Isotope General Atomic (TRIGA), Enrico Fermi, Shippingport pressurized water reactor, Shippingport light water breeder reactor (LWBR), N-Reactor, Melt and Dilute, and Fort Saint Vrain Reactor spent nuclear fuel (SNF). The scope of this analysis is to document the criticality computational method. The criticality

  14. Criticality Calculations for a Typical Nuclear Fuel Fabrication Plant with Low Enriched Uranium

    International Nuclear Information System (INIS)

    Elsayed, Hade; Nagy, Mohamed; Agamy, Said; Shaat, Mohmaed

    2013-01-01

    The operations with the fissile materials such as U 235 introduce the risk of a criticality accident that may be lethal to nearby personnel and can lead the facility to shutdown. Therefore, the prevention of a nuclear criticality accident should play a major role in the design of a nuclear facility. The objectives of criticality safety are to prevent a self-sustained nuclear chain reaction and to minimize the consequences. Sixty criticality accidents were occurred in the world. These are accidents divided into two categories, 22 accidents occurred in process facilities and 38 accidents occurred during critical experiments or operations with research reactor. About 21 criticality accidents including Japan Nuclear Fuel Conversion Co. (JCO) accident took place with fuel solution or slurry and only one accident occurred with metal fuel. In this study the nuclear criticality calculations have been performed for a typical nuclear fuel fabrication plant producing nuclear fuel elements for nuclear research reactors with low enriched uranium up to 20%. The calculations were performed for both normal and abnormal operation conditions. The effective multiplication factor (k eff ) during the nuclear fuel fabrication process (Uranium hexafluoride - Ammonium Diuranate conversion process) was determined. Several accident scenarios were postulated and the criticalities of these accidents were evaluated. The computer code MCNP-4B which based on Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations Monte Carlo method was used to calculate neutron multiplication factor. The criticality calculations were performed for the cases of, change of moderator to fuel ratio, solution density and concentration of the solute in order to prevent or mitigate criticality accidents during the nuclear fuel fabrication process. The calculation results are analyzed and discussed

  15. Accountability control system in plutonium fuel facility

    International Nuclear Information System (INIS)

    Naruki, Kaoru; Aoki, Minoru; Mizuno, Ohichi; Mishima, Tsuyoshi

    1979-01-01

    More than 30 tons of plutonium-uranium mixed-oxide fuel have been manufactured at the Plutonium Facility in PNC for JOYO, FUGEN and DCA (Deuterium Critical Assembly) and for the purpose of irradiation tests. This report reviews the nuclear material accountability control system adopted in the Plutonium Facility. Initially, the main objective of the system was the criticality control of fissible materials at various stages of fuel manufacturing. The first part of this report describes the functions and the structure of the control system. A flow chart is provided to show the various stages of material flow and their associated computer files. The system is composed of the following three sub-systems: procedures of nuclear material transfer; PIT (Physical Inventory Taking); data retrieval, report preparation and file maintenance. OMR (Optical Mark Reader) sheets are used to record the nuclear material transfer. The MUF (Materials Unaccounted For) are evaluated by PIT every three months through computer processing based on the OMR sheets. The MUF ratio of Pu handled in the facility every year from 1966 to 1977 are presented by a curve, indicating that the MUF ratio was kept well under 0.5% for every project (JOYO, FUGEN, and DCA). As for the Pu safeguards, the MBA (Material Balance Area) and the KMP (Key Measurement Point) in the facility of PNC are illustrated. The general idea of the projected PINC (Plutonium Inventory Control) system in PNC is also shortly explained. (Aoki, K.)

  16. Nuclear criticality experiments from 1943 to 1978: an annotated bibliography. Volume 1. Main listing

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, B.L.; Wilcox, T.P.; Hampel, V.E.

    1979-04-24

    The bibliography contains 1067 citations from the literature of critical and near-critical nuclear experiments. It provides an up-to-date index to reports containing useful data for many types of criticality studies. Most of the reports can provide specifications for relatively simple critical configurations necessary for validating nuclear constants and calculational techniques. The reports of more than 1143 experimentors at 38 international facilities since 1943 are cross-referenced. The collection contains the prototypes of many different designs of nuclear reactors and studies performed to insure the safe use of fissile materials in chemical processing plants, storage facilities, and transportation containers. The bibliography has three volumes. Volume 1 contains the main listing of citations with abstracts. Volume 2 is a set of indexes organized by report number, publication date, experimental facility, and author name. Volume 3 provides a subject index, concorded on the significant keyphrases derived from titles, and an index of keyterms derived from titles, and an index of keyterms extracted from titles and abstracts. The bibliography was printed by computer as a selection from a computerized system at Lawrence Livermore Laboratory contaning information and data on criticality experiments.

  17. Nuclear criticality experiments from 1943 to 1978. An annotated bibliography: Volume 1, main listing

    International Nuclear Information System (INIS)

    Koponen, B.L.; Wilcox, T.P.; Hampel, V.E.

    1979-05-01

    This report only describes the bibliography which contains 1067 citations from the literature of critical and near-critical nuclear experiments. The bibliography provides an up-to-date index to reports containing useful data for many types of criticality studies. Most of the reports can provide specifications for relatively simple critical configurations necessary for validating nuclear constants and calculational techniques. The reports of more than 1143 experimentors at 38 international facilities since 1943 are cross-referenced. This collection contains the prototypes of many different designs of nuclear reactors and studies performed to ensure the safe use of fissile materials in chemical processing plants, storage facilities, and transportation containers. The bibliography has three volumes. Volume 1 contains the main listing of citations with abstracts. Volume 2 is a set of indexes organized by report number, publication date, experimental facility, and author name. Volume 3 provides a subject index, concorded on the significant keyphrases derived from titles, and an index of key terms extracted from titles and abstracts. The bibliography was printed by computer as a selection from a computerized system at Lawrence Livermore Laboratory containing information and data on criticality experiments

  18. Primer for criticality calculations with DANTSYS

    International Nuclear Information System (INIS)

    Busch, R.D.

    1996-01-01

    With the closure of many experimental facilities, the nuclear criticality safety analyst is increasingly required to rely on computer calculations to identify safe limits for the handling and storage of fissile materials. However, in many cases, the analyst has little experience with the specific codes available at his or her facility. Typically, two types of codes are available: deterministic codes such as ANISN or DANTSYS that solve an approximate model exactly and Monte Carlo Codes such as KENO or MCNP that solve an exact model approximately. Often, the analyst feels that the deterministic codes are too simple and will not provide the necessary information, so most modeling uses Monte Carlo methods. This sometimes means that hours of effort are expended to produce results available in minutes from deterministic codes. A substantial amount of reliable information on nuclear systems can be obtained using deterministic methods if the user understands their limitations. To guide criticality specialists in this area, the Nuclear Criticality Safety Group at the University of New Mexico in cooperation with the Radiation Transport Group at Los Alamos National Laboratory has designed a primer to help the analyst understand and use the DANTSYS deterministic transport code for nuclear criticality safety analyses. (DANTSYS is the name of a suite of codes that users more commonly know as ONEDANT, TWODANT, TWOHEX, and THREEDANT.) It assumes a college education in a technical field, but there is no assumption of familiarity with neutronics codes in general or with DANTSYS in particular. The primer is designed to teach by example, with each example illustrating two or three DANTSYS features useful in criticality analyses

  19. Design basis earthquakes for critical industrial facilities and their characteristics, and the Southern Hyogo prefecture earthquake, 17 January 1995

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, Heki

    1998-12-01

    This paper deals with how to establish the concept of the design basis earthquake (DBE) for critical industrial facilities such as nuclear power plants in consideration of disasters such as the Southern Hyogo prefecture earthquake, the so-called Kobe earthquake in 1995. The author once discussed various DBEs at the 7th World Conference on Earthquake Engineering. At that time, the author assumed that the strongest effective PGA would be 0.7 G, and compared the values of accelerations of a structure obtained by various codes in Japan and other countries. The maximum PGA observed by an instrument at the Southern Hyogo prefecture earthquake in 1995 exceeded the previous assumption of the author, even though the results of the previous paper had been pessimistic. According to the experience of the Kobe event, the author will point out the necessity of the third earthquake S{sub s} adding to S{sub 1} and S{sub 2} of previous DBEs.

  20. Nuclear criticality safety: 3-day training course

    International Nuclear Information System (INIS)

    Schlesser, J.A.

    1993-06-01

    The open-quotes 3-Day Training Courseclose quotes is an intensive course in criticality safety consisting of lectures and laboratory sessions, including active student participation in actual critical experiments, a visit to a plutonium processing facility, and in-depth discussions on safety philosophy. The program is directed toward personnel who currently have criticality safety responsibilities in the capacity of supervisory staff and/or line management. This compilation of notes is presented as a source reference for the criticality safety course. It represents the contributions of many people, particularly Tom McLaughlin, the course's primary instructor. It should be noted that when chapters were extracted, an attempt was made to maintain footnotes and references as originally written. Photographs and illustrations are numbered sequentially

  1. Benchmark assemblies of the Los Alamos critical assemblies facility

    International Nuclear Information System (INIS)

    Dowdy, E.J.

    1986-01-01

    Several critical assemblies of precisely known materials composition and easily calculated and reproducible geometries have been constructed at the Los Alamos National Laboratory. Some of these machines, notably Jezebel, Flattop, Big Ten, and Godiva, have been used as benchmark assemblies for the comparison of the results of experimental measurements and computation of certain nuclear reaction parameters. These experiments are used to validate both the input nuclear data and the computational methods. The machines and the applications of these machines for integral nuclear data checks are described. (author)

  2. Benchmark assemblies of the Los Alamos Critical Assemblies Facility

    International Nuclear Information System (INIS)

    Dowdy, E.J.

    1985-01-01

    Several critical assemblies of precisely known materials composition and easily calculated and reproducible geometries have been constructed at the Los Alamos National Laboratory. Some of these machines, notably Jezebel, Flattop, Big Ten, and Godiva, have been used as benchmark assemblies for the comparison of the results of experimental measurements and computation of certain nuclear reaction parameters. These experiments are used to validate both the input nuclear data and the computational methods. The machines and the applications of these machines for integral nuclear data checks are described

  3. Benchmark assemblies of the Los Alamos critical assemblies facility

    International Nuclear Information System (INIS)

    Dowdy, E.J.

    1985-01-01

    Several critical assemblies of precisely known materials composition and easily calculated and reproducible geometries have been constructed at the Los Alamos National Laboratory. Some of these machines, notably Jezebel, Flattop, Big Ten, and Godiva, have been used as benchmark assemblies for the comparison of the results of experimental measurements and computation of certain nuclear reaction parameters. These experiments are used to validate both the input nuclear data and the computational methods. The machines and the applications of these machines for integral nuclear data checks are described

  4. Ergonomics and simulation-based approach in improving facility layout

    Science.gov (United States)

    Abad, Jocelyn D.

    2018-02-01

    The use of the simulation-based technique in facility layout has been a choice in the industry due to its convenience and efficient generation of results. Nevertheless, the solutions generated are not capable of addressing delays due to worker's health and safety which significantly impact overall operational efficiency. It is, therefore, critical to incorporate ergonomics in facility design. In this study, workstation analysis was incorporated into Promodel simulation to improve the facility layout of a garment manufacturing. To test the effectiveness of the method, existing and improved facility designs were measured using comprehensive risk level, efficiency, and productivity. Results indicated that the improved facility layout generated a decrease in comprehensive risk level and rapid upper limb assessment score; an increase of 78% in efficiency and 194% increase in productivity compared to existing design and thus proved that the approach is effective in attaining overall facility design improvement.

  5. 340 Waste handling Facility Hazard Categorization and Safety Analysis

    International Nuclear Information System (INIS)

    Rodovsky, T.J.

    2010-01-01

    decommissioning or pumping of radioactive materials from the vault tanks is prohibited. The Criticality Safety Program, HNF-7098, currently classifies an Exempt facility as one that is less than HC 3 per DOE STD 1027-92, Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports, therefore the 340 Facility is classified as Exempt. Exempt facilities are not required to comply with most of the requirements specified in the Criticality Safety Program. The exceptions, with regards to the 340 Facility, include maintaining an accounting of the facility source term to ensure that the facility hazard category is not changed and ensuring that fissionable materials are appropriately labeled.

  6. Assessment of the facilities on Jackass Flats and other Nevada Test Site facilities for the new nuclear rocket program

    International Nuclear Information System (INIS)

    Chandler, G.; Collins, D.; Dye, K.; Eberhart, C.; Hynes, M.; Kovach, R.; Ortiz, R.; Perea, J.; Sherman, D.

    1992-01-01

    Recent NASA/DOE studies for the Space Exploration Initiative have demonstrated a critical need for the ground-based testing of nuclear rocket engines. Experience in the ROVER/NERVA Program, experience in the Nuclear Weapons Testing Program, and involvement in the new nuclear rocket program has motivated our detailed assessment of the facilities used for the ROVER/NERVA Program and other facilities located at the Nevada Test Site (NTS). The ROVER/NERVA facilities are located in the Nevada Research L, Development Area (NRDA) on Jackass Flats at NTS, approximately 85 miles northwest of Las Vegas. To guide our assessment of facilities for an engine testing program we have defined a program goal, scope, and process. To execute this program scope and process will require ten facilities. We considered the use of all relevant facilities at NTS including existing and new tunnels as well as the facilities at NRDA. Aside from the facilities located at remote sites and the inter-site transportation system, all of the required facilities are available at NRDA. In particular we have studied the refurbishment of E-MAD, ETS-1, R-MAD, and the interconnecting railroad. The total cost for such a refurbishment we estimate to be about $253M which includes additional contractor fees related to indirect, construction management, profit, contingency, and management reserves. This figure also includes the cost of the required NEPA, safety, and security documentation

  7. Accomplishment of 10-year research in NUCEF and future development. Criticality safety research

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori

    2005-01-01

    Since 1995, static and transient critical experiments on low enriched uranyl nitrate solution have been performed using two solution type criticality facilities, STACY and TRACY constructed in NUCEF. The obtained fundamental and systematic data on aqueous solution were used to validate the criticality safety calculation codes and to develop the transient analyses codes for criticality accident evaluation. This paper describes the outline of the criticality safety research conducted in NUCEF. (author)

  8. Benchmark calculations for critical experiments at FKBN-M facility with uranium-plutonium-polyethylene systems using JENDL-3.2 and MVP Monte-Carlo code

    International Nuclear Information System (INIS)

    Obara, Toru; Morozov, A.G.; Kevrolev, V.V.; Kuznetsov, V.V.; Treschalin, S.A.; Lukin, A.V.; Terekhin, V.A.; Sokolov, Yu.A.; Kravchenko, V.G.

    2000-01-01

    Benchmark calculations were performed for critical experiments at FKBN-M facility in RFNC-VNIITF, Russia using JENDL-3.2 nuclear data library and continuous energy Monte-Carlo code MVP. The fissile materials were high-enriched uranium and plutonium. Polyethylene was used as moderator. The neutron spectrum was changed by changing the geometry. Calculation results by MVP showed some errors. Discussion was made by reaction rates and η values obtained by MVP. It showed the possibility that cross sections of U-235 had different trend of error in fast and thermal energy region respectively. It also showed the possibility of some error of cross section of Pu-239 in high energy region. (author)

  9. ALARP considerations in criticality safety assessments

    International Nuclear Information System (INIS)

    Bowden, Russell L.; Barnes, Andrew; Thorne, Peter R.; Venner, Jack

    2003-01-01

    Demonstrating that the risk to the public and workers is As Low As Reasonably Practicable (ALARP) is a fundamental requirement of safety cases for nuclear facilities in the United Kingdom. This is embodied in the Safety Assessment Principles (SAPs) published by the Regulator, the essence of which is incorporated within the safety assessment processes of the various nuclear site licensees. The concept of ALARP within criticality safety assessments has taken some time to establish in the United Kingdom. In principle, the licensee is obliged to search for a deterministic criticality safety solution, such as safe geometry vessels and passive control features, rather than placing reliance on active measurement devices and plant administrative controls. This paper presents a consideration of some ALARP issues in relation to the development of criticality safety cases. The paper utilises some idealised examples covering a range of issues facing the criticality safety assessor, including new plant design, operational plant and decommissioning activities. These examples are used to outline the elements of the criticality safety cases and present a discussion of ALARP in the context of criticality safety assessments. (author)

  10. Study of physical resistance of the disposal facility for accidental artificial event in LLW disposal facility

    International Nuclear Information System (INIS)

    Ogawa, Suihei; Irie, Masaaki; Uchida, Masahiro

    2013-11-01

    This report refer to results of examine what follows for structural stability evaluation for the LLW disposal facility in depth over general human activity in underground. Study of physically resistance on the facility for accidental artificial event, namely tunneling an operation facing the disposal facility in future. Physically resistance to excavation of tunneling etc. in disposal facility is studied based on supposing of Tunnel Boring Machine as an excavator, paying attention to reinforcement bar in concrete and steel plate of waste package, as feature of strength in these material differs from rock strength. And it is examined not only resistibility on excavation but also about hard situations of excavation in tunneling works, and namely give thorough consideration to critical quantity of cutting to reinforcement bar and steel plate that could keep resistibility on excavation based on tunneling velocity and limits time furthermore. It requests necessity of evaluation in consider with metal corrosion that status alteration on disposal facility is considered with on timescale. Period of keep on the physically resistance is estimated by velocity of metal corrosion consequently. The physically resistance is kept until metal corrosion reach remaining its material, giving a limits of the physically resistance on inside of facility. Main point of physically resistance in the report will be made the good use of a practice to physically resistance evaluation of in safety assessment. (author)

  11. Criticality Safety Information Resource Center Web portal: www.csirc.net

    International Nuclear Information System (INIS)

    Harmon, C.D. II; Jones, T.

    2000-01-01

    The Nuclear Criticality Safety Group (ESH-6) at Los Alamos National Laboratory (LANL) is in the process of collecting and archiving historical and technical information related to nuclear criticality safety from LANL and other facilities. In an ongoing effort, this information is being made available via the Criticality Safety Information Resource Center (CSIRC) web site, which is hosted and maintained by ESH-6 staff. Recently, the CSIRC Web site was recreated as a Web portal that provides the criticality safety community with much more than just archived data

  12. Critical mass experiment using U-235 foils and lucite plates

    International Nuclear Information System (INIS)

    Sanchez, R.; Butterfield, K.; Kimpland, R.; Jaegers, P.

    1998-01-01

    The main objective of this experiment was to show how the multiplication of the system increases as moderated material is placed between highly enriched uranium foils. In addition, this experiment served to demonstrate the hand-stacking techniques, and approach to criticality by remote operation. This experiment was designed by Tom McLaughlin in the mid seventies as part of the criticality safety course that is taught at Los Alamos Critical Experiment Facility (LACEF). The W-U-235 ratio for this experiment was 215 which is where the minimum critical mass for this configuration occurs

  13. Critical mass experiment using 235U foils and lucite plates

    International Nuclear Information System (INIS)

    Sanchez, R.; Butterfield, K.; Kimpland, R.; Jaegers, P.

    1998-01-01

    This experiment demonstrated how the neutron multiplication of a system increases as moderated material is placed between highly enriched uranium foils. In addition, this experiment served to demonstrate the hand-stacking technique and approach to criticality be remote operation. This experiment was designed by McLaughlin in the mid-seventies as part of the criticality safety course that is taught at the Los Alamos Critical Experiments Facility. The H/ 235 U ratio for this experiment was 215, which is the ratio at which the minimum critical mass for this configuration occurs

  14. Safeguarding uranium enrichment facilities. Review and analysis of the status of safeguards technology for uranium enrichment facilities

    International Nuclear Information System (INIS)

    1977-09-01

    The objective of this paper is to examine critically the diversion potential at uranium enrichment facilities and to outline a basic safeguards strategy which counters all identified hazards as completely as possible yet with a minimum of non-essential redundancy. Where existing technology does not appear to be adequate for effective safeguards, the limitations are examined, and suggestions for further R and D effort are made. Parts of this report are generally applicable to all currently known enrichment processes, while other parts are specifically directed toward facilities based on the gas centrifuge process. It is hoped that additional sections discussing a safeguards strategy for gas diffusion facilities can be added later. It should be emphasized that this is a technical report, and does not reflect any legal positions. The safeguards strategy and subsequent inspection procedures are intended as guidelines, not as negotiating positions

  15. Application of the “best representativity” method to a PWR fuel calculation using the critical experiments at the Toshiba NCA facility

    International Nuclear Information System (INIS)

    Umano, Takuya; Yoshioka, Kenichi; Obara, Toru

    2015-01-01

    Highlights: • Calculation procedures are easier than those of the cross-section adjustment method. • In addition, few cases of experimental results, two or three cases for example, can be well managed with the method. • Different from the bias factor method, a representativity factor is simultaneously obtained to know the whole quality of utilized experiments. • After easier process of calculations, it is possible to obtain a correction value of a particular physical property with the method. • Our proposed method was considered to be applicable to the correction of the infinite neutron multiplication factor in LWR studies. - Abstract: To judge the applicability of a critical experiment, it is necessary to confirm the similarities of the experiment with actual reactor conditions or equipment. The concept of the “representativity factor” has been well adopted since the late 1970’s, particularly for fast breeder reactors (FBRs) and future reactor studies. In our previous study, we extended this concept to the design of a light water reactor (LWR) system, and derived mathematical formulas for a new numerical evaluation method to correct a physical property of a target system. This method is different from the cross-section adjustment method and the bias factor method. For the first qualification of the method, sample calculations were carried out to correct the effective neutron multiplication factor through critical experiments at the Toshiba Nuclear Critical Assembly (NCA) facility. We also compared the result with that of the Product of Exponentiated experimental values method (PE method) of the extended bias factor methods. A good agreement was observed. The purpose of this study was to demonstrate the applicability of the method to the infinite neutron multiplication factor. Using the method and three kinds of critical experiments of NCA, calculations were performed to correct the infinite neutron multiplication factor of a pressurized water

  16. Proceedings of the Nuclear Criticality Technology and Safety Project Workshop

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, R.G. [comp.

    1994-01-01

    This report is the proceedings of the annual Nuclear Criticality Technology and Safety Project (NCTSP) Workshop held in Monterey, California, on April 16--28, 1993. The NCTSP was sponsored by the Department of Energy and organized by the Los Alamos Critical Experiments Facility. The report is divided into six sections reflecting the sessions outlined on the workshop agenda.

  17. Proceedings of the Nuclear Criticality Technology and Safety Project Workshop

    International Nuclear Information System (INIS)

    Sanchez, R.G.

    1994-01-01

    This report is the proceedings of the annual Nuclear Criticality Technology and Safety Project (NCTSP) Workshop held in Monterey, California, on April 16--28, 1993. The NCTSP was sponsored by the Department of Energy and organized by the Los Alamos Critical Experiments Facility. The report is divided into six sections reflecting the sessions outlined on the workshop agenda

  18. The NMT-5 criticality database

    International Nuclear Information System (INIS)

    Cort, B.; Perkins, B.; Cort, G.

    1995-05-01

    The NMT-5 Criticality Database maintains criticality-related data and documentation to ensure the safety of workers handling special nuclear materials at the Plutonium Facility (TA-55) at Los Alamos National Laboratory. The database contains pertinent criticality safety limit information for more than 150 separate locations at which special nuclear materials are handled. Written in 4th Dimension for the Macintosh, it facilitates the production of signs for posting at these areas, tracks the history of postings and related authorizing documentation, and generates in Microsoft Word a current, comprehensive representation of all signs and supporting documentation, such as standard operating procedures and signature approvals. It facilitates the auditing process and is crucial to full and effective compliance with Department of Energy regulations. It has been recommended for installation throughout the Nuclear Materials Technology Division at Los Alamos

  19. Organization and staffing of the regulatory body for nuclear facilities. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this safety guide is to provide recommendations for national authorities on the appropriate management system, organization and staffing for the regulatory body responsible for the regulation of nuclear facilities in order to achieve compliance with the applicable safety requirements. This safety guide covers the organization and staffing in relation to nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And radioactive waste management facilities such as treatment, storage and disposal facilities. This safety guide also covers issues related to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation

  20. Preliminary Design of the AEGIS Test Facility

    CERN Document Server

    Dassa, Luca; Cambiaghi, Danilo

    2010-01-01

    The AEGIS experiment is expected to be installed at the CERN Antiproton Decelerator in a very close future, since the main goal of the AEGIS experiment is the measurement of gravity impact on antihydrogen, which will be produced on the purpose. Antihydrogen production implies very challenging environmental conditions: at the heart of the AEGIS facility 50 mK temperature, 1e-12 mbar pressure and a 1 T magnetic field are required. Interfacing extreme cryogenics with ultra high vacuum will affect very strongly the design of the whole facility, requiring a very careful mechanical design. This paper presents an overview of the actual design of the AEGIS experimental facility, paying special care to mechanical aspects. Each subsystem of the facility – ranging from the positron source to the recombination region and the measurement region – will be shortly described. The ultra cold region, which is the most critical with respect to the antihydrogen formation, will be dealt in detail. The assembly procedures will...

  1. Criticality safety evaluations - a open-quotes stalking horseclose quotes for integrated safety assessment

    International Nuclear Information System (INIS)

    Williams, R.A.

    1995-01-01

    The Columbia Fuel Fabrication Facility of the Westinghouse Commercial Nuclear Fuel Division manufactures low-enriched uranium fuel and associated components for use in commercial pressurized water power reactors. To support development of a comprehensive integrated safety assessment (ISA) for the facility, as well as to address increasing U.S. Nuclear Regulatory Commission (NRC) expectations regarding such a facility's criticality safety assessments, a project is under way to complete criticality safety evaluations (CSEs) of all plant systems used in processing nuclear materials. Each CSE is made up of seven sections, prepared by a multidisciplinary team of process engineers, systems engineers, safety engineers, maintenance representatives, and operators. This paper provides a cursory outline of the type of information presented in a CSE

  2. Criticality safety (prospect of study in NUCEF)

    International Nuclear Information System (INIS)

    Itagaki, Masafumi

    1996-01-01

    Experimental studies of criticality safety are under way using STACY and TRACY in NUCEF. Collection of fundamental data on criticality in a solution system is undergoing with STACY to confirm that the likelihood of criticality safety in the system constructed on the assumption of apparatuses in a reprocessing plant is enough large. Whereas some experiments simulating criticality accidents in a reprocessing plant using TRACY were designed to investigate the behaviors of fuel solution and radioactive matters in order to clarify whether it is possible to safely shut them in the facility even if a critical accident occurs. Both STACY and TRACY reached the criticality in 1995. Up to now a series of criticality experiments have been done using STACY with a core tank φ60 cm and the first periodical examination is now under way. On the other hand, we have a plan using TRACY to investigate the behaviors of nuclear heat solution at a criticality accident, and the releasing, transfer and deposition of radioactive materials. After reaching the criticality for the first, the performance verification test has been conducted. The full-scale study using TRACY is planned to begin in the second half of 1996. (M.N.)

  3. Significant incidents in nuclear fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-03-01

    In contrast to nuclear power plants, events in nuclear fuel cycle facilities are not well documented. The INES database covers all the nuclear fuel cycle facilities; however, it was developed in the early 1990s and does not contain information on events prior to that. The purpose of the present report is to collect significant events and analyze them in order to give a safety related overview of nuclear fuel cycle facilities. Significant incidents were selected using the following criteria: release of radioactive material or exposure to radiation; degradation of items important to safety; and deficiencies in design, quality assurance, etc. which include criticality incidents, fire, explosion, radioactive release and contamination. This report includes an explanation, where possible, of root causes, lessons learned and action taken. 4 refs, 4 tabs.

  4. Significant incidents in nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    1996-03-01

    In contrast to nuclear power plants, events in nuclear fuel cycle facilities are not well documented. The INES database covers all the nuclear fuel cycle facilities; however, it was developed in the early 1990s and does not contain information on events prior to that. The purpose of the present report is to collect significant events and analyze them in order to give a safety related overview of nuclear fuel cycle facilities. Significant incidents were selected using the following criteria: release of radioactive material or exposure to radiation; degradation of items important to safety; and deficiencies in design, quality assurance, etc. which include criticality incidents, fire, explosion, radioactive release and contamination. This report includes an explanation, where possible, of root causes, lessons learned and action taken. 4 refs, 4 tabs

  5. Is the closest facility the one actually used? An assessment of travel time estimation based on mammography facilities.

    Science.gov (United States)

    Alford-Teaster, Jennifer; Lange, Jane M; Hubbard, Rebecca A; Lee, Christoph I; Haas, Jennifer S; Shi, Xun; Carlos, Heather A; Henderson, Louise; Hill, Deirdre; Tosteson, Anna N A; Onega, Tracy

    2016-02-18

    Characterizing geographic access depends on a broad range of methods available to researchers and the healthcare context to which the method is applied. Globally, travel time is one frequently used measure of geographic access with known limitations associated with data availability. Specifically, due to lack of available utilization data, many travel time studies assume that patients use the closest facility. To examine this assumption, an example using mammography screening data, which is considered a geographically abundant health care service in the United States, is explored. This work makes an important methodological contribution to measuring access--which is a critical component of health care planning and equity almost everywhere. We analyzed one mammogram from each of 646,553 women participating in the US based Breast Cancer Surveillance Consortium for years 2005-2012. We geocoded each record to street level address data in order to calculate travel time to the closest and to the actually used mammography facility. Travel time between the closest and the actual facility used was explored by woman-level and facility characteristics. Only 35% of women in the study population used their closest facility, but nearly three-quarters of women not using their closest facility used a facility within 5 min of the closest facility. Individuals that by-passed the closest facility tended to live in an urban core, within higher income neighborhoods, or in areas where the average travel times to work was longer. Those living in small towns or isolated rural areas had longer closer and actual median drive times. Since the majority of US women accessed a facility within a few minutes of their closest facility this suggests that distance to the closest facility may serve as an adequate proxy for utilization studies of geographically abundant services like mammography in areas where the transportation networks are well established.

  6. Status of safety in nuclear facilities - 2012. AREVA General Inspectorate Annual report

    International Nuclear Information System (INIS)

    2013-05-01

    After a message from the Areva's Chief Executive Officer and a message from the senior Vice President of safety, health, security, sustainable development, a text by the inspector general comments the key safety results (events, dose levels, radiological impacts), the inspection findings, the areas of vigilance (relationship with the ASN, the management of the criticality risk, and facility compliance), some significant topics after the Fukushima accident. Then this report addresses the status of nuclear safety and radiation protection in the group's facilities and operations. It more specifically addresses the context and findings (lessons learned from the inspections, operating experience from event, employee radiation monitoring, environmental monitoring), crosscutting processes (safety management, controlling facility compliance, subcontractor guidance and management, crisis management), specific risks (criticality risk, fire hazards, transportation safety, radioactive waste management, pollution prevention, liability mitigation and dismantling), and areas for improvement and outlook

  7. Proceedings of the first annual Nuclear Criticality Safety Technology Project

    International Nuclear Information System (INIS)

    Rutherford, D.A.

    1994-09-01

    This document represents the published proceedings of the first annual Nuclear Criticality Safety Technology Project (NCSTP) Workshop, which took place May 12--14, 1992, in Gaithersburg, Md. The conference consisted of four sessions, each dealing with a specific aspect of nuclear criticality safety issues. The session titles were ''Criticality Code Development, Usage, and Validation,'' ''Experimental Needs, Facilities, and Measurements,'' ''Regulation, Compliance, and Their Effects on Nuclear Criticality Technology and Safety,'' and ''The Nuclear Criticality Community Response to the USDOE Regulations and Compliance Directives.'' The conference also sponsored a Working Group session, a report of the NCSTP Working Group is also presented. Individual papers have been cataloged separately

  8. Automated approach to nuclear facility safeguards effectiveness evaluation

    International Nuclear Information System (INIS)

    1977-01-01

    Concern over the security of nuclear facilities has generated a need for a reliable, time efficient, and easily applied method of evaluating the effectiveness of safeguards systems. Such an evaluation technique could be used (1) by the Nuclear Regulatory Commission to evaluate a licensee's proposal, (2) to assess the security status of a system, or (3) to design and/or upgrade nuclear facilities. The technique should be capable of starting with basic information, such as the facility layout and performance parameters for physical protection components, and analyzing that information so that a reliable overall facility evaluation is obtained. Responding to this expressed need, an automated approach to facility safeguards effectiveness evaluation has been developed. This procedure consists of a collection of functional modules for facility characterization, critical path generation, and path evaluation combined into a continuous stream of operations. The technique has been implemented on an interactive computer-timesharing system and makes use of computer graphics for the handling and presentation of information. Using this technique a thorough facility evaluation can be made by systematically varying parameters that characterize the physical protection components of a facility according to changes in perceived adversary attributes and strategy, environmental conditions, and site status

  9. Rethinking regulations for disposal criticality

    International Nuclear Information System (INIS)

    Scott, M.; Doering, T.

    1997-01-01

    This paper provides the basis for the position that the current U.S. Nuclear Regulatory Commission (NRC) criticality regulation is in need of revision to address problems in implementing it for the postclosure period in a geologic high-level waste repository. The authors believe that the applicant for such a facility should be able to demonstrate that postulated postclosure criticality events will not cause unacceptable risk of deleterious effects on public health and safety. In addition, the applicant should be expected to take practical and feasible measures to reduce the probability of a criticality occurring, even if (as expected) the consequences of such a criticality for repository performance and public health and safety would be negligible. This approach, while recognizing the probabilistic nature of analyses of events and conditions in the distant future, is also arguably consistent with the defense in depth concept that has been successfully applied to nuclear reactor regulation. The authors believe regulations for postclosure criticality control should support this dual approach, rather than require a deterministic prohibition of criticality as does the current rule. The existing rule seems appropriate for the preclosure period, as long as it is clearly specified to apply only to that period

  10. Geographic Hotspots of Critical National Infrastructure.

    Science.gov (United States)

    Thacker, Scott; Barr, Stuart; Pant, Raghav; Hall, Jim W; Alderson, David

    2017-12-01

    Failure of critical national infrastructures can result in major disruptions to society and the economy. Understanding the criticality of individual assets and the geographic areas in which they are located is essential for targeting investments to reduce risks and enhance system resilience. Within this study we provide new insights into the criticality of real-life critical infrastructure networks by integrating high-resolution data on infrastructure location, connectivity, interdependence, and usage. We propose a metric of infrastructure criticality in terms of the number of users who may be directly or indirectly disrupted by the failure of physically interdependent infrastructures. Kernel density estimation is used to integrate spatially discrete criticality values associated with individual infrastructure assets, producing a continuous surface from which statistically significant infrastructure criticality hotspots are identified. We develop a comprehensive and unique national-scale demonstration for England and Wales that utilizes previously unavailable data from the energy, transport, water, waste, and digital communications sectors. The testing of 200,000 failure scenarios identifies that hotspots are typically located around the periphery of urban areas where there are large facilities upon which many users depend or where several critical infrastructures are concentrated in one location. © 2017 Society for Risk Analysis.

  11. 2011 Annual Criticality Safety Program Performance Summary

    Energy Technology Data Exchange (ETDEWEB)

    Andrea Hoffman

    2011-12-01

    The 2011 review of the INL Criticality Safety Program has determined that the program is robust and effective. The review was prepared for, and fulfills Contract Data Requirements List (CDRL) item H.20, 'Annual Criticality Safety Program performance summary that includes the status of assessments, issues, corrective actions, infractions, requirements management, training, and programmatic support.' This performance summary addresses the status of these important elements of the INL Criticality Safety Program. Assessments - Assessments in 2011 were planned and scheduled. The scheduled assessments included a Criticality Safety Program Effectiveness Review, Criticality Control Area Inspections, a Protection of Controlled Unclassified Information Inspection, an Assessment of Criticality Safety SQA, and this management assessment of the Criticality Safety Program. All of the assessments were completed with the exception of the 'Effectiveness Review' for SSPSF, which was delayed due to emerging work. Although minor issues were identified in the assessments, no issues or combination of issues indicated that the INL Criticality Safety Program was ineffective. The identification of issues demonstrates the importance of an assessment program to the overall health and effectiveness of the INL Criticality Safety Program. Issues and Corrective Actions - There are relatively few criticality safety related issues in the Laboratory ICAMS system. Most were identified by Criticality Safety Program assessments. No issues indicate ineffectiveness in the INL Criticality Safety Program. All of the issues are being worked and there are no imminent criticality concerns. Infractions - There was one criticality safety related violation in 2011. On January 18, 2011, it was discovered that a fuel plate bundle in the Nuclear Materials Inspection and Storage (NMIS) facility exceeded the fissionable mass limit, resulting in a technical safety requirement (TSR) violation. The

  12. Safety analysis of the Los Alamos Critical Experiments Facility. Volume II

    International Nuclear Information System (INIS)

    Paxton, H.C.

    1976-04-01

    The Los Alamos critical assembly layout is designed to facilitate personnel protection by means of remote operation and stringent procedural controls during nonoperating periods. Public protection is straightforward because of the small fission-product inventory, essentially ambient pressures, and moderate temperatures

  13. Plant critical concept

    International Nuclear Information System (INIS)

    O'Regan, P.J.

    1996-01-01

    The objective of the effort summarized in this paper is to support O and M cost reduction efforts by focusing resources on components and processes critical to plant performance. This effort will identify where resources on nonplant critical components and processes can be reduced or eliminated. This method will use a functional assessment as the basis for component-specific evaluations and ranking. This effort consists of two stages conducted in series. The first stage is to deterministically identify that set of plant components that are relevant from a plant performance perspective (i.e., safety, economics, reliability). The second stage probabilistically ranks that set of plant components from an importance perspective, where importance pertains to the particular application and is probabilistically weighted. The results of a pilot study identified that only a relatively small set of components are truly critical from an integrated plant performance perspective. These results are consistent with work being conducted at other nuclear power plants, as well as other commercial facilities. Initial implementation of this effort is estimated to reduce O and M costs on the order of $1 million per year. Subsequent applications are anticipated to increase that savings to $4--$5 million per year

  14. Study on expansion power recovery in CO2 trans-critical cycle

    International Nuclear Information System (INIS)

    Tian Hua; Ma Yitai; Li Minxia; Wang Wei

    2010-01-01

    Due to the ozone depletion potential and global warming potential of CFCs and HCFCs, CO 2 is considered as most potential alternative refrigerant. However, there are serious throttle losses and low system efficiency to CO 2 trans-critical cycle because of its low critical temperature and high operating pressure. The aim of this paper is to design an expander to recover expansion power in CO 2 trans-critical cycle. The theoretical analysis and calculation show that 14-23% of input power of compressor can be recovered. A prototype of rolling piston expander is designed and manufactured and its test facility is established. The test facility consists of CO 2 trans-critical cycle, the expander, the chilling water system and the cooling water system. The experimental results show that the recovery ratio and expander efficiency are affected by rotational speed, inlet temperature and mass flow of expander. The highest recovery ratio can reach to 0.145, which means 14.5% of input power of compressor can be recovered. The expander efficiency can reach to 45%.

  15. Nuclear criticality safety: 300 Area

    International Nuclear Information System (INIS)

    1991-01-01

    This Standard applies to the receipt, processing, storage, and shipment of fissionable material in the 300 Area and in any other facility under the control of the Reactor Materials Project Management Team (PMT). The objective is to establish practices and process conditions for the storage and handling of fissionable material that prevent the accidental assembly of a critical mass and that comply with DOE Orders as well as accepted industry practice

  16. Rockwell International's Critical Mass Laboratory Program at the Rocky Flats Plant

    International Nuclear Information System (INIS)

    McCarthy, J.D.

    1984-01-01

    The primary mission of the laboratory is to provide data in support of plant operations. To fulfill this task, the facility has unique capabilities for perfoming general purpose critical mass experiment. The critical mass laboratory performed over 1000 critical measurements, primarily with plutonium metal and uranium metal, oxide and solution; it worked also on the NRC program (high-enriched uranium measurements). Presently the laboratory staff prepares for a series of critical measurements on a poisoned tube tank; the laboratory intends to continue to pursue basic plant support programs in the future

  17. Proceedings of the nuclear criticality technology safety project

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, R.G. [comp.

    1997-06-01

    This document contains summaries of the most of the papers presented at the 1994 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 10 and 11 at Williamsburg, Va. The meeting was broken up into seven sessions, which covered the following topics: (1) Validation and Application of Calculations; (2) Relevant Experiments for Criticality Safety; (3) Experimental Facilities and Capabilities; (4) Rad-Waste and Weapons Disassembly; (5) Criticality Safety Software and Development; (6) Criticality Safety Studies at Universities; and (7) Training. The minutes and list of participants of the Critical Experiment Needs Identification Workgroup meeting, which was held on May 9 at the same venue, has been included as an appendix. A second appendix contains the names and addresses of all NCTSP meeting participants. Separate abstracts have been indexed to the database for contributions to this proceedings.

  18. Proceedings of the nuclear criticality technology safety project

    International Nuclear Information System (INIS)

    Sanchez, R.G.

    1997-06-01

    This document contains summaries of the most of the papers presented at the 1994 Nuclear Criticality Technology Safety Project (NCTSP) meeting, which was held May 10 and 11 at Williamsburg, Va. The meeting was broken up into seven sessions, which covered the following topics: (1) Validation and Application of Calculations; (2) Relevant Experiments for Criticality Safety; (3) Experimental Facilities and Capabilities; (4) Rad-Waste and Weapons Disassembly; (5) Criticality Safety Software and Development; (6) Criticality Safety Studies at Universities; and (7) Training. The minutes and list of participants of the Critical Experiment Needs Identification Workgroup meeting, which was held on May 9 at the same venue, has been included as an appendix. A second appendix contains the names and addresses of all NCTSP meeting participants. Separate abstracts have been indexed to the database for contributions to this proceedings

  19. Study on uncertainty evaluation system for the safety evaluation of interim spent fuel storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyeon; Shin, Myeong Won; Rhy, Seok Jin; Cho, Dong Keon; Park, Dong Hwan [Kyunghee Univ., Seoul (Korea, Republic of); Cheong, Beom Jin [Minstry of Science and Technology, Gwacheon (Korea, Republic of)

    1998-03-15

    The main objective os to develop a technical standards for the facility operation of the interm, spent fuel storage facility and to develop a draft for the technical criteria to be legislated. The another objective os to define a uncertainty evaluation system for burn up credit application in criticality analysis and to investigate an applicability of this topic for future regulatory activity. Investigate a status of art for the operational criteria of spent fuel interm wet storage. Collect relevant laws, decree, notices and standards related to the operation of storage facility and study on the legislation system. Develop a draft of technical standards and criteria to be legislated. Define an evaluation system for the uncertainty analysis and study on the status of art in the field of criticality safety analysis. Develop an uncertainty evaluation system in criticality analysis with burnup credit and investigate an applicability as well as its benefits of this policy.

  20. Tank waste remediation system nuclear criticality safety program management review

    International Nuclear Information System (INIS)

    BRADY RAAP, M.C.

    1999-01-01

    This document provides the results of an internal management review of the Tank Waste Remediation System (TWRS) criticality safety program, performed in advance of the DOE/RL assessment for closure of the TWRS Nuclear Criticality Safety Issue, March 1994. Resolution of the safety issue was identified as Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-40-12, due September 1999

  1. Fabrication of Separator Demonstration Facility process vessel

    International Nuclear Information System (INIS)

    Oberst, E.F.

    1985-01-01

    The process vessel system is the central element in the Separator Development Facility (SDF). It houses the two major process components, i.e., the laser-beam folding optics and the separators pods. This major subsystem is the critical-path procurement for the SDF project. Details of the vaious parts of the process vessel are given

  2. PND fuel handling decontamination: facilities and techniques

    International Nuclear Information System (INIS)

    Pan, R.Y.

    1996-01-01

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at Ontario Hydro's Pickering Nuclear Division. This paper presents an overview of the set up and techniques used for decontamination in the PND Fuel Handling Maintenance Facility and the effectiveness of each. (author). 1 tab., 9 figs

  3. J-PARC Transmutation Experimental Facility Programme

    International Nuclear Information System (INIS)

    Sasa, T.; Takei, H.; Saito, S.; Obayashi, H.; Nishihara, K.; Sugawara, T.; Iwamoto, H.; Yamaguchi, K.; Tsujimoto, K.; Oigawa, H.

    2015-01-01

    Since the Fukushima accident, nuclear transmutation is considered as an option for waste management. Japan Atomic Energy Agency proposes the transmutation of minor actinides (MA) in accelerator-driven system (ADS) using lead-bismuth eutectic alloy (LBE) as a spallation target and a coolant of subcritical core. To obtain the data required for ADS design, we plan the building of a transmutation experimental facility (TEF) is planned within the J-PARC project. TEF consists of an ADS target test facility (TEF-T), which will be installed 400 MeV-250 kW LBE spallation target for material irradiations, and a transmutation physics experimental facility (TEF-P), which set up a fast critical/subcritical assembly driven by low power proton beam with MA fuel to study ADS neutronics. At TEF-T, various research plans to use emitted neutrons from LBE target are discussed. The paper summarises a road-map to establish the ADS transmuter and latest design activities for TEF construction. (authors)

  4. Cyclotron produced radionuclides: Guidelines for setting up a facility

    International Nuclear Information System (INIS)

    2009-01-01

    Establishment of a cyclotron based radionuclide and radiopharmaceutical production facility is a major undertaking regardless of the scope and size of the facility. Regulatory demands of radiation protection and pharmaceutical manufacturing only add to the need for meticulous attention to a large number of factors in overall planning and successful implementation. Also, a significant commitment of resources, not just during the onset of the project, but also for sustained continuity further adds to the already difficult decision making process. In this publication, all these issues have been addressed and discussed conceptually for the benefit of planners and stakeholders of a new facility. The need for a clear vision and realistic scope of the programme has been repeatedly emphasized throughout this book as this aspect of project planning is absolutely vital for defining and achieving the mission and objectives of the facility. The most critical aspects in conceptualization, planning and subsequent implementation have been discussed in detail and are highlighted below as necessary actions: - Performing a feasibility study which balances wishful thinking with project viability; - Forming a task force composed of the stakeholders to evaluate strategically the various aspects of project planning to ultimately recommend the scope and objectives of the facility, and also to formulate the project plan; - Assessing the financial aspects of the project, including the set-up and operating costs of the facility through the development of business models and cost-benefit analyses; - Designing a facility and layout that encompasses the scope of the project and also takes into consideration the regulatory requirements; - Ensuring the availability of appropriately qualified and trained staff, critical for efficient and high quality operation of the facility; - Applying GMP regulations for the production of radiopharmaceutical products which are consistently safe for human use

  5. The Australian National Proton Facility

    International Nuclear Information System (INIS)

    Jackson, M.; Rozenfeld, A.; Bishop, J.

    2002-01-01

    Full text: Protons have been used in the treatment of cancer since 1954 and over 30,000 patients have been treated around the world. Their precise dose distribution allows the treatment of small tumours in critical locations such as the base of skull and orbit and is an alternative to stereotactic radiotherapy in other sites. With the development of hospital-based systems in the 1990's, common tumours such as prostate, breast and lung cancer can now also be treated using simple techniques. The therapeutic ratio is improved as the dose to the tumour can be increased while sparing normal tissues. The well defined high dose region and low integral dose compared with photon treatments is a particular advantage in children and other situations where long-term survival is expected and when used in combination with chemotherapy. In January 2002, the NSW Health Department initiated a Feasibility Study for an Australian National Proton Facility. This Study will address the complex medical, scientific, engineering, commercial and legal issues required to design and build a proton facility in Australia. The Facility will be mainly designed for patient treatment but will also provide facilities for biological, physical and engineering research. The proposed facility will have a combination of fixed and rotating beams with an energy range of 70-250 MeV. Such a centre will enable the conduct of randomised clinical trials and a comparison with other radiotherapy techniques such as Intensity Modulated Radiation Therapy. Cost-utility comparisons with other medical treatments will also be made and further facilities developed if the expected benefit is confirmed. When patients are not being treated, the beam will be available for commercial and research purposes. This presentation will summarize the progress of the Study and discuss the important issues that need to be resolved before the Facility is approved and constructed

  6. Review of the nuclear safety exercises carried out in French industrial facilities

    International Nuclear Information System (INIS)

    Kissel, Ph.P.; Renard, C.; Meramedjian, H.N.

    1977-01-01

    For several years the Commissariat a l'Energie Atomique (CEA) has been organizing nuclear safety exercises in most nuclear industrial facilities, especially in fuel element fabrication plants, many of which are classified as basic nuclear facilities. The subject and extent of each exercise are decided by mutual agreement between the management of the facility and the CEA officials in charge of Assistance in Protection and Nuclear Safety (APSN). The authors deal with such subjects as criticality accidents (evacuation of facilities, regrouping of personnel, rescue operations etc.) and fire involving large quantities of radioactive material (protection of the environment by spraying water on fumes laden with radioactive aerosols etc.). During these exercises use is made of the resources available with the safety services of the facility, one or more mobile nuclear action teams of the CEA and the appropriate resources within the competence of public authorities, e.g. Civil Defence, the fire brigades, the Gendarmerie etc. Each exercise is followed by a meeting which gives an opportunity for constructive criticism and for the adoption of measures best suited for solving problems which invariably arise, such as choice of methods and resources, co-ordination of their simultaneous or gradual application and so on. (author)

  7. Introduction to 'International Handbook of Criticality Safety Benchmark Experiments'

    International Nuclear Information System (INIS)

    Komuro, Yuichi

    1998-01-01

    The Criticality Safety Benchmark Evaluation Project (CSBEP) was initiated in 1992 by the United States Department of Energy. The project quickly became an international effort as scientists from other interested countries became involved. The International Criticality Safety Benchmark Evaluation Project (ICSBEP) is now an official activity of the Organization for Economic Cooperation and Development-Nuclear Energy Agency (OECD-NEA). 'International Handbook of Criticality Safety Benchmark Experiments' was prepared and is updated year by year by the working group of the project. This handbook contains criticality safety benchmark specifications that have been derived from experiments that were performed at various nuclear critical facilities around the world. The benchmark specifications are intended for use by criticality safety engineers to validate calculation techniques used. The author briefly introduces the informative handbook and would like to encourage Japanese engineers who are in charge of nuclear criticality safety to use the handbook. (author)

  8. ACRR fuel storage racks criticality safety analysis

    International Nuclear Information System (INIS)

    Bodette, D.E.; Naegeli, R.E.

    1997-10-01

    This document presents the criticality safety analysis for a new fuel storage rack to support modification of the Annular Core Research Reactor for production of molybdenum-99 at Sandia National Laboratories, Technical Area V facilities. Criticality calculations with the MCNP code investigated various contingencies for the criticality control parameters. Important contingencies included mix of fuel element types stored, water density due to air bubbles or water level for the over-moderated racks, interaction with existing fuel storage racks and fuel storage holsters in the fuel storage pool, neutron absorption of planned rack design and materials, and criticality changes due to manufacturing tolerances or damage. Some limitations or restrictions on use of the new fuel storage rack for storage operations were developed through the criticality analysis and are required to meet the double contingency requirements of criticality safety. As shown in the analysis, this system will remain subcritical under all credible upset conditions. Administrative controls are necessary for loading, moving, and handling the storage rack as well as for control of operations around it. 21 refs., 16 figs., 4 tabs

  9. Computerized Italian criticality guide, description and validation

    International Nuclear Information System (INIS)

    Carotenuto, M.; Landeyro, P.A.

    1988-10-01

    Our group is developing an 'expert system' for collecting engineering know-how on back-end nuclear plant design. An expert system is the most suitable software tool for our problem. During the analysis, the design process was divided into different branches. At each branch of the design process the Expert System relates a computerized design procedure. Any design procedure is composed of a set of design methods, together with their condition of application and reliability limits. In the framework of this expert system, the nuclear criticality safety analysis procedure was developed, in the form of a computerized criticality guide, attempting to reproduce the designer's normal 'reasoning' process. The criticality guide is composed of two parts: A computerized text, including theory, a description of the accidents occurred in the past and a description of the italian design experience; An interactive computer aided calculation module, containing a graphical facility for critical parameter curves. In the present report are presented the criticality guide (computerized Italian Criticality Guide) and its validation test. (author)

  10. Computerized Italian criticality guide, description and validation

    Energy Technology Data Exchange (ETDEWEB)

    Carotenuto, M; Landeyro, P A [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Casaccia (Italy)

    1988-10-15

    Our group is developing an 'expert system' for collecting engineering know-how on back-end nuclear plant design. An expert system is the most suitable software tool for our problem. During the analysis, the design process was divided into different branches. At each branch of the design process the Expert System relates a computerized design procedure. Any design procedure is composed of a set of design methods, together with their condition of application and reliability limits. In the framework of this expert system, the nuclear criticality safety analysis procedure was developed, in the form of a computerized criticality guide, attempting to reproduce the designer's normal 'reasoning' process. The criticality guide is composed of two parts: A computerized text, including theory, a description of the accidents occurred in the past and a description of the italian design experience; An interactive computer aided calculation module, containing a graphical facility for critical parameter curves. In the present report are presented the criticality guide (computerized Italian Criticality Guide) and its validation test. (author)

  11. Lessons learned from early criticality accidents

    International Nuclear Information System (INIS)

    Malenfant, R.E.

    1996-01-01

    Four accidents involving the approach to criticality occurred during the period July, 1945, through May, 1996. These have been described in the format of the OPERATING EXPERIENCE WEEKLY SUMMARY which is distributed by the Office of Nuclear and Facility Safety. Although the lessons learned have been incorporated in standards, codes, and formal procedures during the last fifty years, this is their first presentation in this format. It is particularly appropriate that they be presented in the forum of the Nuclear Criticality Technology Safety Project Workshop closest to the fiftieth anniversary of the last of the four accidents, and that which was most instrumental in demonstrating the need to incorporate lessons learned

  12. Analysis on burn-up behaviors for accelerator-driven sub-critical facility

    International Nuclear Information System (INIS)

    Liu Guisheng; Zhao Zhixiang; Zhang Baocheng; Shen Qinbiao; Ding Dazhao

    2000-01-01

    An analysis is performed on burn-up behaviors for accelerator-driven sub-critical reactor by means of the code PASC-1 for neutronics calculation, the code CBURN for burn-up calculation and 44 group constants is processed by CENDL-2 and ENDF/B-6 using NJOY-91.91

  13. Tritium Systems Test Facility. Volume I

    International Nuclear Information System (INIS)

    Anderson, G.W.; Battleson, K.W.; Bauer, W.

    1976-10-01

    Sandia Laboratories proposes to build and operate a Tritium Systems Test Facility (TSTF) in its newly completed Tritium Research Laboratory at Livermore, California (see frontispiece). The facility will demonstrate at a scale factor of 1:200 the tritium fuel cycle systems for an Experimental Power Reactor (EPR). This scale for each of the TSTF subsystems--torus, pumping system, fuel purifier, isotope separator, and tritium store--will allow confident extrapolation to EPR dimensions. Coolant loop and reactor hall cleanup facilities are also reproduced, but to different scales. It is believed that all critical details of an EPR tritium system will be simulated correctly in the facility. Tritium systems necessary for interim devices such as the Ignition Test Reactor (ITR) or The Next Step (TNS) can also be simulated in TSTF at other scale values. The active tritium system will be completely enclosed in an inert atmosphere glove box which will be connected to the existing Gas Purification System (GPS) of the Tritium Research Laboratory. In effect, the GPS will become the scaled environmental control system which otherwise would have to be built especially for the TSTF

  14. Source term analysis for a criticality accident in metal production line glove boxes

    International Nuclear Information System (INIS)

    Nguyen, D.H.

    1991-06-01

    A recent development in criticality accident analysis is the deterministic calculations of the transport of fission products and actinides through the barriers of the physical facility. The knowledge of the redistribution of the materials inside the facility will help determine the reentry and clean-up procedures. The amount of radioactive materials released to the environment is the source term for dispersion calculations. We have used an integrated computer model to determine the release of fission products to the environment from a hypothetical criticality event in a glove box of the metal production line (MPL) at the Lawrence Livermore National Laboratory (LLNL)

  15. Proposal for a verification facility of ADS in China

    International Nuclear Information System (INIS)

    Guan Xialing; Luo Zhanglin

    1999-01-01

    The concept, general layout and some specifications of a proposed verification facility of the accelerator driven radioactive clean nuclear power system (AD-RCNPS) in China are described. It is composed of a 150 MeV/3 mA low energy accelerator, a swimming pool reactor and some basic research facilities. The 150 MeV accelerator consists of an ECR proton source, LEBT, RFQ, CCDTL and SCC. As the sub-critical reactor, the swimming pool reactor is an existing research reactor at the China Institute of Atomic Energy, whose maximum output power is 3.5 MW. The effect of the instability of proton beam and possibility of simulation tests on the verification facility have been analysed. (author)

  16. Proposal for a verification facility of ADS in China

    International Nuclear Information System (INIS)

    Guan Xialing; Luo Zhanglin

    2000-01-01

    The concept, the general layout and some specifications of a proposed verification facility of the accelerator driven radioactive clean nuclear power system (AD-RCNPS) in China has been described. It is composed of a 150 MeV/3 mA low energy accelerator, a swimming pool reactor and some basic research facility. The 150 MeV accelerator consists of an ECR proton source, LEBT, RFQ, CCDTL and SCC. As the sub-critical reactor, the swimming pool reactor is an existing research reactor in China Institute of Atomic Energy, its maximum output power is 3.5 MW. The effect of the instability of proton beam and possibility of simulation test on the verification facility have been analyzed

  17. ANSI/ANS-8.23-1997: nuclear criticality accident emergency planning and response

    International Nuclear Information System (INIS)

    Baker, J.S.

    2004-01-01

    American National Standard ANSUANS-8.23 was developed to expand upon the basic emergency response guidance given in American National Standard, 'Administrative Practices for Nuclear Criticality Safety' ANSI/ANS-8.19-1996 (Ref. 1). This standard provides guidance for minimizing risks to personnel during emergency response to a nuclear criticality accident outside reactors. This standard is intended to apply to those facilities for which a criticality accident alarm system, as specified in American National Standard, 'Criticality Accident Alarm System', ANSI/ANS-8.3-1997 (Ref. 2) is in use. The Working Group was established in 1990, with Norman L. Pruvost as chairman. The Working Group had up to twenty-three members representing a broad range of the nuclear industry, and has included members from Canada, Japan and the United Kingdom. The initial edition of ANSI/ANS-8.23 was approved by the American National Standards Institute on December 30, 1997. It provides guidance for the following topics: (1) Management and technical staff responsibilities; (2) Evaluation of a potential criticality accident; (3) Emergency plan provisions; (4) Evacuation; (5) Re-entry, rescue and stabilization; and (6) Classroom training, exercises and evacuation drills. This guidance is not for generic emergency planning issues, but is specific to nuclear criticality accidents. For example, it assumes that an Emergency Plan is already established at facilities that implement the standard. During the development of the initial edition of ANSI/ANS-8.23, each Working Group member evaluated potential use of the standard at a facility with which the member was familiar. This revealed areas where a facility could have difficulty complying with the standard. These reviews helped identify and eliminate many potential problems and ambiguities with the guidance. The Working Group has received very limited feedback from the user community since the first edition of the standard was published. Suggestions

  18. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

    Ayer, J.E.; Clark, A.T.; Loysen, P.; Ballinger, M.Y.; Mishima, J.; Owczarski, P.C.; Gregory, W.S.; Nichols, B.D.

    1988-05-01

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  19. Impact of Distributed Energy Resources on the Reliability of Critical Telecommunications Facilities: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, D. G.; Arent, D. J.; Johnson, L.

    2006-06-01

    This paper documents a probabilistic risk assessment of existing and alternative power supply systems at a large telecommunications office. The analysis characterizes the increase in the reliability of power supply through the use of two alternative power configurations. Failures in the power systems supporting major telecommunications service nodes are a main contributor to significant telecommunications outages. A logical approach to improving the robustness of telecommunication facilities is to increase the depth and breadth of technologies available to restore power during power outages. Distributed energy resources such as fuel cells and gas turbines could provide additional on-site electric power sources to provide backup power, if batteries and diesel generators fail. The analysis is based on a hierarchical Bayesian approach and focuses on the failure probability associated with each of three possible facility configurations, along with assessment of the uncertainty or confidence level in the probability of failure. A risk-based characterization of final best configuration is presented.

  20. INTEGRATION OF FACILITY MODELING CAPABILITIES FOR NUCLEAR NONPROLIFERATION ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.; Hamm, L.; Garcia, H.; Burr, T.; Coles, G.; Edmunds, T.; Garrett, A.; Krebs, J.; Kress, R.; Lamberti, V.; Schoenwald, D.; Tzanos, C.; Ward, R.

    2011-07-18

    Developing automated methods for data collection and analysis that can facilitate nuclear nonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facility modeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facility modeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facility modeling capabilities and illustrates how they could be integrated and utilized for nonproliferation analysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facility modeling tools. After considering a representative sampling of key facility modeling capabilities, the proposed integration framework is illustrated with several examples.

  1. Integration of facility modeling capabilities for nuclear nonproliferation analysis

    International Nuclear Information System (INIS)

    Garcia, Humberto; Burr, Tom; Coles, Garill A.; Edmunds, Thomas A.; Garrett, Alfred; Gorensek, Maximilian; Hamm, Luther; Krebs, John; Kress, Reid L.; Lamberti, Vincent; Schoenwald, David; Tzanos, Constantine P.; Ward, Richard C.

    2012-01-01

    Developing automated methods for data collection and analysis that can facilitate nuclear nonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facility modeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facility modeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facility modeling capabilities and illustrates how they could be integrated and utilized for nonproliferation analysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facility modeling tools. After considering a representative sampling of key facility modeling capabilities, the proposed integration framework is illustrated with several examples.

  2. Integration Of Facility Modeling Capabilities For Nuclear Nonproliferation Analysis

    International Nuclear Information System (INIS)

    Gorensek, M.; Hamm, L.; Garcia, H.; Burr, T.; Coles, G.; Edmunds, T.; Garrett, A.; Krebs, J.; Kress, R.; Lamberti, V.; Schoenwald, D.; Tzanos, C.; Ward, R.

    2011-01-01

    Developing automated methods for data collection and analysis that can facilitate nuclear nonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facility modeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facility modeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facility modeling capabilities and illustrates how they could be integrated and utilized for nonproliferation analysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facility modeling tools. After considering a representative sampling of key facility modeling capabilities, the proposed integration framework is illustrated with several examples.

  3. PND fuel handling decontamination: facilities and techniques

    Energy Technology Data Exchange (ETDEWEB)

    Pan, R Y [Ontario Hydro, Toronto, ON (Canada)

    1997-12-31

    The use of various decontamination techniques and equipment has become a critical part of Fuel Handling maintenance work at Ontario Hydro`s Pickering Nuclear Division. This paper presents an overview of the set up and techniques used for decontamination in the PND Fuel Handling Maintenance Facility and the effectiveness of each. (author). 1 tab., 9 figs.

  4. R.4. Innovative concept for plutonium finishing facility

    International Nuclear Information System (INIS)

    Bertolotti, G.; Laguerie, I.V. de; Richter, R.; Gillet, B.

    1998-01-01

    After complete shutdown of the units of the previous UP2 plant, the new R4 facility will ensure the purification of Plutonium of the UP2-800 plant for the whole range of fuel to be reprocessed in the years to come. This facility features four main units: - Purification of plutonium nitrate; - Conversion into plutonium oxide; - PuO 2 conditioning into cans; Acid recovery. An extensive R and D program resulted in significant innovations. From a technological aspect the centrifugal extractor and the sub-critical tube bundle tank contributed to the reduction of the building dimensions. The extensive use of on-line analyses enables a more efficient follow-up of the process while minimizing the effluent production. On the other hand, the organization of the building which consists in grouping the rooms presenting the same risk of dispersal of nuclear materials also contributed to reduce the active zone volume. This facility, as any other facilities on the LA HAGUE site, will be remotely operated. (author)

  5. Status, plans, and capabilities of the Nuclear Criticality Information System

    International Nuclear Information System (INIS)

    Koponen, B.L.

    1984-01-01

    The Nuclear Criticality Information System (NCIS), in preparation since 1981, has substantially evolved and now contains a growing number of resources pertinent to nuclear criticality safety. These resources include bibliographic compilations, experimental data, communications media, and the International Directory of Nuclear Criticality Safety Personnel. These resources are part of the LLNL Technology Information System (TIS) which provides the host computer for NCIS. The TIS provides nationwide access to authorized members of the nuclear criticality community via interactive dial-up from computer terminals that utilize communication facilities such as commercial and federal telephone networks, toll-free WATS lines, TYMNET, and the ARPANET/MILNET computer network

  6. HEROICA: A fast screening facility for the characterization of germanium detectors

    Energy Technology Data Exchange (ETDEWEB)

    Andreotti, Erica [Universität Tübingen, Auf der Morgenstelle 14, 72076 Tübingen (Germany); Collaboration: GERDA Collaboration

    2013-08-08

    In the course of 2012, a facility for the fast screening of germanium detectors called HEROICA (Hades Experimental Research Of Intrinsic Crystal Appliances) has been installed at the HADES underground laboratory in the premises of the Belgian Nuclear Research Centre SCK•CEN, in Mol (Belgium). The facility allows performing a complete characterization of the critical germanium detectors' operational parameters with a rate of about two detectors per week.

  7. A whole-school approach to facilities maintenance / Velaphi Aaron Nhlapo.

    OpenAIRE

    Nhlapo, Velaphi Aaron

    2009-01-01

    The area of school facilities maintenance as an integral component of schools' educational programmes is only beginning to receive attention in South Africa, through the publishing of Notice 1438 of 2008 of the National Education Policy, which is a call for comments on the National Policy for an Equitable Provision of an Enabling School Physical Teaching and Learning Environment. This implies that, while it is a critical aspect of teaching and learning, school facilities maintenance has not b...

  8. Vibration isolation in a free-piston driven expansion tube facility

    Science.gov (United States)

    Gildfind, D. E.; Jacobs, P. A.; Morgan, R. G.

    2013-09-01

    The stress waves produced by rapid piston deceleration are a fundamental feature of free-piston driven expansion tubes, and wave propagation has to be considered in the design process. For lower enthalpy test conditions, these waves can traverse the tube ahead of critical flow processes, severely interfering with static pressure measurements of the passing flow. This paper details a new device which decouples the driven tube from the free-piston driver, and thus prevents transmission of stress waves. Following successful incorporation of the concept in the smaller X2 facility, it has now been applied to the larger X3 facility, and results for both facilities are presented.

  9. Recent trends of plutonium facilities and their control

    Energy Technology Data Exchange (ETDEWEB)

    Muto, T [Power Reactor and Nuclear Fuel Development Corp., Tokai, Ibaraki (Japan). Tokai Works

    1974-02-01

    Much interest has been focussed on Pu recycle since the oil crisis because of an expected shortage of enriched uranium. Plutonium handling techniques and plutonium fuel fabricating facilities should be developed to meet the future demand of plutonium, but the radioactive property of plutonium to be reprocessed from spent fuel and recycled plutonium is remarkably different, and it has to be handled safely. Technical criteria for plutonium facilities are specified in the USAEC regulatory guides and other rules. Some of these criteria are location condition, quality of confinement, protection against accidents and so on. The control conditions for plutonium facilities are exposure control, criticality control, measurement control and new system of safeguard. These problems are under development to meet the future requirement for the safe handling of Pu material.

  10. Criticality Safety Evaluation for the TACS at DAF

    Energy Technology Data Exchange (ETDEWEB)

    Percher, C. M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinrichs, D. P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2011-06-10

    Hands-on experimental training in the physical behavior of multiplying systems is one of ten key areas of training required for practitioners to become qualified in the discipline of criticality safety as identified in DOE-STD-1135-99, Guidance for Nuclear Criticality Safety Engineer Training and Qualification. This document is a criticality safety evaluation of the training activities and operations associated with HS-3201-P, Nuclear Criticality 4-Day Training Course (Practical). This course was designed to also address the training needs of nuclear criticality safety professionals under the auspices of the NNSA Nuclear Criticality Safety Program1. The hands-on, or laboratory, portion of the course will utilize the Training Assembly for Criticality Safety (TACS) and will be conducted in the Device Assembly Facility (DAF) at the Nevada Nuclear Security Site (NNSS). The training activities will be conducted by Lawrence Livermore National Laboratory following the requirements of an Integrated Work Sheet (IWS) and associated Safety Plan. Students will be allowed to handle the fissile material under the supervision of an LLNL Certified Fissile Material Handler.

  11. Criticality management organization in the alpha incinerator

    International Nuclear Information System (INIS)

    Devillard, D.; Thiebaut, C.; Poinso, J.Y.; Huin, M.

    2004-01-01

    The Valduc Research Center, which reports to the CEA's Military applications Division, generates solid wastes contaminated with alpha emitters in the operation of its installations. An incineration plant has been built to treat these contaminated wastes. Criticality risk prevention is based on limiting the mass of active material undergoing treatment in the facility. A balance is compiled continuously by calculating the difference between the mass of active material entering the facility and the mass leaving it. Due to measurement uncertainties, the balance must be zeroed periodically by cleaning and drainage of all the equipment and the absence of holdup in the components must be checked. (authors)

  12. Method of preventing criticality of fresh fuel assembly in storage facility

    International Nuclear Information System (INIS)

    Kawamura, Makoto.

    1990-01-01

    With an aim of improving the operation efficiency of a reactor, extention of the operation cycle by increasing U 235 enrichment degree of fuel uranium is planned. However, along with the increase of the enrichment degree of the fuel uranium, there occurs a problem of criticality upon fuel handling. Then, in the present invention, boric acid incorporating B-10 of great neutron absorption effect are packed with water soluble polymeric materials which are further packed with a fuel packing sheet, or the water soluble polymeric materials incorporating boric acids are packed with fuel packing sheets which are disposed to a fresh fuel assembly and stored in a store house as they are. The fuel packing sheet is a perforated sheet having a plurality of water intruding pores. Then, if water should intrude to the store house accidentally, the water soluble polymeric materials are dissolved, so that the intruded water is converted into aqueous boric acid easily and absorbs neutrons effectively to thereby attain the prevention of criticality. (T.M.)

  13. Forecast of criticality experiments and experimental programs needed to support nuclear operations in the United States of America: 1994-1999

    International Nuclear Information System (INIS)

    Rutherford, D.

    1995-01-01

    This Forecast is generated by the Chair of the Experiment Needs Identification Workgroup (ENIWG), with input from Department of Energy and the nuclear community. One of the current concerns addressed by ENIWG was the Defense Nuclear Facilities Safety Board's Recommendation 93-2. This Recommendation delineated the need for a critical experimental capability, which includes (1) a program of general-purpose experiments, (2) improving the information base, and (3) ongoing departmental programs. The nuclear community also recognizes the importance of criticality theory, which, as a stepping stone to computational analysis and safety code development, needs to be benchmarked against well-characterized critical experiments. A summary projection of the Department's needs with respect to criticality information includes (1) hands-on training, (2) criticality and nuclear data, (3) detector systems, (4) uranium- and plutonium-based reactors, and (5) accident analysis. The Workgroup has evaluated, prioritized, and categorized each proposed experiment and program. Transportation/Applications is a new category intended to cover the areas of storage, training, emergency response, and standards. This category has the highest number of priority-1 experiments (nine). Facilities capable of performing experiments include the Los Alamos Critical Experiment Facility (LACEF) along with Area V at Sandia National Laboratory. The LACEF continues to house the most significant collection of critical assemblies in the Western Hemisphere. The staff of this facility and Area V are trained and certified, and documentation is current. ENIWG will continue to work with the nuclear community to identify and prioritize experiments because there is an overwhelming need for critical experiments to be performed for basic research and code validation

  14. Forecast of criticality experiments and experimental programs needed to support nuclear operations in the United States of America: 1994--1999

    International Nuclear Information System (INIS)

    Rutherford, D.

    1994-03-01

    This Forecast is generated by the Chair of the Experiment Needs Identification Workgroup (ENIWG), with input from Department of Energy and the nuclear community. One of the current concerns addressed by ENIWG was the Defense Nuclear Facilities Safety Board's Recommendation 93-2. This Recommendation delineated the need for a critical experimental capability, which includes (1) a program of general-purpose experiments, (2) improving the information base, and (3) ongoing departmental programs. The nuclear community also recognizes the importance of criticality theory, which, as a stepping stone to computational analysis and safety code development, needs to be benchmarked against well-characterized critical experiments. A summary project of the Department's needs with respect to criticality information includes (1) hands-on training, (2) criticality and nuclear data, (3) detector systems, (4) uranium- and plutonium-based reactors, and (5) accident analysis. The Workgroup has evaluated, prioritized, and categorized each proposed experiment and program. Transportation/Applications is a new category intended to cover the areas of storage, training, emergency response, and standards. This category has the highest number of priority-1 experiments (nine). Facilities capable of performing experiments include the Los Alamos Critical Experiment Facility (LACEF) along with Area V at Sandia National Laboratory. The LACEF continues to house the most significant collection of critical assemblies in the Western Hemisphere. The staff of this facility and Area V are trained and certified, and documentation is current. ENIWG will continue to work with the nuclear community to identify and prioritize experiments because there is an overwhelming need for critical experiments to be performed for basic research and code validation

  15. Automated Critical Peak Pricing Field Tests: Program Descriptionand Results

    Energy Technology Data Exchange (ETDEWEB)

    Piette, Mary Ann; Watson, David; Motegi, Naoya; Kiliccote, Sila; Xu, Peng

    2006-04-06

    California utilities have been exploring the use of critical peak prices (CPP) to help reduce needle peaks in customer end-use loads. CPP is a form of price-responsive demand response (DR). Recent experience has shown that customers have limited knowledge of how to operate their facilities in order to reduce their electricity costs under CPP (Quantum 2004). While the lack of knowledge about how to develop and implement DR control strategies is a barrier to participation in DR programs like CPP, another barrier is the lack of automation of DR systems. During 2003 and 2004, the PIER Demand Response Research Center (DRRC) conducted a series of tests of fully automated electric demand response (Auto-DR) at 18 facilities. Overall, the average of the site-specific average coincident demand reductions was 8% from a variety of building types and facilities. Many electricity customers have suggested that automation will help them institutionalize their electric demand savings and improve their overall response and DR repeatability. This report focuses on and discusses the specific results of the Automated Critical Peak Pricing (Auto-CPP, a specific type of Auto-DR) tests that took place during 2005, which build on the automated demand response (Auto-DR) research conducted through PIER and the DRRC in 2003 and 2004. The long-term goal of this project is to understand the technical opportunities of automating demand response and to remove technical and market impediments to large-scale implementation of automated demand response (Auto-DR) in buildings and industry. A second goal of this research is to understand and identify best practices for DR strategies and opportunities. The specific objectives of the Automated Critical Peak Pricing test were as follows: (1) Demonstrate how an automated notification system for critical peak pricing can be used in large commercial facilities for demand response (DR). (2) Evaluate effectiveness of such a system. (3) Determine how customers

  16. AEC Regulatory view of the reliability of air cleaning systems in nuclear facilities

    International Nuclear Information System (INIS)

    Bellamy, R.R.; Zavadoski, R.W.

    1975-01-01

    Air cleaning systems in nuclear facilities can be divided into three categories: ventilation exhaust systems, containment atmosphere cleanup systems, and process offgas systems. These systems have been the subject of numerous reports, regulatory guides, discussions, and meetings. Some of the analyses have been critical of the operation and design of these air cleaning systems--in particular, the engineered safety features containment atmosphere cleanup systems. Although for the most part the criticism is applicable, and recognizing that there are a number of unresolved issues pertaining to gaseous waste management systems, there are data to show that air cleaning systems in use in nuclear facilities are performing their intended function. (U.S.)

  17. Safety of fuel cycle facilities. Topical issues paper no. 3

    International Nuclear Information System (INIS)

    Ranguelova, V.; Niehaus, F.; Delattre, D.

    2001-01-01

    A wide range of nuclear fuel cycle facilities are in operation. These installations process, use, store and dispose of radioactive material and cover: mining and milling, conversion, enrichment, fuel fabrication (including mixed oxide fuel), reactor, interim spent fuel storage, reprocessing, waste treatment and waste disposal facilities. For the purposes of this paper, reactors and waste disposal facilities are not considered. The term 'fuel cycle facilities' covers only the remainder of the installations listed above. The IAEA Secretariat maintains a database of fuel cycle facilities in its Member States. Known as the Nuclear Fuel Cycle Information System (NFCIS), it is available as an on-line service through the Internet. More than 500 such facilities have been reported under this system. The facilities are listed by facility type and operating status. Approximately one third of all of the facilities are located in developing States. About half of all facilities are reported to be operating, of which approximately 40% are operating in developing States. In addition, some 60 facilities are either in the design stage or under construction. Although the radioactive source term for most fuel cycle facilities is lower than the source term for reactors, which results in less severe consequences to the public from potential accidents at these fuel cycle installations, recent events at some fuel cycle facilities have given rise to public concern which has to be addressed adequately by national regulatory bodies and at the international level. Worldwide, operational experience feedback warrants improvements in the safety of these facilities. Some of the hazards are similar for reactor and non-reactor facilities. However, the differences between these installations give rise to specific safety concerns at fuel cycle facilities. In particular, these concerns include: criticality, radiation protection of workers, chemical hazards, fire and explosion hazards. It is recognized

  18. Abnormal Events for Reactor System and Facilities in HANARO

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Ho Young; Lee, B. H.; Lee, M.; Kang, I. H.; Lee, U. G.; Sin, H. C.; Park, C. Y.; Song, B. S.; Lee, S. H.; Han, J. S

    2006-12-15

    This report gathers abnormal events related to reactor system and facilities of HANARO that happened during its operation over 10 years since the first criticality on February 1995. The collected examples will be utilized to the HANARO's operators as a useful guide.

  19. Test facilities for radioactive material transport packages (AEA Technology plc, Winfrith,UK)

    International Nuclear Information System (INIS)

    Gillard, J.E.

    2001-01-01

    Transport containers for radioactive materials are tested to demonstrate compliance with national and international standards. Transport package design, testing, assessment and approval requires a wide range of skills and facilities. The comprehensive capability of AEA Technology in these areas is described. The facilities described include drop-test cranes and targets (up to 700 tonne); pool fires, furnaces and rigs for thermal tests, including heat dissipation on prototype flasks; shielding facilities; criticality simulations and leak test techniques. These are illustrated with photographs demonstrating the comprehensive nature of package testing services supplied to customers. (author)

  20. Test facilities for radioactive material transport packages (AEA Technology plc, Winfrith,UK)

    Energy Technology Data Exchange (ETDEWEB)

    Gillard, J.E

    2001-07-01

    Transport containers for radioactive materials are tested to demonstrate compliance with national and international standards. Transport package design, testing, assessment and approval requires a wide range of skills and facilities. The comprehensive capability of AEA Technology in these areas is described. The facilities described include drop-test cranes and targets (up to 700 tonne); pool fires, furnaces and rigs for thermal tests, including heat dissipation on prototype flasks; shielding facilities; criticality simulations and leak test techniques. These are illustrated with photographs demonstrating the comprehensive nature of package testing services supplied to customers. (author)

  1. Assessment of critical flow models of RELAP5-MOD2 and CATHARE codes

    International Nuclear Information System (INIS)

    Hao Laomi; Zhu Zhanchuan

    1992-01-01

    The critical flow tests for the long and short nozzles conducted on the SUPER MOBY-DICK facility were analyzed using the RELAP5-MOD2 and CATHARE 1.3 codes to assess the critical flow models of two codes. The critical mass flux calculated for two nozzles are given. The CATHARE code has used the thermodynamic nonequilibrium sound velocity of the two-phase fluid as the critical flow criterion, and has the better interphase transfer models and calculates the critical flow velocities with the completely implicit solution. Therefore, it can well calculate the critical flowrate and can describe the effect of the geometry L/D on the critical flowrate

  2. Criticality Safety Support to a Project Addressing SNM Legacy Items at LLNL

    International Nuclear Information System (INIS)

    Pearson, J S; Burch, J G; Dodson, K E; Huang, S T

    2005-01-01

    The programmatic, facility and criticality safety support staffs at the LLNL Plutonium Facility worked together to successfully develop and implement a project to process legacy (DNFSB Recommendation 94-1 and non-Environmental, Safety, and Health (ES and H) labeled) materials in storage. Over many years, material had accumulated in storage that lacked information to adequately characterize the material for current criticality safety controls used in the facility. Generally, the fissionable material mass information was well known, but other information such as form, impurities, internal packaging, and presence of internal moderating or reflecting materials were not well documented. In many cases, the material was excess to programmatic need, but such a determination was difficult with the little information given on MC and A labels and in the MC and A database. The material was not packaged as efficiently as possible, so it also occupied much more valuable storage space than was necessary. Although safe as stored, the inadequately characterized material posed a risk for criticality safety noncompliances if moved within the facility under current criticality safety controls. A Legacy Item Implementation Plan was developed and implemented to deal with this problem. Reasonable bounding conditions were determined for the material involved, and criticality safety evaluations were completed. Two appropriately designated glove boxes were identified and criticality safety controls were developed to safely inspect the material. Inspecting the material involved identifying containers of legacy material, followed by opening, evaluating, processing if necessary, characterizing and repackaging the material. Material from multiple containers was consolidated more efficiently thus decreasing the total number of stored items to about one half of the highest count. Current packaging requirements were implemented. Detailed characterization of the material was captured in databases

  3. Stakeholders' perception of critical success factors for sustainable ...

    African Journals Online (AJOL)

    interval' and 'adherence to the tenets of the SD agenda (supply chain)' were selected as most critical of the success factors identified. It is expected that the study's findings will contribute to the development of a viable SFM strategy in SSA universities. Keywords: Facilities management, sub-Saharan Africa, success factors, ...

  4. A medical facility proposal to use the SSC linac

    International Nuclear Information System (INIS)

    Funk, L.W.

    1994-01-01

    A consortium organized by the Texas National Research Laboratory Commission under a Department of Energy grant proposes to build and operate a Regional Medical Technology Center to function as a combined medical radioisotope production complex and proton cancer therapy facility using the Linear Accelerator (Linac) assets of the Superconducting Super Collider (SSC). The radioisotope production complex will serve as a domestic source of radioisotopes critically needed by the U.S. pharmaceutical industry and nuclear medicine facilities throughout North America. Presently, more than 70 percent of radioisotopes used in U.S. nuclear medicine procedures are produced outside the country. The Center's state-of-the-art proton cancer therapy facility will serve the Central United States, providing advanced capabilities and augmenting facilities in California and Massachusetts. Long-term, it is anticipated that the RMTC also will stimulate nuclear medicine research, advance medical diagnostic technologies, and generate new industrial applications for linear accelerator technology

  5. Material accountancy and control practice at a research reactor facility

    International Nuclear Information System (INIS)

    Bouchard, J.; Maurel, J.J.; Tromeur, Y.

    1982-01-01

    This session surveys the regulations, organization, and accountancy practice that compose the French State System of Accountancy and Control. Practical examples are discussed showing how inventories are verified at a critical assembly facility and at a materials testing reactor

  6. Material selection for Multi-Function Waste Tank Facility tanks

    International Nuclear Information System (INIS)

    Carlos, W.C.

    1994-01-01

    This report briefly summarizes the history of the materials selection for the US Department of Energy's high-level waste carbon steel storage tanks. It also provide an evaluation of the materials for the construction of new tanks at the Multi-Function Waste Tank Facility. The evaluation included a materials matrix that summarized the critical design, fabrication, construction, and corrosion resistance requirements; assessed each requirement; and cataloged the advantages and disadvantages of each material. This evaluation is based on the mission of the Multi-Function Waste Tank Facility. On the basis of the compositions of the wastes stored in Hanford waste tanks, it is recommended that tanks for the Multi-Function Waste Tank Facility be constructed of normalized ASME SA 516, Grade 70, carbon steel

  7. Review of Sodium and Plutonium related Technical Standards in Trans-Uranium Fuel Fabrication Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF.

  8. Review of Sodium and Plutonium related Technical Standards in Trans-Uranium Fuel Fabrication Facilities

    International Nuclear Information System (INIS)

    Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae

    2016-01-01

    In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF

  9. Maintaining Mission Critical Systems in a 247 Environment

    CERN Document Server

    Curtis, Peter M

    2011-01-01

    "This book is meant to offer Architects, Property Mangers, Facility Managers, Building Engineers, Information Technology Professionals, Data Center Personnel, Electrical & Mechanical Technicians and students in undergraduate, graduate, or continuing education programs relevant insight into the Mission Critical Environment with an emphasis on business resiliency, data center efficiency, and green power technology. Industry improvements, standards, and techniques have been incorporated into the text and address the latest issues prevalent in the Mission Critical Industry. An emphasis on green technologies and certifications is presented throughout the book. In addition, a description of the United States energy infrastructure's dependency on oil, in relation to energy security in the mission critical industry, is discussed. In conjunction with this, either a new chapter will be created on updated policies and regulations specifically related to the mission critical industry or updates to policies and regula...

  10. Criticality incident detection assessment methodology

    International Nuclear Information System (INIS)

    Haley, Richard M.; Warburton, Simon J.; Bowden, Russell L.

    2003-01-01

    In the United Kingdom, all nuclear facilities that handle, treat or store fissile material require a Criticality Incident Detection and Alarm System (CIDAS) to be installed, unless a case is made for the omission of such a system. Where it is concluded that a CIDAS is required, the primary objective is the reliable detection of criticality and the initiation of prompt evacuation of plant workers from the vicinity of the incident. This paper will examine and compare various methods that can be used to demonstrate that a CIDAS will satisfy the detection criterion. The paper will focus on fit-for-purpose and cost-effective methods for the assessment of gamma-based systems. In the experience of the authors this is particularly useful in demonstrating the efficacy of existing systems in operational plant. (author)

  11. Emergency preparedness source term development for the Office of Nuclear Material Safety and Safeguards-Licensed Facilities

    International Nuclear Information System (INIS)

    Sutter, S.L.; Mishima, J.; Ballinger, M.Y.; Lindsey, C.G.

    1984-08-01

    In order to establish requirements for emergency preparedness plans at facilities licensed by the Office of Nuclear Materials Safety and Safeguards, the Nuclear Regulatory Commission (NRC) needs to develop source terms (the amount of material made airborne) in accidents. These source terms are used to estimate the potential public doses from the events, which, in turn, will be used to judge whether emergency preparedness plans are needed for a particular type of facility. Pacific Northwest Laboratory is providing the NRC with source terms by developing several accident scenarios for eleven types of fuel cycle and by-product operations. Several scenarios are developed for each operation, leading to the identification of the maximum release considered for emergency preparedness planning (MREPP) scenario. The MREPP scenarios postulated were of three types: fire, tornado, and criticality. Fire was significant at oxide fuel fabrication, UF 6 production, radiopharmaceutical manufacturing, radiopharmacy, sealed source manufacturing, waste warehousing, and university research and development facilities. Tornadoes were MREPP events for uranium mills and plutonium contaminated facilities, and criticalities were significant at nonoxide fuel fabrication and nuclear research and development facilities. Techniques for adjusting the MREPP release to different facilities are also described

  12. Critical Experiments With Aqueous Solutions of 233UO2(NO3)2

    International Nuclear Information System (INIS)

    Thomas, J.T.

    2001-01-01

    This report provides the critical experimenter's interpretations and descriptions of informal critical experiment logbook notes and associated information (e.g., experimental equipment designs/sketches, chemical and isotopic analyses, etc.) for the purpose of formally documenting the results of critical experiments performed in the late 1960s at the Oak Ridge Critical Experiments Facility. The experiments were conducted with aqueous solutions of 97.6 wt % 233 U uranyl nitrate having uranium densities varying between about 346 g U/l and 45 g U/l. Criticality was achieved with single simple units (e.g., cylinders and spheres) and with spaced subcritical simple cylindrical units arranged in unreflected, water-reflected, and polyethylene reflected critical arrays

  13. Status of the joint United States/Japanese criticality data development program for the Fast Breeder fuel cycle

    International Nuclear Information System (INIS)

    Takeda, H.; Kishimoto, Y.; Matsumoto, T.; Haire, M.J.

    1987-01-01

    An experimental program is described to accumulate the basic criticality data on uranium-plutonium aqueous solution systems, including pin and water systems. This experimental program, managed by the Consolidated Fuel Reprocessing Program, utilized existing critical experiment facilities. However, Japan at least for the near future, does not have a facility for conducting critical experiments where fissible solution can be treated and processed. Therefore, Japan has remoted the existing US experimental program as well as enlarged the program so that a close cooperative relationship may exist in this area. This paper is an update of a description of the program presented in September 1985. 9 refs., 5 figs., 1 tab

  14. Neutron-induced nuclear data for the MYRRHA fast spectrum facility

    OpenAIRE

    Romojaro Pablo; Žerovnik Gašper; Álvarez-Velarde Francisco; Stankovskiy Alexey; Kodeli Ivan; Fiorito Luca; Díez Carlos Javier; Cabellos Óscar; García-Herranz Nuria; Heyse Jan; Paradela Carlos; Schillebeeckx Peter; Eynde Gert Van den

    2017-01-01

    The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) concept is a flexible experimental lead-bismuth cooled and mixed-oxide (MOX) fueled fast spectrum facility designed to operate both in sub-critical (accelerator driven) and critical modes. One of the key issues for the safe operation of the reactor is the uncertainty assessment during the design works. The main objective of the European project CHANDA (solving CHAllenges in Nuclear DAta) Work Package 10 is to improv...

  15. Basis for Interim Operation for Fuel Supply Shutdown Facility

    International Nuclear Information System (INIS)

    BENECKE, M.W.

    2003-01-01

    This document establishes the Basis for Interim Operation (BIO) for the Fuel Supply Shutdown Facility (FSS) as managed by the 300 Area Deactivation Project (300 ADP) organization in accordance with the requirements of the Project Hanford Management Contract procedure (PHMC) HNF-PRO-700, ''Safety Analysis and Technical Safety Requirements''. A hazard classification (Benecke 2003a) has been prepared for the facility in accordance with DOE-STD-1027-92 resulting in the assignment of Hazard Category 3 for FSS Facility buildings that store N Reactor fuel materials (303-B, 3712, and 3716). All others are designated Industrial buildings. It is concluded that the risks associated with the current and planned operational mode of the FSS Facility (uranium storage, uranium repackaging and shipment, cleanup, and transition activities, etc.) are acceptable. The potential radiological dose and toxicological consequences for a range of credible uranium storage building have been analyzed using Hanford accepted methods. Risk Class designations are summarized for representative events in Table 1.6-1. Mitigation was not considered for any event except the random fire event that exceeds predicted consequences based on existing source and combustible loading because of an inadvertent increase in combustible loading. For that event, a housekeeping program to manage transient combustibles is credited to reduce the probability. An additional administrative control is established to protect assumptions regarding source term by limiting inventories of fuel and combustible materials. Another is established to maintain the criticality safety program. Additional defense-in-depth controls are established to perform fire protection system testing, inspection, and maintenance to ensure predicted availability of those systems, and to maintain the radiological control program. It is also concluded that because an accidental nuclear criticality is not credible based on the low uranium enrichment

  16. Applicability of base-isolation R and D in non-reactor facilities to a nuclear reactor plant

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1989-01-01

    Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. This paper reviews the research and development (R and D) programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R and D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R and D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant

  17. Use of antibiotics in paediatric long-term care facilities.

    Science.gov (United States)

    Murray, M T; Johnson, C L; Cohen, B; Jackson, O; Jones, L K; Saiman, L; Larson, E L; Neu, N

    2018-06-01

    Adult long-term care (LTC) facilities have high rates of antibiotic use, raising concerns about antimicrobial resistance. Few studies have examined antibiotic use in paediatric LTC facilities. To describe antibiotic use in three paediatric LTC facilities and to describe the factors associated with use. A retrospective cohort study was conducted from September 2012 to December 2015 in three paediatric LTC facilities. Medical records were reviewed for demographics, healthcare-associated infections (HAIs), antimicrobial use and diagnostic testing. Logistic regression was used to identify predictors for antibiotic use. The association between susceptibility testing results and appropriate antibiotic coverage was determined using Chi-squared test. Fifty-eight percent (413/717) of residents had at least one HAI, and 79% (325/413) of these residents were treated with at least one antibiotic course, totalling 2.75 antibiotic courses per 1000 resident-days. Length of enrolment greater than one year, having a neurological disorder, having a tracheostomy, and being hospitalized at least once during the study period were significantly associated with receiving antibiotics when controlling for facility (all P facilities is widespread. There is further need to assess antibiotic use in paediatric LTC facilities. Evaluation of the adverse outcomes associated with inappropriate antibiotic use, including the prevalence of resistant organisms in paediatric LTC facilities, is critical. Copyright © 2017 The Healthcare Infection Society. Published by Elsevier Ltd. All rights reserved.

  18. Forecast of criticality experiments and experimental programs needed to support nuclear operations in the United States of America: 1994--1999

    Energy Technology Data Exchange (ETDEWEB)

    Rutherford, D.

    1994-03-01

    This Forecast is generated by the Chair of the Experiment Needs Identification Workgroup (ENIWG), with input from Department of Energy and the nuclear community. One of the current concerns addressed by ENIWG was the Defense Nuclear Facilities Safety Board`s Recommendation 93-2. This Recommendation delineated the need for a critical experimental capability, which includes (1) a program of general-purpose experiments, (2) improving the information base, and (3) ongoing departmental programs. The nuclear community also recognizes the importance of criticality theory, which, as a stepping stone to computational analysis and safety code development, needs to be benchmarked against well-characterized critical experiments. A summary project of the Department`s needs with respect to criticality information includes (1) hands-on training, (2) criticality and nuclear data, (3) detector systems, (4) uranium- and plutonium-based reactors, and (5) accident analysis. The Workgroup has evaluated, prioritized, and categorized each proposed experiment and program. Transportation/Applications is a new category intended to cover the areas of storage, training, emergency response, and standards. This category has the highest number of priority-1 experiments (nine). Facilities capable of performing experiments include the Los Alamos Critical Experiment Facility (LACEF) along with Area V at Sandia National Laboratory. The LACEF continues to house the most significant collection of critical assemblies in the Western Hemisphere. The staff of this facility and Area V are trained and certified, and documentation is current. ENIWG will continue to work with the nuclear community to identify and prioritize experiments because there is an overwhelming need for critical experiments to be performed for basic research and code validation.

  19. Criticality safety engineering at the Savannah River Site - the 1990s

    International Nuclear Information System (INIS)

    Chandler, J.R.; Apperson, C.E. Jr.

    1996-01-01

    The privatization and downsizing effort that is ongoing within the U.S. Department of Energy (DOE) is requiring a change in the management of criticality safety engineering resources at the Savannah River Site (SRS). Downsizing affects the number of criticality engineers employed by the prime contractor, Westinghouse Savannah River Company (WSRC), and privatization affects the manner in which business is conducted. In the past, criticality engineers at the SRS have been part of the engineering organizations that support each facility handling fissile material. This practice led to different criticality safety engineering organizations dedicated to fuel fabrication activities, reactor loading and unloading activities, separation and waste management operations, and research and development

  20. Massachusetts Large Blade Test Facility Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Rahul Yarala; Rob Priore

    2011-09-02

    Project Objective: The Massachusetts Clean Energy Center (CEC) will design, construct, and ultimately have responsibility for the operation of the Large Wind Turbine Blade Test Facility, which is an advanced blade testing facility capable of testing wind turbine blades up to at least 90 meters in length on three test stands. Background: Wind turbine blade testing is required to meet international design standards, and is a critical factor in maintaining high levels of reliability and mitigating the technical and financial risk of deploying massproduced wind turbine models. Testing is also needed to identify specific blade design issues that may contribute to reduced wind turbine reliability and performance. Testing is also required to optimize aerodynamics, structural performance, encourage new technologies and materials development making wind even more competitive. The objective of this project is to accelerate the design and construction of a large wind blade testing facility capable of testing blades with minimum queue times at a reasonable cost. This testing facility will encourage and provide the opportunity for the U.S wind industry to conduct more rigorous testing of blades to improve wind turbine reliability.

  1. Activity of the Atomic Energy Society of Japan for compiling the consensus standard on nuclear criticality safety control

    International Nuclear Information System (INIS)

    Yamane, Yoshihiro; Matsumoto, Tadakuni

    2003-01-01

    Activity of the Atomic Energy Society of Japan for compiling the consensus standard on nuclear criticality safety control is presented. The standard recommends an enhancement of nuclear criticality safety throughout a life cycle of facility in terms of a concept of 'barriers against criticality'. (author)

  2. A medical facility proposal to use the SSC linac

    International Nuclear Information System (INIS)

    Funk, L.W.

    1995-01-01

    A consortium organized by the Texas National Research Laboratory Commission (TNRLC) under a Department of Energy (DOE) grant proposes to build and operate a Regional Medical Technology Center (RMTC) to function as a combined medical radioisotope production complex and proton cancer therapy facility using the linear accelerator (linac) assets of the cancelled Superconducting Super Collider (SSC). The radioisotope production complex will serve as a domestic source of radioisotopes critically needed by the U.S. pharmaceutical industry and nuclear medicine facilities throughout North America. Presently, more than 70 percent of radioisotopes used in U.S. nuclear medicine procedures are produced outside the country. The Center's state-of-the-art proton cancer therapy facility will serve the Central United States, providing advanced capabilities and augmenting facilities in California and Massachusetts. Long-term, it is anticipated that the RMTC also will stimulate nuclear medicine research, advance medical diagnostic technologies, and generate new industrial applications of linear accelerator technology. (orig.)

  3. A medical facility proposal to use the SSC linac

    Science.gov (United States)

    Warren Funk, L.

    1995-05-01

    A consortium organized by the Texas National Research Laboratory Commission (TNRLC) under a Department of Energy (DOE) grant proposes to build and operate a Regional Medical Technology Center (RMTC) to function as a combined medical radioisotope production complex and proton cancer therapy facility using the linear accelerator (linac) assets of the cancelled Superconducting Super Collider (SSC). The radioisotope production complex will serve as a domestic source of radioisotopes critically needed by the U.S. pharmaceutical industry and nuclear medicine facilities throughout North America. Presently, more than 70 percent of radioisotopes used in U.S. nuclear medicine procedures are produced outside the country. The Center's state-of-the-art proton cancer therapy facility will serve the Central United States, providing advanced capabilities and augmenting facilities in California and Massachusetts. Long-term, it is anticipated that the RMTC also will stimulate nuclear medicine research, advance medical diagnostic technologies, and generate new industrial applications of linear accelerator technology.

  4. Criticality impacts on LWR fuel storage efficiency

    International Nuclear Information System (INIS)

    Napolitano, D.

    1992-01-01

    This presentation discusses the criticality impacts throughout storage of fuel onsite including new fuel storage, spent fuel storage, consolidation, and dry storage. The general principles for criticality safety are also be discussed. There is first an introduction which explains today's situation for criticality safety concerns. This is followed by a discussion of criticality safety Regulatory Guides, safety limits and fundamental principles. Design objectives for criticality safety in the 1990's include higher burnups, longer cycles, and higher enrichments which impact the criticality safety design. Criticality safety for new fuel storage, spent fuel storage, fuel consolidation, and dry storage are followed by conclusions. Today's situation is one in which the US does not reprocess, and does not have an operating MRS facility or repository. High density fuel storage rack designs of the 1980s, are filling up. Dry cask storage systems for spent fuel storage are being utilized. Enrichments continue to increase PWR fuel assemblies with enrichments of 4.5 to 5.0 weight percent U-235 and BWR fuel assemblies with enrichments of 3.25 to 3.5 weight percent U-235 are common. Criticality concerns affect the capacity and the economics of light water reactor (LWR) fuel storage arrays by dictating the spacing of fuel assemblies in a storage system, or the use of poisons or exotic materials in the storage system design

  5. Feasibility study for a postaccident heat removal facility

    International Nuclear Information System (INIS)

    Barts, E.W.; Apperson, C.E. Jr.; Dunwoody, W.E.; Bennett, J.G.

    1978-01-01

    An initial feasibility investigation for PAHRTEF, a Postaccident Heat Removal Test Facility, is presented. The facility would provide an experimental capability for PAHR experiments beyond that available in any currently existing or proposed U.S. safety test facility. The facility design presented in this report is based upon the technology developed for the ROVER nuclear rocket propulsion program. The core is a graphite-moderated, helium-cooled, epithermal core with radial reflector control. The PAHR experiments are located just below the reactor containment vessel, very near the bottom of the core. The experiments (up to 55% enriched) are driven and controlled by neutrons leaking axially from the core such that the PAHRTEF core and the experiment form a coupled reactor system. The experiment can be designed so that it is extremely unlikely that the test fuel by itself could form a critical system. The investigation indicates that adequate fission heating of large PAHR experiments could be provided at low driver core power levels. Both the reactor and the experiment handling and examination equipment can use available technology and, whenever possible, existing equipment and buildings

  6. Nuclear engineering experiments at experimental facilities of JNC in graduate course of Tokyo Institute of Technology

    International Nuclear Information System (INIS)

    Hayashizaki, Noriyosu; Takahashi, Minoru; Aoyama, Takafumi; Onose, Shoji

    2005-01-01

    Nuclear engineering experiments using outside facilities of the campus have been offered for graduate students in the nuclear engineering course in Tokyo Institute of Technology (Tokyo Tech.). The experiments are managed with the collaboration of Japan Nuclear Cycle Development Institute (JNC), Japan Atomic Energy Research Institute (JAERI) and Research Reactor Institute, Kyoto University (KUR). This report presents the new curriculum of the nuclear engineering experiments at JNC since 2002. The change is due to the shutdown of Deuterium Criticality Assembly Facility (DCA) that was used as an experimental facility until 2001. Reactor physics experiment using the training simulator of the experimental fast reactor JOYO is continued from the previous curriculum with the addition of the criticality approach experiment and control rods calibration. A new experimental subject is an irradiated material experiment at the Material Monitoring Facility (MMF). As a result, both are acceptable as the student experiments on the fast reactor. (author)

  7. Results and Analysis of the Infrastructure Request for Information (DE-SOL-0008318)

    Energy Technology Data Exchange (ETDEWEB)

    Heidrich, Brenden John [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-07-01

    The Department of Energy (DOE) Office of Nuclear Energy (NE) released a request for information (RFI) (DE-SOL-0008318) for “University, National Laboratory, Industry and International Input on Potential Office of Nuclear Energy Infrastructure Investments” on April 13, 2015. DOE-NE solicited information on five specific types of capabilities as well as any others suggested by the community. The RFI proposal period closed on June 19, 2015. From the 26 responses, 34 individual proposals were extracted. Eighteen were associated with a DOE national laboratory, including Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Idaho National Laboratory (INL), Los Alamos National Laboratory (LANL), Pacific Northwest National Laboratory (PNNL) and Sandia National Laboratory (SNL). Oak Ridge National Laboratory (ORNL) was referenced in a proposal as a proposed capability location, although the proposal did not originate with ORNL. Five US universities submitted proposals (Massachusetts Institute of Technology, Pennsylvania State University, Rensselaer Polytechnic Institute, University of Houston and the University of Michigan). Three industrial/commercial institutions submitted proposals (AREVA NP, Babcock and Wilcox (B&W) and the Electric Power Research Institute (EPRI)). Eight major themes emerged from the submissions as areas needing additional capability or support for existing capabilities. Two submissions supported multiple areas. The major themes are: Advanced Manufacturing (AM), High Performance Computing (HPC), Ion Irradiation with X-Ray Diagnostics (IIX), Ion Irradiation with TEM Visualization (IIT), Radiochemistry Laboratories (RCL), Test Reactors, Neutron Sources and Critical Facilities (RX) , Sample Preparation and Post-Irradiation Examination (PIE) and Thermal-Hydraulics Test Facilities (THF).

  8. The legal imperative to protect critical energy infrastructure

    Energy Technology Data Exchange (ETDEWEB)

    Shore, J.J.M.

    2008-03-15

    Canada's critical infrastructure is comprised of energy facilities, communications centres, finance, health care, food, government and transportation sectors. All sectors face a range of physical or cyber threats from terrorism and natural phenomenon. Failures or disruptions in the sectors can cascade through other systems and disrupt essential services. The power outage in 2003 demonstrated gaps in North America's emergency preparedness. In 2006, al-Qaida called for terrorist attacks on North American oil fields and pipelines, specifically targeting Canada. Studies have confirmed that Canada is vulnerable to attacks on energy infrastructure. Government agencies and the private sector must work ensure the safety of Canada's energy infrastructure, as the primary responsibility of government is the protection of its citizenry. The fulfilment of the government's commitment to national security cannot be achieved without protecting Canada's critical energy infrastructure. However, Canada has not yet provided a framework linking federal government with critical infrastructures, despite the fact that a draft strategy has been under development for several years. It was concluded that governments and the private sector should work together to reduce risks, protect the public, and secure the economy. National security litigation against the government and legal imperatives for energy facility owners and operators were also reviewed. 98 refs., 20 figs.

  9. Reliability design of a critical facility: An application of PRA methods

    International Nuclear Information System (INIS)

    Souza Vieira Neto, A.; Souza Borges, W. de

    1987-01-01

    Although a general agreement concerning the enforcement of reliability (probabilistic) design criteria for nuclear utilities is yet to be achieved. PRA methodology can still be used successfully as a project design and review tool, aimed at improving system's prospective performance or minimizing expected accident consequences. In this paper, the potential of such an application of PRA methods is examined in the special case of a critical design project currently being developed in Brazil. (orig.)

  10. Use of compensation and incentives in siting low-level radioactive waste disposal facilities

    International Nuclear Information System (INIS)

    1985-04-01

    This report assumes that local opposition is a critical issue in siting low-level radioactive waste disposal facilities. Although it recognizes the importance of local health and safety concerns, this report only addresses the economic issues facing local officials in the siting process. Finding ways to overcome local opposition through economic compensation and incentives is a basic step in the waste facility siting process. The report argues that the use of these compensation and incentive mechanisms can help achieve greater local acceptance of waste facilities and also help ease the economic burdens that many communities bear when they agree to host a low-level waste disposal facility. The growing national need for low-level radioactive waste disposal facilities requires that state and local planning agencies develop creative new procedures for siting facilities, procedures that are sensitive to local perceptions and effects

  11. Criticality analysis for weapon disassembly at the Pantex-Plant part II: Staging

    International Nuclear Information System (INIS)

    Knief, R.A.

    1997-01-01

    This paper very briefly describes criticality investigations for nuclear weapon dismantlement at the Pantex Plant. The investigations performed were for pit staging, and build on previous criticality calculations for single pits. The KENO and MCNP computer models were used for pit and container combinations. Scenarios were based on administrative limits and actual or potential physical conditions in the facilities. Essentially all of the pit configurations modeled were subcritical by a substantial amount. It was concluded that a critical configuration involving pit/container combinations is not credible

  12. Increasing Registered Nurse Retention Using Mentors in Critical Care Services.

    Science.gov (United States)

    Schroyer, Coreena C; Zellers, Rebecca; Abraham, Sam

    2016-01-01

    Recruiting and training 1 newly hired registered nurse can cost thousands of dollars. With a high percentage of these newly hired nurses leaving their first place of employment within their first year, the financial implications may be enormous. It is imperative that health care facilities invest in recruiting and retention programs that retain high-quality nurses. Mentorship programs in retaining and easing the transition to practice for new graduate nurses, re-entry nurses, and nurses new to a specialty area are critical in nurse retention. Discussion in this study includes the effect of implementing a mentor program into the critical care services area of a 325-bed not-for-profit community hospital in northern Indiana. Based on this study, nurses with a mentor were retained at a 25% higher rate than those not mentored. Implementation of a mentor program reduced the training cost to the facility and increased retention and morale.

  13. Using Gamification to Raise Awareness of Cyber Threats to Critical National Infrastructure

    OpenAIRE

    Cook, Allan; Smith, Richard; Maglaras, Leandros; Janicke, Helge

    2016-01-01

    Linked to the SCIPS tabletop game Senior executives of critical national infrastructure facilities face competing requirements for investment budgets. Whilst the impact of a cyber attack upon such utilities is potentially catastrophic, the risks to continued operations from failing to upgrade ageing infrastructure, or not meeting mandated regulatory regimes, are considered higher given the demonstrable impact of such circumstances. As cyber attacks on critical national infrastructure remai...

  14. Development of computer model for radionuclide released from shallow-land disposal facility

    International Nuclear Information System (INIS)

    Suganda, D.; Sucipta; Sastrowardoyo, P.B.; Eriendi

    1998-01-01

    Development of 1-dimensional computer model for radionuclide release from shallow land disposal facility (SLDF) has been done. This computer model is used for the SLDF facility at PPTA Serpong. The SLDF facility is above 1.8 metres from groundwater and 150 metres from Cisalak river. Numerical method by implicit method of finite difference solution is chosen to predict the migration of radionuclide with any concentration.The migration starts vertically from the bottom of SLDF until the groundwater layer, then horizontally in the groundwater until the critical population group. Radionuclide Cs-137 is chosen as a sample to know its migration. The result of the assessment shows that the SLDF facility at PPTA Serpong has the high safety criteria. (author)

  15. Development of Radiation-Hardening Ceramic Composites for Fusion Applications

    International Nuclear Information System (INIS)

    Don Steiner

    2004-01-01

    This Progress Report describes work performed as a collaborative effort between Rensselaer Polytechnic Institute (RPI) and Oak Ridge National Laboratory (ORNL). This research is focused in four areas considered to be critical issues for using SiC fiber-reinforced SiC matrix composites (SiC/SiC) as structural materials in a fusion environment: (1) Calculation of the critical dose and temperature for amorphization of SiC by using the TRIM computer code to analyze ORNL and literature data; (2) Measurement of irradiation-induced creep in monolithic SiC or stoichiometric SiC fibers; (3) Determining the effects of high-temperature irradiation on monolithic SiC as part of ORNL's METS experiment; and (4) Gauging the effectiveness of polymer impregnation pyrolysis in improving SiC/SiC composite hermicity. Progress in each area is described, as well as plans for next year

  16. Criticality evaluation for the 233-S decontamination and decommissioning project

    International Nuclear Information System (INIS)

    1996-08-01

    This criticality evaluation document analyzes the potential of a criticality event as a result of decontaminating and decommissioning the 233-S Plutonium Concentration Facility. These calculations supplement the previous set of calculations performed under this same contract, which were performed on March 13, 1996. These calculations were performed using the same MCNP computer code as for the previous set; the validation calculations performed then are valid for this set as well. Hand calculations, using the method of Solid Angle, were also developed

  17. Optimal Facility Location Tool for Logistics Battle Command (LBC)

    Science.gov (United States)

    2015-08-01

    64 Appendix B. VBA Code . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 Appendix C. Story...should city planners have located emergency service facilities so that all households (the demand) had equal access to coverage?” The critical...programming language called Visual Basic for Applications ( VBA ). CPLEX is a commercial solver for linear, integer, and mixed integer linear programming problems

  18. Missed vaccination opportunities at a secondary health facility in ...

    African Journals Online (AJOL)

    The study aimed to identify missed vaccination visits and the associated factors in children presenting at the general out-patient clinic of a secondary health facility in Ilorin, Nigeria. Method: Through a descriptive cross-sectional study, the vaccination data of all children seen at the out-patient clinic were critically reviewed ...

  19. Applications of PRA in nuclear criticality safety

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1992-01-01

    Traditionally, criticality accident prevention at Los Alamos has been based on a thorough review and understanding of proposed operations of changes to operations, involving both process supervision and criticality safety staff. The outcome of this communication was usually an agreement, based on professional judgement, that certain accident sequences were credible and had to be reduced in likelihood either by administrative controls or by equipment design and others were not credible, and thus did not warrant expenditures to further reduce their likelihood. The extent of analysis and documentation was generally in proportion to the complexity of the operation but did not include quantified risk assessments. During the last three years nuclear criticality safety related Probabilistic Risk Assessments (PRAs) have been preformed on operations in two Los Alamos facilities. Both of these were conducted in order to better understand the cost/benefit aspects of PRA's as they apply to largely ''hands-on'' operations with fissile material for which human errors or equipment failures significant to criticality safety are both rare and unique. Based on these two applications and an appreciation of the historical criticality accident record (frequency and consequences) it is apparent that quantified risk assessments should be performed very selectively

  20. Nuclear criticality safety specialist training and qualification programs

    International Nuclear Information System (INIS)

    Hopper, C.M.

    1993-01-01

    Since the beginning of the Nuclear Criticality Safety Division of the American Nuclear Society (ANS) in 1967, the nuclear criticality safety (NCS) community has sought to provide an exchange of information at a national level to facilitate the education and development of NCS specialists. In addition, individual criticality safety organizations within government contractor and licensed commercial nonreactor facilities have developed training and qualification programs for their NCS specialists. However, there has been substantial variability in the content and quality of these program requirements and personnel qualifications, at least as measured within the government contractor community. The purpose of this paper is to provide a brief, general history of staff training and to describe the current direction and focus of US DOE guidance for the content of training and qualification programs designed to develop NCS specialists

  1. Do Urban Rail Transit Facilities Affect Housing Prices? Evidence from China

    Directory of Open Access Journals (Sweden)

    Xu Zhang

    2016-04-01

    Full Text Available Urban rail transit facilities play a critical role in citizen’s social activities (e.g., residence, work and education. Using panel data on housing prices and urban rail transit facilities for 35 Chinese cities for 2002 to 2013, this study constructs a panel data model to evaluate the effect of rail transit facilities on housing prices quantitatively. A correlation test reveals significant correlations between housing prices and rail transit facilities. Empirical results demonstrate that rail transit facilities can markedly elevate real estate prices. Quantitatively, a 1% increase in rail transit mileage improves housing prices by 0.0233%. The results highlight the importance of other factors (e.g., per capita GDP, land price, investment in real estate and population density in determining housing prices. We also assess the effects of expectations of new rail transit lines on housing prices, and the results show that expectation effects are insignificant. These findings encourage Chinese policy makers to take rail transit facilities into account in achieving sustainable development of real estate markets.

  2. Engineering and technology in the deconstruction of nuclear materials production facilities

    International Nuclear Information System (INIS)

    Kingsley, R.S.; Reynolds, W.E.; Heffner, D.C.

    1996-01-01

    Technology and equipment exist to support nuclear facility deactivation, decontamination, and decommissioning. In reality, this statement is not surprising because the nuclear industry has been decontaminating and decommissioning production plants for decades as new generations of production technology were introduced. Since the 1950s, the Babcock and Wilcox Company (B ampersand W) has operated a number of nuclear materials processing facilities to manufacture nuclear fuel for the commercial power industry and the U.S. Navy. These manufacturing facilities included a mixed oxide (PuO 2 -UO 2 ) nuclear fuel manufacturing plant, low- and high-enriched uranium (HEU/LEU) chemical and fuel plants, and fuel assembly plants. In addition, B ampersand W designed and build a major nuclear research center in Lynchburg, Virginia, to support these nuclear fuel manufacturing activities and to conduct nuclear power research. These nuclear research facilities included two research reactors, a hot-cell complex for nuclear materials research, four critical experiment facilities, and a plutonium fuels research and development facility. This article describes the B ampersand W deactivation, decomtanimation, and decommisioning program

  3. Applications of PRA in nuclear criticality safety

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1992-01-01

    Traditionally, criticality accident prevention at Los Alamos National Laboratory (LANL) has been based on a thorough review and understanding of proposed operations or changes to operations involving both process supervision and criticality safety staff. The outcome of this communication was usually an agreement, based on professional judgment, that certain accident sequences were credible and had to be precluded by design; others were incredible and thus did not warrant expenditures to further reduce their likelihood. The extent of documentation was generally in proportion to the complexity of the operation but never as detailed as that associated with quantified risk assessments. During the last 3 yr, nuclear criticality safety-related probabilistic risk assessments (PRAs) have been performed on operations in two LANL facilities. Both of these were conducted in order to better understand the cost/benefit aspects of PRAs as they apply to largely hands-on operations with fissile material

  4. Realistic evaluation of new fuel storage criticality

    International Nuclear Information System (INIS)

    Hamasaki, M.; Itahara, K.; Shimada, S.; Etsu, M.; Kuroda, M.; Watanabe, H.; Kuragasaki, M.; Kawaguchi, K.

    1987-01-01

    In the criticality safety design of dry fuel storage facility, the optimum moderation condition which appears at rather low water density (0.05 ∼ 0.1 gH 2 O/cc) has been a barrier to incorporate economical design. In this study we have shown that this optimum moderation does not occur, even in fire-fighting with water supply system, by criticality analyses and experiments. Evaluated critical amount of water in the space has found itself far from the level estimated actually attainable through the experiment with fire-fighting system. In another experiment we have taken the holograms of 0.003 ∼ 0.35 g/cc water which was realized in the very narrow space by the nozzle manufactured for this special purpose. Those holograms demonstrate that the droplets in the low density water are more closely packed than we anticipate they are. (author)

  5. Thermal-Hydraulic Analysis of SWAMUP Facility Using ATHLET-SC Code

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Zidi; Cao, Zhen; Liu, Xiaojing, E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai (China)

    2015-03-16

    During the loss of coolant accident (LOCA) of supercritical water-cooled reactor (SCWR), the pressure in the reactor system will undergo a rapid decrease from the supercritical pressure to the subcritical condition. This process is called trans-critical transients, which is of crucial importance for the LOCA analysis of SCWR. In order to simulate the trans-critical transient, a number of system codes for SCWR have been developed up to date. However, the validation work for the trans-critical models in these codes is still missing. The test facility Supercritical WAter MUltiPurpose loop (SWAMUP) with 2 × 2 rod bundle in Shanghai Jiao Tong University (SJTU) will be applied to provide test data for code validation. Some pre-test calculations are important and necessary to show the feasibility of the experiment. In this study, trans-critical transient analysis is performed for the SWAMUP facility with the system code ATHLET-SC, which is modified in SJTU, for supercritical water system. This paper presents the system behavior, e.g., system pressure, coolant mass flow, cladding temperature during the depressurization. The effects of some important parameters such as heating power, depressurization rate on the system characteristics are also investigated in this paper. Additionally, some sensitivities study of the code models, e.g., heat transfer coefficient, critical heat flux correlation are analyzed and discussed. The results indicate that the revised system code ATHLET-SC is capable of simulating thermal-hydraulic behavior during the trans-critical transient. According to the results, the cladding temperature during the transient is kept at a low value. However, the pressure difference of the heat exchanger after depressurization could reach 6 MPa, which should be considered in the experiment.

  6. Molecular Science Computing Facility Scientific Challenges: Linking Across Scales

    Energy Technology Data Exchange (ETDEWEB)

    De Jong, Wibe A.; Windus, Theresa L.

    2005-07-01

    The purpose of this document is to define the evolving science drivers for performing environmental molecular research at the William R. Wiley Environmental Molecular Sciences Laboratory (EMSL) and to provide guidance associated with the next-generation high-performance computing center that must be developed at EMSL's Molecular Science Computing Facility (MSCF) in order to address this critical research. The MSCF is the pre-eminent computing facility?supported by the U.S. Department of Energy's (DOE's) Office of Biological and Environmental Research (BER)?tailored to provide the fastest time-to-solution for current computational challenges in chemistry and biology, as well as providing the means for broad research in the molecular and environmental sciences. The MSCF provides integral resources and expertise to emerging EMSL Scientific Grand Challenges and Collaborative Access Teams that are designed to leverage the multiple integrated research capabilities of EMSL, thereby creating a synergy between computation and experiment to address environmental molecular science challenges critical to DOE and the nation.

  7. Criticality of a 237Np sphere

    International Nuclear Information System (INIS)

    Sanchez, Rene; Loaiza, David; Kimpland, Robert; Hayes, David; Cappiello, Charlene; Chadwick, Mark

    2006-01-01

    For the past five years, scientists at Los Alamos National Laboratory have mounted an unprecedented effort to get a better estimate of the critical mass of 237 Np. To accomplish this task, a 6-kg neptunium sphere was recently cast at the Chemical and Metallurgy Research (CMR) facility, which is part of the Los Alamos National Laboratory. The neptunium sphere was clad with tungsten and nickel to reduce the dose rates from the 310 keV gamma rays from the first daughter of neptunium, namely, 233 Pa. 237 Np is a byproduct of power production in nuclear reactors. It is primarily produced by successive neutron captures in 235 U or through the (n,2n) reaction in 238 U. These nuclear reactions lead to the production of 237 U, which decays by beta emission into 237 Np, namely, 235 U(n,γ) 236 U, 236 U(n,γ) 237 U→β→ 237 Np, 238 U (n,2n) 237 U→β→ 237 Np. It is estimated that a typical 1000 MW(e) produces on the order of 12 to 13 kg of neptunium in a year. Some of this neptunium in irradiated fuel elements has been separated and is presently stored in containers in a liquid form. This method of storage is quite adequate because the fission cross section for 237 Np at thermal energies is quite low and any moderation of the neutron population by diluting the configurations with water would increase the critical mass to infinity. However, for long term storage, the neptunium liquid solutions must be converted into oxides and metals because these form are less movable and less likely to leak out of containers. Metals and oxides made out of neptunium have finite critical masses but there is a great uncertainty about these values because of the lack of experimental criticality data. Knowing precisely the critical mass of neptunium not only will help to validate mass storage limits or optimize storage configurations for safe disposition of these materials, but will also save thousands of dollars in transportation and disposition costs. The experimental results presented in

  8. Review and assessment of nuclear facilities by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on reviewing and assessing the various safety related submissions made by the operator of a nuclear facility at different stages (siting, design, construction, commissioning, operation and decommissioning or closure) in the facility's lifetime to determine whether the facility complies with the applicable safety objectives and requirements. This Safety Guide covers the review and assessment of submissions in relation to the safety of nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Objectives, management, planning and organizational matters relating to the review and assessment process are presented in Section 2. Section 3 deals with the bases for decision making and conduct of the review and assessment process. Section 4 covers aspects relating to the assessment of this process. The Appendix provides a generic list of topics to be covered in the review and assessment process

  9. Review and assessment of nuclear facilities by the regulatory body. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    The purpose of this Safety Guide is to provide recommendations for regulatory bodies on reviewing and assessing the various safety related submissions made by the operator of a nuclear facility at different stages (siting, design, construction, commissioning, operation and decommissioning or closure) in the facility's lifetime to determine whether the facility complies with the applicable safety objectives and requirements. This Safety Guide covers the review and assessment of submissions in relation to the safety of nuclear facilities such as: enrichment and fuel manufacturing plants. Nuclear power plants. Other reactors such as research reactors and critical assemblies. Spent fuel reprocessing plants. And facilities for radioactive waste management, such as treatment, storage and disposal facilities. This Safety Guide also covers issues relating to the decommissioning of nuclear facilities, the closure of waste disposal facilities and site rehabilitation. Objectives, management, planning and organizational matters relating to the review and assessment process are presented in Section 2. Section 3 deals with the bases for decision making and conduct of the review and assessment process. Section 4 covers aspects relating to the assessment of this process. The Appendix provides a generic list of topics to be covered in the review and assessment process

  10. (Some) Present and planned facilities for the R&D support of MYRRHA

    International Nuclear Information System (INIS)

    Schuurmans, Paul

    2013-01-01

    MYRRHA (Multipurpose hYbrid Research Reactor for High-tech Applications) - Accelerator Driven System is a flexible fast spectrum irradiation facility able to operate in sub-critical and critical mode for material and fuel developments for GEN IV and fusion reactors and in a back-up role for radioisotopes production. Some presented topics: • Structural Materials; • Advanced Fuel; • Thermal-hydraulics; • Reactor components; • Instrumentation; • Coolant chemistry control; • Operational control & Neutronics

  11. Experience with performance based training of nuclear criticality safety engineers

    International Nuclear Information System (INIS)

    Taylor, R.G.

    1993-01-01

    For non-reactor nuclear facilities, the U.S. Department of Energy (DOE) does not require that nuclear criticality safety engineers demonstrate qualification for their job. It is likely, however, that more formalism will be required in the future. Current DOE requirements for those positions which do have to demonstrate qualification indicate that qualification should be achieved by using a systematic approach such as performance based training (PBT). Assuming that PBT would be an acceptable mechanism for nuclear criticality safety engineer training in a more formal environment, a site-specific analysis of the nuclear criticality safety engineer job was performed. Based on this analysis, classes are being developed and delivered to a target audience of newer nuclear criticality safety engineers. Because current interest is in developing training for selected aspects of the nuclear criticality safety engineer job, the analysis is incompletely developed in some areas

  12. Comparison of the dose evaluation methods for criticality accident

    International Nuclear Information System (INIS)

    Shimizu, Yoshio; Oka, Tsutomu

    2004-01-01

    The improvement of the dose evaluation method for criticality accidents is important to rationalize design of the nuclear fuel cycle facilities. The source spectrums of neutron and gamma ray of a criticality accident depend on the condition of the source, its materials, moderation, density and so on. The comparison of the dose evaluation methods for a criticality accident is made. Some methods, which are combination of criticality calculation and shielding calculation, are proposed. Prompt neutron and gamma ray doses from nuclear criticality of some uranium systems have been evaluated as the Nuclear Criticality Slide Rule. The uranium metal source (unmoderated system) and the uranyl nitrate solution source (moderated system) in the rule are evaluated by some calculation methods, which are combinations of code and cross section library, as follows: (a) SAS1X (ENDF/B-IV), (b) MCNP4C (ENDF/B-VI)-ANISN (DLC23E or JSD120), (c) MCNP4C-MCNP4C (ENDF/B-VI). They have consisted of criticality calculation and shielding calculation. These calculation methods are compared about the tissue absorbed dose and the spectrums at 2 m from the source. (author)

  13. Neutron irradiation facilities for fission and fusion reactor materials studies

    International Nuclear Information System (INIS)

    Rowcliffe, A.F.

    1985-01-01

    The successful development of energy-conversion machines based upon nuclear fission or fusion reactors is critically dependent upon the behavior of the engineering materials used to construct the full containment and primary heat extraction systems. The development of radiation damage-resistant materials requires irradiation testing facilities which reproduce, as closely as possible, the thermal and neutronic environment expected in a power-producing reactor. The Oak Ridge National Laboratory (ORNL) reference core design for the Center for Neutron Research (CNR) reactor provides for instrumented facilities in regions of both hard and mixed neutron spectra, with substantially higher fluxes than are currently available. The benefits of these new facilities to the development of radiation damage resistant materials are discussed in terms of the major US fission and fusion reactor programs

  14. Sustainability of utility-scale solar energy: Critical environmental concepts

    Science.gov (United States)

    Hernandez, R. R.; Moore-O'Leary, K. A.; Johnston, D. S.; Abella, S.; Tanner, K.; Swanson, A.; Kreitler, J.; Lovich, J.

    2017-12-01

    Renewable energy development is an arena where ecological, political, and socioeconomic values collide. Advances in renewable energy will incur steep environmental costs to landscapes in which facilities are constructed and operated. Scientists - including those from academia, industry, and government agencies - have only recently begun to quantify trade-off in this arena, often using ground-mounted, utility-scale solar energy facilities (USSE, ≥ 1 megawatt) as a model. Here, we discuss five critical ecological concepts applicable to the development of more sustainable USSE with benefits over fossil-fuel-generated energy: (1) more sustainable USSE development requires careful evaluation of trade-offs between land, energy, and ecology; (2) species responses to habitat modification by USSE vary; (3) cumulative and large-scale ecological impacts are complex and challenging to mitigate; (4) USSE development affects different types of ecosystems and requires customized design and management strategies; and (5) long-term ecological consequences associated with USSE sites must be carefully considered. These critical concepts provide a framework for reducing adverse environmental impacts, informing policy to establish and address conservation priorities, and improving energy production sustainability.

  15. Criticality analysis in uranium enrichment plant

    International Nuclear Information System (INIS)

    Okamoto, Tsuyoshi; Kiyose, Ryohei

    1977-01-01

    In a large scale uranium enrichment plant, uranium inventory in cascade rooms is not very large in quantity, but the facilities dealing with the largest quantity of uranium in that process are the UF 6 gas supply system and the blending system for controlling the product concentration. When UF 6 spills out of these systems, the enriched uranium is accumulated, and the danger of criticality accident is feared. If a NaF trap is placed at the forestage of waste gas treatment system, plenty of UF 6 and HF are adsorbed together in the NaF trap. Thus, here is the necessity of checking the safety against criticality. Various assumptions were made to perform the computation surveying the criticality of the system composed of UF 6 and HF adsorbed on NaF traps with WIMS code (transport analysis). The minimum critical radius resulted in about 53 cm in case of 3.5% enriched fuel for light water reactors. The optimum volume ratio of fissile material in the double salt UF 6 .2NaF and NaF.HF is about 40 vol. %. While, criticality survey computation was also made for the annular NaF trap having the central cooling tube, and it was found that the effect of cooling tube radius did not decrease the multiplication factor up to the cooling tube radius of about 5 cm. (Wakatsuki, Y.)

  16. On the Viability of Supporting Institutional Sharing of Remote Laboratory Facilities

    Science.gov (United States)

    Lowe, David; Dang, Bridgette; Daniel, Keith; Murray, Stephen; Lindsay, Euan

    2015-01-01

    Laboratories are generally regarded as critical to engineering education, and yet educational institutions face significant challenges in developing and maintaining high-quality laboratory facilities. Remote laboratories are increasingly being explored as a partial solution to this challenge, with research showing that--for the right learning…

  17. Characterization and consequences from CEA nuclear fuel cycle facilities effluents releases - 1995 up to 2007 period

    International Nuclear Information System (INIS)

    Ferreira, Nelson Luiz Dias; Fonseca, Lizandra Pereira de Souza

    2009-01-01

    Discharges to the environment of airborne and/or liquid radioactive effluents from the normal operation of nuclear facilities can become a potential source of radiation exposure to humans. The highest exposed members of the public are defined as the critical group. The requirements for the control and monitoring of radioactive discharges to the environment and the degree of environmental monitoring required are linked to the assessed critical group dose. The assessed dose can be compared to dose constraint, which is a fraction of the annual effective dose to members of the public, as well as the level of exemption specified by the National Commission for Nuclear Energy (CNEN). Effluents releases from the Centro Experimental Aramar (CEA) facilities are registered and described at CEA Effluent Report, semestrally sent to CNEN. Basically, that report provides information related to the type and the quantity of chemical and radioactive substances released to the environment due the routine operation of CEA nuclear fuel cycle facilities (LEI - Isotopic Enrichment Laboratory, USIDE - Pilot Plant for Industrial Verification of Uranium Enrichment and LABMAT - Nuclear Materials Laboratory). CEA Annual Effluent Report includes assessment of the annual effective doses for members of the critical group for the CEA site. This work presents the characterization of the radioactive release source terms and a historical of the critical group annual doses from 1995 up to 2007. (author)

  18. Validating analysis methodologies used in burnup credit criticality calculations

    International Nuclear Information System (INIS)

    Brady, M.C.; Napolitano, D.G.

    1992-01-01

    The concept of allowing reactivity credit for the depleted (or burned) state of pressurized water reactor fuel in the licensing of spent fuel facilities introduces a new challenge to members of the nuclear criticality community. The primary difference in this analysis approach is the technical ability to calculate spent fuel compositions (or inventories) and to predict their effect on the system multiplication factor. Isotopic prediction codes are used routinely for in-core physics calculations and the prediction of radiation source terms for both thermal and shielding analyses, but represent an innovation for criticality specialists. This paper discusses two methodologies currently being developed to specifically evaluate isotopic composition and reactivity for the burnup credit concept. A comprehensive approach to benchmarking and validating the methods is also presented. This approach involves the analysis of commercial reactor critical data, fuel storage critical experiments, chemical assay isotopic data, and numerical benchmark calculations

  19. An Adaptive Critic Approach to Reference Model Adaptation

    Science.gov (United States)

    Krishnakumar, K.; Limes, G.; Gundy-Burlet, K.; Bryant, D.

    2003-01-01

    Neural networks have been successfully used for implementing control architectures for different applications. In this work, we examine a neural network augmented adaptive critic as a Level 2 intelligent controller for a C- 17 aircraft. This intelligent control architecture utilizes an adaptive critic to tune the parameters of a reference model, which is then used to define the angular rate command for a Level 1 intelligent controller. The present architecture is implemented on a high-fidelity non-linear model of a C-17 aircraft. The goal of this research is to improve the performance of the C-17 under degraded conditions such as control failures and battle damage. Pilot ratings using a motion based simulation facility are included in this paper. The benefits of using an adaptive critic are documented using time response comparisons for severe damage situations.

  20. A discussion on establishment of GIP management system for food irradiation facilities

    International Nuclear Information System (INIS)

    Lu Jiang; Shi Hua; Li Ruisong; Li Shurong; Zhou Hongjie; Ha Yiming

    2005-01-01

    This article analyses the hazard factors and selects Critical Control Point (CCP) for food irradiation process (including staff, facilities and processing) using HACCP version. The principles and method of GIP system for food irradiation plant are also discussed. (authors)

  1. Emission Facilities - Erosion & Sediment Control Facilities

    Data.gov (United States)

    NSGIC Education | GIS Inventory — An Erosion and Sediment Control Facility is a DEP primary facility type related to the Water Pollution Control program. The following sub-facility types related to...

  2. HSE's safety assessment principles for criticality safety

    International Nuclear Information System (INIS)

    Simister, D N; Finnerty, M D; Warburton, S J; Thomas, E A; Macphail, M R

    2008-01-01

    The Health and Safety Executive (HSE) published its revised Safety Assessment Principles for Nuclear Facilities (SAPs) in December 2006. The SAPs are primarily intended for use by HSE's inspectors when judging the adequacy of safety cases for nuclear facilities. The revised SAPs relate to all aspects of safety in nuclear facilities including the technical discipline of criticality safety. The purpose of this paper is to set out for the benefit of a wider audience some of the thinking behind the final published words and to provide an insight into the development of UK regulatory guidance. The paper notes that it is HSE's intention that the Safety Assessment Principles should be viewed as a reflection of good practice in the context of interpreting primary legislation such as the requirements under site licence conditions for arrangements for producing an adequate safety case and for producing a suitable and sufficient risk assessment under the Ionising Radiations Regulations 1999 (SI1999/3232 www.opsi.gov.uk/si/si1999/uksi_19993232_en.pdf). (memorandum)

  3. Administrative Aspects of the Criticality Controls Used in Programmes for Basic Criticality Research, Reactor Development and Materials Processing

    Energy Technology Data Exchange (ETDEWEB)

    Wood, D. P.; Giessing, D. F. [Operational Safety Division, USAEC Albuquerque Operations Office, NM (United States)

    1966-05-15

    This paper describes the administrative and procedural aspects of criticality controls used by a field office of the United States Atomic Energy Commission in programmes that include reactor criticals, research and materials testing reactors, and power reactor development. Situations encountered include handling, storing, and processing large quantities of uranium-235 and plutonium-239 of various configurations and compositions in laboratories and operations which gather basic criticality data, processing of fissile material, and varied reactor research and development, programmes including fuel materials. Similar situations exist for uranium-233 and plutonium-238 on a smaller laboratory scale. The administrative controls and interactions of the USAEC field office and the operating contractors, who operate these installations for the USAEC, are outlined. Also, the purpose and scope of the direct examination by USAEC personnel of these contractor facilities are analysed. The programme has been in effect for three years and is believed to be successful in maintaining efficient operations and an acceptable low level of risk of inadvertent criticality. Success of this programme is in good measure due to the close working relationship between the staffs of the USAEC field office and the operating contractors. (author)

  4. Critical infrastructure – content, structure and problems of its protection

    Directory of Open Access Journals (Sweden)

    Ladislav Hofreiter

    2014-06-01

    Full Text Available Security, economic and social stability of the country, its functionality but also protecting the lives and property of citizens are dependent on the proper functioning of many infrastructure systems of state. Disruptions, lack or destruction of such systems, institutions, facilities and other services could cause disruption of social stability and national security, provoke a crisis situation or seriously affect the operation of state and local governments in crisis situations. This is known as critical infrastructure. It is in the interest of the State to the critical infrastructure effectively protected.

  5. A method for managing the storage of fissile materials using criticality indices

    International Nuclear Information System (INIS)

    Philbin, J.S.; Harms, G.A.

    1995-01-01

    This paper describes a method for criticality control at fissile material storage facilities. The method involves the use criticiality indices for storage canisters. The logic, methodology, and results for selected canisters are presented. A concept for an interactive computer program using the method is also introduced. The computer program can be used in real time (using precalulated data) to select a Criticality Index (CI) for a container when it is delivered to or packaged at a site. Criticality safety is assured by controlling the sum of the CIs at each storage location below a defined Emit value when containers are moved

  6. Criticality safety of transuranic storage arrays at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Boyd, W.A.; Fecteau, M.W.

    1993-01-01

    The Waste Isolation Pilot Plant (WIPP) facility is designed to store transuranic waste that will consist mainly of surface contaminate articles and sludge. The fissile material in the waste is predominantly 239 Pu. The waste is grouped into two categories: contact-handled waste, which will be stored in 55-gal steel drums or in steel boxes, and remote-handled waste, which will be stored in specially designed cylindrical steel canisters. To show that criticality safety will be acceptable, criticality analyses were performed to demonstrate that a large number of containers with limiting loadings of fissile material could be stored at the site and meet a k eff limit of 0.95. Criticality analyses based on the classic worst-case moderated plutonium sphere approach would severely limit the capacity for storage of waste at the facility. Therefore, these analyses use realistic or credible worst-case assumptions to better represent the actual storage situation without compromising the margin of safety. Numerous sensitivity studies were performed to determine the importance of various parameters on the criticality of the configuration. It was determined that the plutonium loading has the dominant effect on the system reactivity. Nearly all other reactivity variations from the sensitivity studies were found to be relatively small. The analysis shows that criticality of the contact-handled waste storage drums and boxes and the remote-handled canisters is prevented by restrictions on maximum fissile loading per container and on the size of handling/storage areas

  7. Present status of Monte Carlo seminar for sub-criticality safety analysis in Japan

    International Nuclear Information System (INIS)

    Sakurai, Kiyoshi

    2003-01-01

    This paper provides overview of the methods and results of a series of sub-criticality safety analysis seminars for nuclear fuel cycle facility with the Monte Carlo method held in Japan from July 2000 to July 2003. In these seminars, MCNP-4C2 system (MS-DOS version) was installed in note-type personal computers for participants. Fundamental theory of reactor physics and Monte Carlo simulation as well as the contents of the MCNP manual were lectured. Effective neutron multiplication factors and neutron spectra were calculated for some examples such as JCO deposit tank, JNC uranium solution storage tank, JNC plutonium solution storage tank and JAERI TCA core. Management for safety of nuclear fuel cycle facilities was discussed in order to prevent criticality accidents in some of the seminars. (author)

  8. Australian national networked tele-test facility for integrated systems

    Science.gov (United States)

    Eshraghian, Kamran; Lachowicz, Stefan W.; Eshraghian, Sholeh

    2001-11-01

    The Australian Commonwealth government recently announced a grant of 4.75 million as part of a 13.5 million program to establish a world class networked IC tele-test facility in Australia. The facility will be based on a state-of-the-art semiconductor tester located at Edith Cowan University in Perth that will operate as a virtual centre spanning Australia. Satellite nodes will be located at the University of Western Australia, Griffith University, Macquarie University, Victoria University and the University of Adelaide. The facility will provide vital equipment to take Australia to the frontier of critically important and expanding fields in microelectronics research and development. The tele-test network will provide state of the art environment for the electronics and microelectronics research and the industry community around Australia to test and prototype Very Large Scale Integrated (VLSI) circuits and other System On a Chip (SOC) devices, prior to moving to the manufacturing stage. Such testing is absolutely essential to ensure that the device performs to specification. This paper presents the current context in which the testing facility is being established, the methodologies behind the integration of design and test strategies and the target shape of the tele-testing Facility.

  9. Integration of facility modeling capabilities for nuclear nonproliferation analysis

    International Nuclear Information System (INIS)

    Burr, Tom; Gorensek, M.B.; Krebs, John; Kress, Reid L.; Lamberti, Vincent; Schoenwald, David; Ward, Richard C.

    2012-01-01

    Developing automated methods for data collection and analysis that can facilitate nuclearnonproliferation assessment is an important research area with significant consequences for the effective global deployment of nuclear energy. Facilitymodeling that can integrate and interpret observations collected from monitored facilities in order to ascertain their functional details will be a critical element of these methods. Although improvements are continually sought, existing facilitymodeling tools can characterize all aspects of reactor operations and the majority of nuclear fuel cycle processing steps, and include algorithms for data processing and interpretation. Assessing nonproliferation status is challenging because observations can come from many sources, including local and remote sensors that monitor facility operations, as well as open sources that provide specific business information about the monitored facilities, and can be of many different types. Although many current facility models are capable of analyzing large amounts of information, they have not been integrated in an analyst-friendly manner. This paper addresses some of these facilitymodelingcapabilities and illustrates how they could be integrated and utilized for nonproliferationanalysis. The inverse problem of inferring facility conditions based on collected observations is described, along with a proposed architecture and computer framework for utilizing facilitymodeling tools. After considering a representative sampling of key facilitymodelingcapabilities, the proposed integration framework is illustrated with several examples.

  10. Engineering evaluation/cost analysis for the 105-DR and 105-F Reactor facilities and ancillary facilities

    International Nuclear Information System (INIS)

    Coenenberg, E.T.

    1998-01-01

    This document presents the results of an engineering evaluation/cost analysis (EE/CA) that was conducted to evaluate alternatives to address final disposition of the 105-DR and 105-F Reactor Buildings (subsequently referred to as facilities), including the fuel storage basins (FSB) and below-grade portions of the reactors, excluding the reactor blocks. The reactor blocks will remain in a safe storage mode for up to 75 years as identified in the Record the Decision (ROD) (58 FR 48509) for the Environmental Impact Statement (EIS), Decommissioning of Eight Surplus Production Reactors at the Hanford Site, Richland, Washington (DOE 1992a). This EE/CA also addresses final disposition of four ancillary facilities: 116-D and 116-DR Exhaust Air Stacks, 117-DR Exhaust Filter Building, and 119-DR Exhaust Air Sample Building. The 105-DR and 105-F facilities are located in the 100-D and 100-F Areas of the Hanford Site. In November 1989, the 100 Area of the Hanford Site was placed on the U.S. Environmental Protection Agency's (EPA) National Priorities List (NPL) under the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA). The 100 Area NPL includes the 100-D Area (which includes the 100-DR site) and the 100-F Area, which are in various stages of the remediation process. It has been determined by DOE that hazardous substances in the 105-DR, 105-F, and the four ancillary facilities may present a potential threat to human health or the environment, and that a non-time critical removal action at these facilities is warranted. To help determine the most appropriate action, DOE, in cooperation with the Washington State Department of Ecology (Ecology) and the EPA, has prepared this EE/CA. The scope of the evaluation includes the 105-DR and 105-F facilities and the four ancillary facilities. The 116-DR and 117-DR facilities are located within the boundaries of the 105-DR Large Sodium Fire Facility Treatment, Storage, and Disposal (TSD) unit, which is

  11. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

    In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

  12. Predictors of Mortality in a Critical Care Unit in South Western Kenya

    African Journals Online (AJOL)

    Abstract. Background: Critical care in developing countries has been ... which may impact the quality of care. Hospitals also ... and referral facility located in South Western Kenya in Bomet .... p=0.01). As regards end of life care; 40.4% of those.

  13. Development of radiation safety monitoring system at gamma greenhouse gamma facility

    International Nuclear Information System (INIS)

    Hairul Nizam Idris; Azimawati Ahmad, Ahmad Zaki Hussain; Ahmad Fairuz Mohd Nasir

    2009-01-01

    This paper is discussing about installation of radiation safety monitoring system at Gamma Greenhouse Gamma facility, Agrotechnology and Bioscience Division (BAB). This facility actually is an outdoor type irradiation facility, which first in Nuclear Malaysia and the only one in Malaysia. Source Cs-137 (801 Curie) was use as radiation source and it located at the centre of 30 metres diameter size of open irradiation area. The radiation measurement and monitoring system to be equipped in this facility were required the proper equipment and devices, specially purpose for application at outside of building. Research review, literature study and discussion with the equipment manufacturers was being carried out, in effort to identify the best system should be developed. Factors such as tropical climate, environment surrounding and security were considered during selecting the proper system. Since this facility involving with panoramic radiation type, several critical and strategic locations have been fixed with radiation detectors, up to the distance at 200 meter from the radiation source. Apart from that, this developed system also was built for capable to provide the online real-time reading (using internet). In general, it can be summarized that the radiation safety monitoring system for outdoor type irradiation facility was found much different and complex compared to the system for indoor type facility. Keyword: radiation monitoring, radiation safety, Gamma Greenhouse, outdoor irradiation facility, panoramic radiation. (Author)

  14. A study on safety analysis methodology in spent fuel dry storage facility

    Energy Technology Data Exchange (ETDEWEB)

    Che, M. S.; Ryu, J. H.; Kang, K. M.; Cho, N. C.; Kim, M. S. [Hanyang Univ., Seoul (Korea, Republic of)

    2004-02-15

    Collection and review of the domestic and foreign technology related to spent fuel dry storage facility. Analysis of a reference system. Establishment of a framework for criticality safety analysis. Review of accident analysis methodology. Establishment of accident scenarios. Establishment of scenario analysis methodology.

  15. Research on Methodology to Prioritize Critical Digital Assets based on Nuclear Risk Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Wonjik; Kwon, Kookheui; Kim, Hyundoo [Korea Institute of Nuclear Nonproliferation and Control, Daejeon (Korea, Republic of)

    2016-10-15

    Digital systems are used in nuclear facilities to monitor and control various types of field devices, as well as to obtain and store vital information. Therefore, it is getting important for nuclear facilities to protect digital systems from cyber-attack in terms of safety operation and public health since cyber compromise of these systems could lead to unacceptable radiological consequences. Based on KINAC/RS-015 which is a cyber security regulatory standard, regulatory activities for cyber security at nuclear facilities generally focus on critical digital assets (CDAs) which are safety, security, and emergency preparedness related digital assets. Critical digital assets are estimated over 60% among all digital assets in a nuclear power plant. Therefore, it was required to prioritize critical digital assets to improve efficiency of regulation and implementation. In this paper, the research status on methodology development to prioritize critical digital assets based on nuclear risk assessment will be introduced. In this paper, to derive digital asset directly affect accident, PRA results (ET, FT, and minimal cut set) are analyzed. According to result of analysis, digital systems related to CD are derived ESF-CCS (safety-related component control system) and Process-CCS (non-safety-related component control system) as well as Engineered Safety Features Actuation System (ESFAS). These digital assets can be identified Vital Digital Asset (VDA). Hereafter, to develop general methodology which was identified VDA related to accident among CDAs, (1) method using result of minimal cut set in PRA model will be studied and (2) method quantifying result of Digital I and C PRA which is performed to reflect all digital cabinet related to system in FT will be studied.

  16. Research on Methodology to Prioritize Critical Digital Assets based on Nuclear Risk Assessment

    International Nuclear Information System (INIS)

    Kim, Wonjik; Kwon, Kookheui; Kim, Hyundoo

    2016-01-01

    Digital systems are used in nuclear facilities to monitor and control various types of field devices, as well as to obtain and store vital information. Therefore, it is getting important for nuclear facilities to protect digital systems from cyber-attack in terms of safety operation and public health since cyber compromise of these systems could lead to unacceptable radiological consequences. Based on KINAC/RS-015 which is a cyber security regulatory standard, regulatory activities for cyber security at nuclear facilities generally focus on critical digital assets (CDAs) which are safety, security, and emergency preparedness related digital assets. Critical digital assets are estimated over 60% among all digital assets in a nuclear power plant. Therefore, it was required to prioritize critical digital assets to improve efficiency of regulation and implementation. In this paper, the research status on methodology development to prioritize critical digital assets based on nuclear risk assessment will be introduced. In this paper, to derive digital asset directly affect accident, PRA results (ET, FT, and minimal cut set) are analyzed. According to result of analysis, digital systems related to CD are derived ESF-CCS (safety-related component control system) and Process-CCS (non-safety-related component control system) as well as Engineered Safety Features Actuation System (ESFAS). These digital assets can be identified Vital Digital Asset (VDA). Hereafter, to develop general methodology which was identified VDA related to accident among CDAs, (1) method using result of minimal cut set in PRA model will be studied and (2) method quantifying result of Digital I and C PRA which is performed to reflect all digital cabinet related to system in FT will be studied

  17. Review of Nuclear Criticality Safety Requirements Implementation for Hanford Tank Farms Facility

    International Nuclear Information System (INIS)

    DEFIGH PRICE, C.

    2000-01-01

    In November 1999, the Deputy Secretary of the Department of Energy directed a series of actions to strengthen the Department's ongoing nuclear criticality safety programs. A Review Plan describing lines of inquiry for assessing contractor programs was included. The Office of River Protection completed their assessment of the Tank Farm Contractor program in May 2000. This document supports that assessment by providing a compliance statement for each line of inquiry

  18. Critical experiments AT Pacific Northwest Laboratory

    International Nuclear Information System (INIS)

    Clayton, E.D.; Bierman, S.R.

    1984-01-01

    After a short description of the facility, a brief listing of the principal types of fuel forms and assembly geometries is provided. A number of experiments have recently been performed on plain fissionable units, or isolated assemblies of single units, that include measurements on solutions composed of Pu-U mixtures and critical experiment data on lattices of low enriched uranium in water. Experiments have been performed on planar arrays of containers with Pu solutions because of the lack of data in this field concerning the safe storage of nuclear fuel; others have been conducted on arrays of low enriched U lattice assemblies. Neutronic measurements to date have shown they can be used to provide additional benchmark data for improvement and validation of criticality codes. Studies have previously been made to ascertain the need for critical experiments in support of fuel recycle operations. The result of an effort to update the list of needed critical experiments is summarized in this section. Experiments are listed in support of uranium based fuels and fast breeder reactor fuels. An effort is made to identify those areas within the fuel cycle wherein the critical experiment data would be applied and to identify the experiments (and data) required to fulfill the needs in each of these areas. The type and form of fuel on which the data would be obtained also are identified. In presenting this information, no attempt is made to describe the experiments in detail, or to define the actual number of critical experiments that might be needed to provide the required data

  19. Proceedings of KURRI symposium on criticality safety

    International Nuclear Information System (INIS)

    Nishina, Kojiro; Kanda, Keiji

    1984-01-01

    On August 8, 1984, at the Reactor Application Center of the Research Reactor Institute, Kyoto University, the symposium on criticality safety was held, and 81 participants from various fields of reactor physics, nuclear fuel cycle engineering, reactor chemistry, nuclear chemistry, health physics and so on discussed the problem. The gists of the presentation are collected in this report. The contents are the techniques of evaluating criticality safety in respective fuel facilities, the system of control and its concept, the course and plan of the research on criticality safety in Japan and foreign countries, the techniques of determining multiplication factor and so on, and the review of present status, the pointing-out of problems and the report of new techniques were made. The measures coping with criticality safety have been mostly to meet urgent demand, but its fundamental examination and long term research should be carried out. This symposium was planned as the preparation for such research project, and favorable comment was given by the participants. In the next symposium, it is considered better to limit the themes and to allot more time to respective lectures. (Kako, I.)

  20. Measurements in Los Alamos benchmark criticals and the central reactivity discrepancy

    International Nuclear Information System (INIS)

    Davey, W.G.; Hansen, G.E.; Koelling, J.J.; McLaughlin, T.P.

    1978-01-01

    Measurements in seven Los Alamos fast critical facilities are described; all are related to elucidating the causes of the central reactivity discrepancy in fast reactors. Specific capabilities of these specialized assemblies permit measurements well-above delayed critical and these confirm the validity of the delayed neutron data used for calibration; there is therefore no reactivity-scale error. Reactivity measurements in these homogeneous assemblies exhibit no discrepancy. It is concluded that nuclear data should not be adjusted to eliminate the discrepancy found in other, heterogeneous assemblies

  1. Helium generation reaction rates for 6Li and 10B in benchmark facilities

    International Nuclear Information System (INIS)

    Farrar, Harry IV; Oliver, B.M.; Lippincott, E.P.

    1980-01-01

    The helium generation rates for 10 B and 6 Li have been measured in two benchmark reactor facilities having neutron spectra similar to those found in a breeder reactor. The irradiations took place in the Coupled Fast Reactivity Measurements Facility (CFRMF) and in the 10% enriched 235 U critical assembly, BIG-10. The helium reaction rates were obtained by precise high-sensitivity gas mass spectrometric analyses of the helium content of numerous small samples. Comparison of these reaction rates with other reaction rates measured in the same facilities, and with rates calculated from published cross sections and from best estimates of the neutron spectral shapes, indicate significant discrepancies in the calculated values. Additional irradiations in other benchmark facilities have been undertaken to better determine the energy ranges where the discrepancies lie

  2. KEOPS and other VENUS experiments dedicated to the criticality safety of a MOX fuel fabrication facility

    International Nuclear Information System (INIS)

    Lance, Benoit; Van Den Hende, Paul; Marloye, Daniel; Basselier, Jacques; Libon, Henri; De Vleeschhauwer, Marc; Moerenhout, Jeremie; Baeten, Peter

    2005-01-01

    The qualification scheme of criticality computer codes for Pu bearing powders lies upon databases which suffer from a lack of recent experimental results. As a MOX manufacture, BELGONUCLEAIRE is especially concerned by criticality safety and would like to address such an issue by launching with SCK-CEN an International Programme called KEOPS. (author)

  3. Results from 2010 Caliban Criticality Dosimetry Intercomparison

    Energy Technology Data Exchange (ETDEWEB)

    Veinot, K. G.

    2011-10-12

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

  4. Information needs critical to implementing the Federal Facility Compliance Act

    Energy Technology Data Exchange (ETDEWEB)

    Rasch, D.N. [Department of Energy-Idaho Operations Office, Idaho Falls, ID (United States); Kristofferson, K. [WINCO/INEL, Idaho Falls, ID (United States); Eaton, D.L. [EG& G Idaho/INEL, Idaho Falls, ID (United States)] [and others

    1994-12-31

    The presented paper summarizes the current status of data collection completed to support the Federal Facility Compliance Act (FFCA) Interim Mixed Waste Inventory Report (IMWIR), current needs, and related lessons learned. The Department of Energy (DOE), as required in Section 3021 of Resource Conservation and Recovery Act (RCRA), is required to prepare waste inventory reports, treatment reports and treatment plans. With this extensive effort, formulation of these requirements has required extensive data collection, validation and revision efforts. The framework for supporting these data needs has been enhanced by establishing a core database capable of supporting the required IMWIR, and has provided the basis for development of the Conceptual Site Treatment Plan (CSTP). The development of the CSTP has shown a need for complex wide standardized information that will ultimately become the basis for major land disposal restriction (LDR) activities such as; site treatment, equity resolution, consent agreement and continued capability to respond to stakeholder requests. DOE is in a position to dramatically demonstrate to the public and the states that mixed waste treatment can be cost effectively realized. To accomplish this program successfully will require use of existing data and expertise. This effort will be enhanced by implementation of basic system management processes which focus on completion of a mutually agreed to goal.

  5. Test facilities for radioactive material transport packages (AEA Technology, Winfrith, UK)

    International Nuclear Information System (INIS)

    Burgess, M.H.

    1991-01-01

    Transport packages for radioactive materials are tested to demonstrate compliance with national and international regulations. The involvement of AEA Technology is traced from the establishment of the early IAEA Regulations. Transport package design, testing, assessment and approval requires a wide variety of skills and facilities. The comprehensive capability of AEA Technology in these areas is described with references to practical experience in the form of a short bibliography. The facilities described include drop-test cranes and targets (up to 700te); air guns for impacts up to sonic velocities; pool fires, furnaces and rigs for thermal tests including heat dissipation on prototype flasks; shielding facilities and instruments; criticality simulations and leak test instruments. These are illustrated with photographs demonstrating the comprehensive nature of package testing services supplied to customers. (author)

  6. Proceedings of the Advanced Hadron Facility accelerator design workshop, February 20--25, 1989

    International Nuclear Information System (INIS)

    Thiessen, H.A.

    1990-04-01

    The International Workshop on Hadron Facility Technology was held February 20--25, 1989, at the Study Center at Los Alamos National Laboratory. This volume (first of two) included papers on architecture, beam diagnostics, compressors, and linacs. Participants included groups from AHF, Brookhaven National Laboratory, European Hadron Facility, Fermilab, and the Moscow Meson Factory. The workshop was well attended by members of the Los Alamos staff. The interchange of information and the opportunity by criticism by peers was important to all who attended

  7. Risk Assessment Methodology for Protecting Our Critical Physical Infrastructures

    Energy Technology Data Exchange (ETDEWEB)

    BIRINGER,BETTY E.; DANNEELS,JEFFREY J.

    2000-12-13

    Critical infrastructures are central to our national defense and our economic well-being, but many are taken for granted. Presidential Decision Directive (PDD) 63 highlights the importance of eight of our critical infrastructures and outlines a plan for action. Greatly enhanced physical security systems will be required to protect these national assets from new and emerging threats. Sandia National Laboratories has been the lead laboratory for the Department of Energy (DOE) in developing and deploying physical security systems for the past twenty-five years. Many of the tools, processes, and systems employed in the protection of high consequence facilities can be adapted to the civilian infrastructure.

  8. Life Management and Safety of Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Fabbri, S.; Diluch, A.; Vega, G., E-mail: fabbri@cnea.gov.ar [Comisión Nacional de Energía Atómica, Buenos Aires (Argentina)

    2014-10-15

    The nuclear programme in Argentina includes: nuclear power and related supplies, medical and industrial applications, waste management, research and development and human training. Nuclear facilities require life management programs that allow a safe operation. Safety is the first priority for designers and operators. This can be attained with defence in depth: regular inspections and maintenance procedures to minimize failure risks. CNEA objectives in this area are to possess the necessary capability to give safe and fast technical support. Within this scheme, one of the main activities undertaken by CNEA is to provide technological assistance to the nuclear plants and research reactors. As a consequence of an increasing concern about safety and ageing a Life Management Department for safe operation was created to take care of these subjects. The goal is to elaborate a Safety Evaluation Process for the critical components of nuclear plants and other facilities. The overall objectives of a safety process are to ensure a continuous safe, reliable and effective operation of nuclear facilities and it means the implementation of the defence in deep concept to enhance safety for the protection of the public, the workers and the environment. (author)

  9. Variations in Financial Performance among Peer Groups of Critical Access Hospitals

    Science.gov (United States)

    Pink, George H.; Holmes, George M.; Thompson, Roger E.; Slifkin, Rebecca T.

    2007-01-01

    Context: Among the large number of hospitals with critical access hospital (CAH) designation, there is substantial variation in facility revenue as well as the number and types of services provided. If these variations have material effects on financial indicators, then performance comparisons among all CAHs are problematic. Purpose: To…

  10. Critical heat flux (CHF) phenomenon on a downward facing curved surface

    Energy Technology Data Exchange (ETDEWEB)

    Cheung, F.B.; Haddad, K.H.; Liu, Y.C. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical Engineering

    1997-06-01

    This report describes a theoretical and experimental study of the boundary layer boiling and critical heat flux phenomena on a downward facing curved heating surface, including both hemispherical and toroidal surfaces. A subscale boundary layer boiling (SBLB) test facility was developed to measure the spatial variation of the critical heat flux and observe the underlying mechanisms. Transient quenching and steady-state boiling experiments were performed in the SBLB facility under both saturated and subcooled conditions to obtain a complete database on the critical heat flux. To complement the experimental effort, an advanced hydrodynamic CHF model was developed from the conservation laws along with sound physical arguments. The model provides a clear physical explanation for the spatial variation of the CHF observed in the SBLB experiments and for the weak dependence of the CHF data on the physical size of the vessel. Based upon the CHF model, a scaling law was established for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water. The scaling law, which compares favorably with all the available local CHF data obtained for various vessel sizes, can be used to predict the local CHF limits on large commercial-size vessels. This technical information represents one of the essential elements that is needed in assessing the efficacy of external cooling of core melt by cavity flooding as a severe accident management strategy. 83 figs., 3 tabs.

  11. Critical heat flux (CHF) phenomenon on a downward facing curved surface

    International Nuclear Information System (INIS)

    Cheung, F.B.; Haddad, K.H.; Liu, Y.C.

    1997-06-01

    This report describes a theoretical and experimental study of the boundary layer boiling and critical heat flux phenomena on a downward facing curved heating surface, including both hemispherical and toroidal surfaces. A subscale boundary layer boiling (SBLB) test facility was developed to measure the spatial variation of the critical heat flux and observe the underlying mechanisms. Transient quenching and steady-state boiling experiments were performed in the SBLB facility under both saturated and subcooled conditions to obtain a complete database on the critical heat flux. To complement the experimental effort, an advanced hydrodynamic CHF model was developed from the conservation laws along with sound physical arguments. The model provides a clear physical explanation for the spatial variation of the CHF observed in the SBLB experiments and for the weak dependence of the CHF data on the physical size of the vessel. Based upon the CHF model, a scaling law was established for estimating the local critical heat flux on the outer surface of a heated hemispherical vessel that is fully submerged in water. The scaling law, which compares favorably with all the available local CHF data obtained for various vessel sizes, can be used to predict the local CHF limits on large commercial-size vessels. This technical information represents one of the essential elements that is needed in assessing the efficacy of external cooling of core melt by cavity flooding as a severe accident management strategy. 83 figs., 3 tabs

  12. An evaluation of the control rod modelling approach used in VSOP by comparing its results to the experiments performed in the ASTRA critical facility

    International Nuclear Information System (INIS)

    Reitsma, F.; Naidoo, D.; Karriem, Z.

    2002-01-01

    The modelling of strong absorber regions in diffusion theory is a well-known problem and many methods have been developed to accommodate the transport effects in diffusion theory. In this work the method of equivalent cross sections is evaluated for the ASTRA critical facility at the Russian Research Centre - Kurchatov Institute in Moscow. The measured reactivity worths of the control rods situated in the side reflector, are compared with the calculated values making use of equivalent diffusion parameters in VSOP. Favourable results were obtained for the control rods positioned within the first ring of reflector blocks with larger errors obtained for control rods positioned further from the core. Furthermore, the use of an equivalent boron concentration to represent the absorber regions was also investigated and shown to be useful if applied correctly and with care. However, the practical difficulties and restrictions imposed by the two approaches make the investigation of an alternative method, which should remove these shortcomings, attractive. (author)

  13. Overview of seismic probabilistic risk assessment for structural analysis in nuclear facilities

    International Nuclear Information System (INIS)

    Reed, J.W.

    1989-01-01

    Probabilistic Risk Assessment (PRA) for seismic events is currently being performed for nuclear and DOE facilities. The background on seismic PRA is presented along with a basic description of the method. The seismic PRA technique is applicable to other critical facilities besides nuclear plants. The different approaches for obtained structure fragility curves are discussed and their applications to structures and equipment, in general, are addressed. It is concluded that seismic PRA is a useful technique for conducting probability analysis for a wide range of classes of structures and equipment

  14. Sustainability of utility-scale solar energy – critical ecological concepts

    Science.gov (United States)

    Moore-O'Leary, Kara A.; Hernandez, Rebecca R.; Johnston, Dave S.; Abella, Scott R.; Tanner, Karen E.; Swanson, Amanda C.; Kreitler, Jason R.; Lovich, Jeffrey E.

    2017-01-01

    Renewable energy development is an arena where ecological, political, and socioeconomic values collide. Advances in renewable energy will incur steep environmental costs to landscapes in which facilities are constructed and operated. Scientists – including those from academia, industry, and government agencies – have only recently begun to quantify trade-offs in this arena, often using ground-mounted, utility-scale solar energy facilities (USSE, ≥1 megawatt) as a model. Here, we discuss five critical ecological concepts applicable to the development of more sustainable USSE with benefits over fossil-fuel-generated energy: (1) more sustainable USSE development requires careful evaluation of trade-offs between land, energy, and ecology; (2) species responses to habitat modification by USSE vary; (3) cumulative and large-scale ecological impacts are complex and challenging to mitigate; (4) USSE development affects different types of ecosystems and requires customized design and management strategies; and (5) long-term ecological consequences associated with USSE sites must be carefully considered. These critical concepts provide a framework for reducing adverse environmental impacts, informing policy to establish and address conservation priorities, and improving energy production sustainability.

  15. Neutron spectral characterization of the PCA-PV benchmark facility

    International Nuclear Information System (INIS)

    Stallmann, F.W.; Kam, F.B.K.; Fabry, A.

    1980-01-01

    The Pool Critical Assembly (PCA) at the Oak Ridge National Laboratory is being used to generate the PCA-PV benchmark neutron field. A configuration consisting of steel blocks and water gaps is used to simulate the thermal shield pressure vessel configurations in power reactors. The distances between the steel blocks can be changed so that the penetration of neutrons through water and steel can be determined and compared for many different configurations. Easy access and low flux levels make it possible to conduct extensive measurements using active and passive neutron dosimetry, which are impossible to perform in commercial reactors. The clean core and simple geometry facilitates neutron transport calculations which can be validated in detail by comparison with measurements. A facility which has the same configuration of water and steel as the PCA-PV facility but contains test specimens for materials testing, will be irradiated in the higher fluxes at the Oak Ridge Research Reactor. Using the results from the PCA-PV facility, the correlation between neutron flux-fluences and radiation damage in steel can be established. This facility is being discussed in a separate paper

  16. American National Standards and the DOE - A cooperative effort to promote nuclear criticality safety

    International Nuclear Information System (INIS)

    Rothleder, B.M.

    1996-01-01

    The U.S. Department of Energy's (DOE's) new criticality safety order, DOE Order 420.1 (open-quotes Facility Safety,close quotes October 13, 1995), Sec. 4.3 (open-quotes Nuclear Criticality Safetyclose quotes), invokes, as an integral part, 12 appropriate American National Standards Institute/American Nuclear Society (ANSI/ANS) Series-8 standards for nuclear criticality safety, but with modifications. (The order that 420.1/4.3 replaced also invoked some ANSI/ANS Series-8 standards.) These modifications include DOE operation-specific exceptions to the standards and elaborations on some of the wording in the standards

  17. Flow analysis of HANARO flow simulated test facility

    International Nuclear Information System (INIS)

    Park, Yong-Chul; Cho, Yeong-Garp; Wu, Jong-Sub; Jun, Byung-Jin

    2002-01-01

    The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial critical in February, 1995. Many experiments should be safely performed to activate the utilization of the NANARO. A flow simulated test facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The half-core structure assembly is composed of plenum, grid plate, core channel with flow tubes, chimney and dummy pool. The flow channels are to be filled with flow orifices to simulate core channels. This test facility must simulate similar flow characteristics to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the test facility. The computational flow analysis has been performed for the verification of flow structure and similarity of this test facility assuming that flow rates and pressure differences of the core channel are constant. The shapes of flow orifices were determined by the trial and error method based on the design requirements of core channel. The computer analysis program with standard k - ε turbulence model was applied to three-dimensional analysis. The results of flow simulation showed a similar flow characteristic with that of the HANARO and satisfied the design requirements of this test facility. The shape of flow orifices used in this numerical simulation can be adapted for manufacturing requirements. The flow rate and the pressure difference through core channel proved by this simulation can be used as the design requirements of the flow system. The analysis results will be verified with the results of the flow test after construction of the flow system. (author)

  18. The CMS CERN Analysis Facility (CAF)

    Energy Technology Data Exchange (ETDEWEB)

    Buchmueller, O [Imperial College (United Kingdom); Bonacorsi, D [Universita and INFN, Bologna (Italy); Fanzago, F [Universita and INFN, Padova (Italy); Gowdy, S; Malgeri, L; Panzer-Steindel, B; Schwickerath, U; Spiga, D; Toebbicke, Rainer [Conseil Europeen Recherche Nucl. (CERN) Switzerland (Switzerland); Kreuzer, P [Rheinisch-Westfaelische Tech. Hoch. (RWTH) (Germany); Mankel, R [Deutsches Elektronen-Synchrotron (DESY) (Germany); Metson, S [University of Bristol (United Kingdom); Sanches, J Afonso; Teodoro, D, E-mail: Peter.Kreuzer@cern.c [Universidade do Estado do Rio De Janeiro (UERJ) (Brazil)

    2010-04-01

    The CMS CERN Analysis Facility (CAF) was primarily designed to host a large variety of latency-critical workflows. These break down into alignment and calibration, detector commissioning and diagnosis, and high-interest physics analysis requiring fast-turnaround. In addition to the low latency requirement on the batch farm, another mandatory condition is the efficient access to the RAW detector data stored at the CERN Tier-0 facility. The CMS CAF also foresees resources for interactive login by a large number of CMS collaborators located at CERN, as an entry point for their day-by-day analysis. These resources will run on a separate partition in order to protect the high-priority use-cases described above. While the CMS CAF represents only a modest fraction of the overall CMS resources on the WLCG GRID, an appropriately sized user-support service needs to be provided. We will describe the building, commissioning and operation of the CMS CAF during the year 2008. The facility was heavily and routinely used by almost 250 users during multiple commissioning and data challenge periods. It reached a CPU capacity of 1.4MSI2K and a disk capacity at the Peta byte scale. In particular, we will focus on the performances in terms of networking, disk access and job efficiency and extrapolate prospects for the upcoming LHC first year data taking. We will also present the experience gained and the limitations observed in operating such a large facility, in which well controlled workflows are combined with more chaotic type analysis by a large number of physicists.

  19. The CMS CERN Analysis Facility (CAF)

    International Nuclear Information System (INIS)

    Buchmueller, O; Bonacorsi, D; Fanzago, F; Gowdy, S; Malgeri, L; Panzer-Steindel, B; Schwickerath, U; Spiga, D; Toebbicke, Rainer; Kreuzer, P; Mankel, R; Metson, S; Sanches, J Afonso; Teodoro, D

    2010-01-01

    The CMS CERN Analysis Facility (CAF) was primarily designed to host a large variety of latency-critical workflows. These break down into alignment and calibration, detector commissioning and diagnosis, and high-interest physics analysis requiring fast-turnaround. In addition to the low latency requirement on the batch farm, another mandatory condition is the efficient access to the RAW detector data stored at the CERN Tier-0 facility. The CMS CAF also foresees resources for interactive login by a large number of CMS collaborators located at CERN, as an entry point for their day-by-day analysis. These resources will run on a separate partition in order to protect the high-priority use-cases described above. While the CMS CAF represents only a modest fraction of the overall CMS resources on the WLCG GRID, an appropriately sized user-support service needs to be provided. We will describe the building, commissioning and operation of the CMS CAF during the year 2008. The facility was heavily and routinely used by almost 250 users during multiple commissioning and data challenge periods. It reached a CPU capacity of 1.4MSI2K and a disk capacity at the Peta byte scale. In particular, we will focus on the performances in terms of networking, disk access and job efficiency and extrapolate prospects for the upcoming LHC first year data taking. We will also present the experience gained and the limitations observed in operating such a large facility, in which well controlled workflows are combined with more chaotic type analysis by a large number of physicists.

  20. New design procedure development of future reactor critical power estimation. (1) Practical design-by-analysis method for BWR critical power design correlation

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Mitsutake, Toru

    2007-01-01

    For present BWR fuels, the full mock-up thermal-hydraulic test, such as the critical power measurement test, pressure drop measurement test and so on, has been needed. However, the full mock-up test required the high costs and large-scale test facility. At present, there are only a few test facilities to perform the full mock-up thermal-hydraulic test in the world. Moreover, for future BWR, the bundle size tends to be larger, because of reducing the plant construction costs and minimizing the routine check period. For instance, AB1600, improved ABWR, was proposed from Toshiba, whose bundle size was 1.2 times larger than the conventional BWR fuel size. It is too expensive and far from realistic to perform the full mock-up thermal-hydraulic test for such a large size fuel bundle. The new design procedure is required to realize the large scale bundle design development, especially for the future reactor. Therefore, the new design procedure, Practical Design-by-Analysis (PDBA) method, has been developed. This new procedure consists of the partial mock-up test and numerical analysis. At present, the subchannel analysis method based on three-fluid two-phase flow model only is a realistic choice. Firstly, the partial mock-up test is performed, for instance, the 1/4 partial mock-up bundle. Then, the first-step critical power correlation coefficients are evaluated with the measured data. The input data, such as the spacer effect model coefficient, on the subchannel analysis are also estimated with the data. Next, the radial power effect on the critical power of the full-bundle size was estimated with the subchannel analysis. Finally, the critical power correlation is modified by the subchannel analysis results. In the present study, the critical power correlation of the conventional 8x8 BWR fuel was developed with the PDBA method by 4x4 partial mock-up tests and the subchannel analysis code. The accuracy of the estimated critical power was 3.8%. The several themes remain to

  1. Study and comparison of different methods control in light water critical facility

    International Nuclear Information System (INIS)

    Michaiel, M.L.; Mahmoud, M.S.

    1980-01-01

    The control of nuclear reactors, may be studied using several control methods, such as control by rod absorbers, by inserting or removing fuel rods (moderator cavities), or by changing reflector thickness. Every method has its advantage, the comparison between these different methods and their effect on the reactivity of a reactor is the purpose of this work. A computer program is written by the authors to calculate the critical radius and worth in any case of the three precedent methods of control

  2. The Maternity Benefit (Amendment Bill, 2016: A Critical Analysis

    Directory of Open Access Journals (Sweden)

    Suman Singh

    2016-11-01

    Full Text Available On 11 August 2016, amending the Maternity Benefit Act, 1961, the new bill, The Maternity Benefit (Amendment Bill, 2016 was introduced and passed in the Rajya Sabha (or Council of States, the upper house of the Parliament of India. Central aim of this article is to critically review the amendments to the bill regarding geographies of maternity leave and its associated facilities.

  3. 77 FR 56231 - National Register of Historic Places; Notification of Pending Nominations and Related Actions

    Science.gov (United States)

    2012-09-12

    ... National Register criteria for evaluation. Comments may be forwarded by United States Postal Service, to... Hook, 12000831 Rensselaer County Cornell--Manchester Farmstead, (Farmsteads of Pittstown, New York MPS...

  4. Radiological impacts from nuclear facilities on non-human species

    International Nuclear Information System (INIS)

    1997-07-01

    This monograph is the Proceedings of a Symposium on Radiological Impacts from Nuclear Facilities on Non-Human Species, held in Ottawa, Canada, December 1 and 2, 1996. The Symposium was held in response to the assessment of radiological impacts from nuclear facilities on non-human biota by Environment Canada and the move by Atomic Energy Control Board to include the radiological impacts in its regulatory regime. The two major goals of the Symposium were to critically evaluate the ecological risk assessment as applied to radionuclides and contribute to the wide consultation sought by the Atomic Energy Control Board on their new environmental initiatives. The series of papers presented at the Symposium discuss issues relevant to the two major objectives of the Symposium

  5. A study on critical heat flux in gap

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Jeong, Ji Whan; Cho, Young Ro; Chang, Young Cho; Kang, Kyung Ho; Kim, Jong Whan; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    The scope and content of this study is to perform the test on critical heat flux in hemispherical narrow gaps using distilled water and Freon R-113 as experimental parameters, such as system pressure from 1 to 10 atm and gap thickness of 0.5, 1.0, 2.0, and 5.0 mm. The CHFG test results have shown that the measured values of critical power are much lower than the predictions made by empirical CHF correlations applicable to flat plate gaps and annuli. The pressure effect on the critical power was found to be much milder than predictions by those CHF correlations. The values and the pressure trend of the critical powers measured in the present experiments are close to the values converted from the CCFL data. This confirms the claim that a CCFL brings about local dryout and finally, global dryout in hemispherical narrow gaps. Increases in the gap thickness lead to increase in critical power. The measured critical power using R-113 in hemispherical narrow gaps are 60 % lower than that using water due to the lower boiling point, which is different from the pool boiling condition. The CCFL (counter counter flow limit) test facility was constructed and the test is being performed to estimate the CCFL phenomena and to evaluate the CHFG test results on critical power in hemispherical narrow gaps. (Author). 35 refs., 2 tabs., 19 figs.

  6. A study on critical heat flux in gap

    International Nuclear Information System (INIS)

    Park, Rae Joon; Jeong, Ji Whan; Cho, Young Ro; Chang, Young Cho; Kang, Kyung Ho; Kim, Jong Whan; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    The scope and content of this study is to perform the test on critical heat flux in hemispherical narrow gaps using distilled water and Freon R-113 as experimental parameters, such as system pressure from 1 to 10 atm and gap thickness of 0.5, 1.0, 2.0, and 5.0 mm. The CHFG test results have shown that the measured values of critical power are much lower than the predictions made by empirical CHF correlations applicable to flat plate gaps and annuli. The pressure effect on the critical power was found to be much milder than predictions by those CHF correlations. The values and the pressure trend of the critical powers measured in the present experiments are close to the values converted from the CCFL data. This confirms the claim that a CCFL brings about local dryout and finally, global dryout in hemispherical narrow gaps. Increases in the gap thickness lead to increase in critical power. The measured critical power using R-113 in hemispherical narrow gaps are 60 % lower than that using water due to the lower boiling point, which is different from the pool boiling condition. The CCFL (counter counter flow limit) test facility was constructed and the test is being performed to estimate the CCFL phenomena and to evaluate the CHFG test results on critical power in hemispherical narrow gaps. (Author). 35 refs., 2 tabs., 19 figs

  7. The Criticality Safety Information Resource Center (CSIRC) at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Henderson, B.D.; Meade, R.A.; Pruvost, N.L.

    1999-01-01

    The Criticality Safety Information Resource Center (CSIRC) at Los Alamos National Laboratory (LANL) is a program jointly funded by the U.S. Department of Energy (DOE) and the U.S. Nuclear Regulatory Commission (NRC) in conjunction with the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 97-2. The goal of CSIRC is to preserve primary criticality safety documentation from U.S. critical experimental sites and to make this information available for the benefit of the technical community. Progress in archiving criticality safety primary documents at the LANL archives as well as efforts to make this information available to researchers are discussed. The CSIRC project has a natural linkage to the International Criticality Safety Benchmark Evaluation Project (ICSBEP). This paper raises the possibility that the CSIRC project will evolve in a fashion similar to the ICSBEP. Exploring the implications of linking the CSIRC to the international criticality safety community is the motivation for this paper

  8. Critical supply network protection against intentional attacks: A game-theoretical model

    International Nuclear Information System (INIS)

    Bricha, Naji; Nourelfath, Mustapha

    2013-01-01

    A crucial issue in today's critical supply chains is how to protect facilities against intentional attacks, since it has become unacceptable to ignore the high impact of low probability disruptions caused by these attacks. This article develops a game-theoretical model to deal with the protection of facilities, in the context of the uncapacitated fixed-charge location problem. Given a set of investment alternatives for protecting the facilities against identified threats, the objective is to select the optimal defence strategy. The attacker is considered as a player who tries to maximise the expected damage while weighing against the attacks expenditures. The conflict on facilities vulnerability is modelled using the concept of contest. The vulnerability of a facility is defined by its destruction probability. Contest success functions determine the vulnerability of each facility dependent on the relative investments of the defender and the attacker on each facility, and on the characteristics of the contest. A method is developed to evaluate the utilities of the players (i.e., the defender and the attacker). This method evaluates many expected costs, including the cost needed to restore disabled facilities, the backorder cost, and the cost incurred because of the increase in transportation costs after attacks. In fact, when one or several facilities are unavailable, transportation costs will increase since reassigned customers may receive shipments from facilities which are farther away. The model considers a non-cooperative two-period game between the players, and an algorithm is presented to determine the equilibrium solution and the optimal defence strategy. An illustrative example is presented. The approach is compared to other suggested strategies, and some managerial insights are provided in the context of facility location

  9. Summary of ORSphere critical and reactor physics measurements

    Directory of Open Access Journals (Sweden)

    Marshall Margaret A.

    2017-01-01

    Full Text Available In the early 1970s Dr. John T. Mihalczo (team leader, J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF with highly enriched uranium (HEU metal (called Oak Ridge Alloy or ORALLOY to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP. Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

  10. Summary of ORSphere critical and reactor physics measurements

    Science.gov (United States)

    Marshall, Margaret A.; Bess, John D.

    2017-09-01

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

  11. Facility effluent monitoring plan determinations for the 400 Area facilities

    International Nuclear Information System (INIS)

    Nickels, J.M.

    1991-09-01

    This Facility Effluent Monitoring Plan determination resulted from an evaluation conducted for the Westinghouse Hanford Company 400 Area facilities on the Hanford Site. The Facility Effluent Monitoring Plan determinations have been prepared in accordance with A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans. Two major Westinghouse Hanford Company facilities in the 400 Area were evaluated: the Fast Flux Test Facility and the Fuels Manufacturing and examination Facility. The determinations were prepared by Westinghouse Hanford Company. Of these two facilities, only the Fast Flux Test Facility will require a Facility Effluent Monitoring Plan. 7 refs., 5 figs., 4 tabs

  12. Study on system integration of robots operated in nuclear fusion facility and nuclear power plant facilities

    International Nuclear Information System (INIS)

    Oka, Kiyoshi

    2004-07-01

    A present robot is required to apply to many fields such as amusement, welfare and protection against disasters. The are however only limited numbers of the robots, which can work under the actual conditions as a robot system. It is caused by the following reasons: (1) the robot system cannot be realized by the only collection of the elemental technologies, (2) the performance of the robot is determined by that of the integrated system composed of the complicated elements with many functions, and (3) the respective elements have to be optimized in the integrated robot system with a well balance among them, through their examination, adjustment and improvement. Therefore, the system integration of the robot composed of a large number of elements is the most critical issue to realize the robot system for actual use. In the present paper, I describe the necessary approaches and elemental technologies to solve the issues on the system integration of the typical robot systems for maintenance in the nuclear fusion facility and rescue in the accident of the nuclear power plant facilities. These robots work under the intense radiation condition and restricted space in place of human. In particular, I propose a new approach to realize the system integration of the robot for actual use from the viewpoints of not only the environment and working conditions but also the restructure and optimization of the required elemental technologies with a well balance in the robot system. Based on the above approach, I have a contribution to realize the robot systems working under the actual conditions for maintenance in the nuclear fusion facility and rescue in the accident of the nuclear power plant facilities. (author)

  13. CSER-98-002: Criticality analysis for the storage of special nuclear material sources and standards in the WRAP facility

    International Nuclear Information System (INIS)

    GOLDBERG, H.J.

    1999-01-01

    The Waste Receiving and Processing (WRAP) Facility will store uranium and transuranic (TRU) sources and standards for certification that WRAP meets the requirements of the Quality Assurance Program Plan (QAPP) for the Waste Isolation Pilot Plant (WIPP). In addition, WRAP must meet internal requirements for testing and validation of measuring instruments for nondestructive assay (NDA). In order to be certified for WIPP, WRAP will participate in the NDA Performance Demonstration Program (PDP). This program is a blind test of the NDA capabilities for TRU waste. It is intended to ensure that the NDA capabilities of this facility satisfy the requirements of the quality assurance program plan for the WIPP. The PDP standards have been provided by the Los Alamos National Laboratory (LANL) for this program. These standards will be used in the WRAP facility

  14. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    International Nuclear Information System (INIS)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs

  15. Operation of the nuclear fuel cycle test facilities -Operation of the hot test loop facilities

    Energy Technology Data Exchange (ETDEWEB)

    Chun, S. Y.; Jeong, M. K.; Park, C. K.; Yang, S. K.; Won, S. Y.; Song, C. H.; Jeon, H. K.; Jeong, H. J.; Cho, S.; Min, K. H.; Jeong, J. H.

    1997-01-01

    A performance and reliability of a advanced nuclear fuel and reactor newly designed should be verified by performing the thermal hydraulics tests. In thermal hydraulics research team, the thermal hydraulics tests associated with the development of an advanced nuclear fuel and reactor haven been carried out with the test facilities, such as the Hot Test Loop operated under high temperature and pressure conditions, Cold Test Loop, RCS Loop and B and C Loop. The objective of this project is to obtain the available experimental data and to develop the advanced measuring techniques through taking full advantage of the facilities. The facilities operated by the thermal hydraulics research team have been maintained and repaired in order to carry out the thermal hydraulics tests necessary for providing the available data. The performance tests for the double grid type bottom end piece which was improved on the debris filtering effectivity were performed using the PWR-Hot Test Loop. The CANDU-Hot Test Loop was operated to carry out the pressure drop tests and strength tests of CANFLEX fuel. The Cold Test Loop was used to obtain the local velocity data in subchannel within HANARO fuel bundle and to study a thermal mixing characteristic of PWR fuel bundle. RCS thermal hydraulic loop was constructed and the experiments have been carried out to measure the critical heat flux. In B and C Loop, the performance tests for each component were carried out. (author). 19 tabs., 78 figs., 19 refs.

  16. General principles of the nuclear criticality safety for handling, processing and transportation fissile materials in the USSR

    International Nuclear Information System (INIS)

    Vnukov, V.S.; Rjazanov, B.G.; Sviridov, V.I.; Frolov, V.V.; Zubkov, Y.N.

    1991-01-01

    The paper describes the general principles of nuclear criticality safety for handling, processing, transportation and fissile materials storing. Measures to limit the consequences of critical accidents are discussed for the fuel processing plants and fissile materials storage. The system of scientific and technical measures on nuclear criticality safety as well as the system of control and state supervision based on the rules, limits and requirements are described. The criticality safety aspects for various stages of handling nuclear materials are considered. The paper gives descriptions of the methods and approaches for critical risk assessments for the processing facilities, plants and storages. (Author)

  17. Australian proton therapy facilities - status report

    International Nuclear Information System (INIS)

    Bleasel, S.; Jackson, M.

    2000-01-01

    Full text: Radiotherapy plays an important role in the treatment of cancer; both in the curative treatment of localised disease and in alleviating symptoms in more advanced disease. Radiotherapy is usually given with megavoltage X-rays which give good penetration at depth and a lower dose on the skin. The aim is to give a high dose to the tumour while keeping the dose to normal tissues as low as possible. While X-rays continue beyond the target volume, protons and other charged particles have a finite range in tissue and this allows the high dose region to closely conform to the tumour, thereby giving the best chance of tumour control with minimum side effects. This is particularly important for small tumours adjacent to critical normal structures. The unmodified Bragg peak is too narrow to be useful but can be spread out to cover the tumour. Protons of energy 70 - 250 MeV are required to achieve the optimal depth in the body. Protons have been used for cancer treatment since 1954 but most of these treatments have been given in physics-based facilities which do not provide an ideal environment for the patient and may have only limited beam time available. A dedicated, hospital based facility was built at Loma Linda in California in 1991 and has now treated over 5,000 patients. Several other centres are being built in the USA, Japan and Europe and one is now being considered for Australia. Early 1998 Hitachi Australia Ltd. was asked to research the possibility of building a facility in Australia to serve Oceania. Two major hospitals showed interest, Royal Brisbane Hospital and Royal Prince Alfred Hospital. Both hospitals are undergoing redevelopment and have space for a building 3 stories high with a foot-print of approximately 50 m x 50 m. What is proposed for Australia is a principally clinical research facility with a dedicated area for physics research. A Steering Committee will be established to develop a document for presentation to Government. The facility

  18. Masters Thesis- Criticality Alarm System Design Guide with Accompanying Alarm System Development for the Radioisotope Production Laboratory in Richland, Washington

    Energy Technology Data Exchange (ETDEWEB)

    Greenfield, Bryce A. [Univ. of New Mexico, Albuquerque, NM (United States)

    2009-12-01

    A detailed instructional manual was created to guide criticality safety engineers through the process of designing a criticality alarm system (CAS) for Department of Energy (DOE) hazard class 1 and 2 facilities. Regulatory and technical requirements were both addressed. A list of design tasks and technical subtasks are thoroughly analyzed to provide concise direction for how to complete the analysis. An example of the application of the design methodology, the Criticality Alarm System developed for the Radioisotope Production Laboratory (RPL) of Richland, Washington is also included. The analysis for RPL utilizes the Monte Carlo code MCNP5 for establishing detector coverage in the facility. Significant improvements to the existing CAS were made that increase the reliability, transparency, and coverage of the system.

  19. Radiological safety assessment of a reference INTOR facility

    International Nuclear Information System (INIS)

    Khan, T.A.; Stasko, R.R.; Watts, R.T.; Shaw, G.; Morrison, C.A.; Russell, S.; Kempe, T.; Zimmerman, R.

    1985-03-01

    This report consists of a number of separate studies all of which were performed in support of INTOR Critical Issue D: Tritium Containment and Personnel Access vs Remote Maintenance. The common thread running through these studies is the radiological safety element in the design and operation of the INTOR facility. The intent is to help establish a firm basis for comparisons between a reactor cell maintenance option which requires personnel access, and one which involves completely remote maintenance

  20. Measurement of the Portsmouth Gaseous Diffusion Plant criticality accident alarm

    International Nuclear Information System (INIS)

    Tayloe, R.W. Jr.; D'Aquila, D.M.; McGinnis, R.B.

    1991-01-01

    The nuclear criticality accident radiation alarm system installed at the Portsmouth Gaseous Diffusion Plant was tested extensively at critical facilities located at the Los Alamos National Laboratory. The ability of the neutron scintillator radiation detection units to respond to a minimum accident of concern as defined in Standard ANSI/ANS-83.-1986 was demonstrated. Detector placement and the established trip point are based on shielding calculations performed by the Oak Ridge National Laboratory and criticality specialists at the Portsmouth plant. Based on these experiments and calculations, a detector trip point of 5 mrad/h in air is used. Any credible criticality accident is expected to produce neutron radiation fields >5 mrad/h in air at one or more radiation alarm locations. Each radiation alarm location has a cluster of three detectors that employs a two-out-of-three alarm logic. Earlier work focused on testing the alarm logic latching circuitry. This work was directed toward measurements involving the actual audible alarm signal delivered