High Flux Isotope Reactor system RELAP5 input model

International Nuclear Information System (INIS)

Morris, D.G.; Wendel, M.W.

1993-01-01

A thermal-hydraulic computational model of the High Flux Isotope Reactor (HFIR) has been developed using the RELAP5 program. The purpose of the model is to provide a state-of-the art thermal-hydraulic simulation tool for analyzing selected hypothetical accident scenarios for a revised HFIR Safety Analysis Report (SAR). The model includes (1) a detailed representation of the reactor core and other vessel components, (2) three heat exchanger/pump cells, (3) pressurizing pumps and letdown valves, and (4) secondary coolant system (with less detail than the primary system). Data from HFIR operation, component tests, tests in facility mockups and the HFIR, HFIR specific experiments, and other pertinent experiments performed independent of HFIR were used to construct the model and validate it to the extent permitted by the data. The detailed version of the model has been used to simulate loss-of-coolant accidents (LOCAs), while the abbreviated version has been developed for the operational transients that allow use of a less detailed nodalization. Analysis of station blackout with core long-term decay heat removal via natural convection has been performed using the core and vessel portions of the detailed model

Low-Enriched Uranium Fuel Design with Two-Dimensional Grading for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01

An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

Determination of the theoretical feasibility for the transmutation of europium isotopes from high flux isotope reactor control cylinders

International Nuclear Information System (INIS)

Elam, K.R.; Reich, W.J.

1995-09-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) is a 100 MWth light-water research reactor designed and built in the 1960s primarily for the production of transuranic isotopes. The HFIR is equipped with two concentric cylindrical blade assemblies, known as control cylinders, that are used to control reactor power. These control cylinders, which become highly radioactive from neutron exposure, are periodically replaced as part of the normal operation of the reactor. The highly radioactive region of the control cylinders is composed of europium oxide in an aluminum matrix. The spent HFIR control cylinders have historically been emplaced in the ORNL Waste Area Grouping (WAG) 6. The control cylinders pose a potential radiological hazard due to the long lived radiotoxic europium isotopes 152 Eu, 154 Eu, and 155 Eu. In a 1991 health evaluation of WAG 6 (ERD 1991) it was shown that these cylinders were a major component of the total radioactivity in WAG 6 and posed a potential exposure hazard to the public in some of the postulated assessment scenarios. These health evaluations, though preliminary and conservative in nature, illustrate the incentive to investigate methods for permanent destruction of the europium radionuclides. When the cost of removing the control cylinders from WAG 6, performing chemical separations and irradiating the material in HFIR are factored in, the option of leaving the control cylinders in place for decay must be considered. Other options, such as construction of an engineered barrier around the disposal silos to reduce the chance of migration, should also be analyzed

Advanced Fuel/Cladding Testing Capabilities in the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Ott, Larry J.; Ellis, Ronald James; McDuffee, Joel Lee; Spellman, Donald J.; Bevard, Bruce Balkcom

2009-01-01

The ability to test advanced fuels and cladding materials under reactor operating conditions in the United States is limited. The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) and the newly expanded post-irradiation examination (PIE) capability at the ORNL Irradiated Fuels Examination Laboratory provide unique support for this type of advanced fuel/cladding development effort. The wide breadth of ORNL's fuels and materials research divisions provides all the necessary fuel development capabilities in one location. At ORNL, facilities are available from test fuel fabrication, to irradiation in HFIR under either thermal or fast reactor conditions, to a complete suite of PIEs, and to final product disposal. There are very few locations in the world where this full range of capabilities exists. New testing capabilities at HFIR have been developed that allow testing of advanced nuclear fuels and cladding materials under prototypic operating conditions (i.e., for both fast-spectrum conditions and light-water-reactor conditions). This paper will describe the HFIR testing capabilities, the new advanced fuel/cladding testing facilities, and the initial cooperative irradiation experiment that begins this year.

Fabrication of control rods for the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Sease, J.D.

1998-01-01

The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A

Fabrication of control rods for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Sease, J.D.

1998-03-01

The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

Lessons Learned in the Update of a Safety Limit for the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Cook, David Howard

2009-01-01

A recent unreviewed safety question (USQ) regarding a portion of the High Flux Isotope Reactor (HFIR) transient decay heat removal analysis focused on applicability of a heat transfer correlation at the low flow end of reactor operations. During resolution of this issue, review of the correlations used to establish the safety limit (SL) on reactor flux-to-flow ratio revealed the need to change the magnitude of the SL at the low flow end of reactor operations and the need to update the hot spot fuel damage criteria to incorporate current knowledge involving parallel channel flow stability. Because of the original safety design strategy for the reactor, resolution of the issues for the flux-to-flow ratio involved reevaluation of all key process variable SLs and limiting control settings (LCSs) using the current version of the heat transfer analysis code for the reactor. Goals of the work involved updating and upgrading the SL analysis where necessary, while preserving the safety design strategy for the reactor. Changes made include revisions to the safety design criteria at low flows to address the USQ, update of the process- and analysis input-variable uncertainty considerations, and upgrade of the safety design criteria at high flow. The challenges faced during update/upgrade of this SL and LCS are typical of the problems found in the integration of safety into the design process for a complex facility. In particular, the problems addressed in the area of instrument uncertainties provide valuable lessons learned for establishment and configuration control of SLs for large facilities

Temperature and void reactivity coefficient calculations for the high flux isotope reactor safety analysis report

International Nuclear Information System (INIS)

Engle, W.W. Jr.; Williams, L.R.

1994-07-01

This report provides documentation of a series of calculations performed in 1991 in order to provide input for the High Flux Isotope Reactor Safety Analysis Report. In particular, temperature and void reactivity coefficients were calculated for beginning-of-life, end-of-life, and xenon equilibrium (29 h) conditions. Much of the data used to prepare the computer models for these calculations was derived from the original HFIR nuclear design study

Extraction of gadolinium from high flux isotope reactor control plates

International Nuclear Information System (INIS)

Kohring, M.W.

1987-04-01

Gadolinium-153 is an important radioisotope used in the diagnosis of various bone disorders. Recent medical and technical developments in the detection and cure of osteoporosis, a bone disease affecting an estimated 50 million people, have greatly increased the demand for this isotope. The Oak Ridge National Laboratory (ORNL) has produced 153 Gd since 1980 primarily through the irradiation of a natural europium-oxide powder followed by the chemical separation of the gadolinium fraction from the europium material. Due to the higher demand for 153 Gd, an alternative production method to supplement this process has been investigated. This process involves the extraction of gadolinium from the europium-bearing region of highly radioactive, spent control plates used at the High Flux Isotope Reactor (HFIR) with a subsequent re-irradiation of the extracted material for the production of the 153 Gd. Based on the results of experimental and calculational analyses, up to 25 grams of valuable gadolinium (≥60% enriched in 152 Gd) resides in the europium-bearing region of the HFIR control components of which 70% is recoverable. At a specific activity yield of 40 curies of 153 Gd for each gram of gadolinium re-irradiated, 700 one-curie sources can be produced from each control plate assayed

Status of High Flux Isotope Reactor (HFIR) post-restart safety analysis and documentation upgrades

International Nuclear Information System (INIS)

Cook, D.H.; Radcliff, T.D.; Rothrock, R.B.; Schreiber, R.E.

1990-01-01

The High Flux Isotope Reactor (HFIR), an experimental reactor located at the Oak Ridge National Laboratory (ORNL) and operated for the US Department of Energy by Martin Marietta Energy Systems, was shut down in November, 1986 after the discovery of unexpected neutron embrittlement of the reactor vessel. The reactor was restarted in April, 1989, following an extensive review by DOE and ORNL of the HFIR design, safety, operation, maintenance and management, and the implementation of several upgrades to HFIR safety-related hardware, analyses, documents and procedures. This included establishing new operating conditions to provide added margin against pressure vessel failure, as well as the addition, or upgrading, of specific safety-related hardware. This paper summarizes the status of some of the follow-on (post-restart) activities which are currently in progress, and which will result in a comprehensive set of safety analyses and documentation for the HFIR, comparable with current practice in commercial nuclear power plants. 8 refs

Management of safety and risk at the HFIR [High-Flux Isotope Reactor

International Nuclear Information System (INIS)

Glovier, H.A.

1990-01-01

This paper discusses the management of safety and risk at the High-Flux Isotope Reactor (HFIR), a category A research reactor at Oak Ridge National Laboratory (ORNL). The HFIR went critical in 1966 and operated at its designed 100 MW for 20 yr until it was shut down on November 14, 1986. It operated at a >90% availability and without significant event during this period. The result was a complacent management program lacking rigor. This complacency came to an end with the Chernobyl accident, which led to the appointment of an internal committee to assess the safety of ORNL reactor operations. This committee found that HFIR pressure vessel material specimens removed several years earlier had not been analyzed. This issue led to a general review of management practices that were found lacking in quality assurance, safety documentation, training process, and emergency planning, among others. Management accountability was lacking, as shown by design basis and safety analyses that were not up to data and by the fact that reactor operators whose requalification examinations had not been graded were allowed to continue operating the reactor over a long period of time. Between shutdown in 1986 and restart in April 1989, significant management changes and initiatives were made in the area of risk and safety management of ORNL reactors. These are presented briefly in this paper

Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

Energy Technology Data Exchange (ETDEWEB)

Chandler, David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Betzler, Ben [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Hirtz, Gregory John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Sunny, Eva [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

2016-09-01

The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length ²⁵²Cf production targets, ⁷⁵Se production capsules, ⁶³Ni production capsules, a ¹⁸⁸W production capsule, and various materials irradiation targets. Fully loaded ²³⁸Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.

Job/task analysis for I ampersand C [Instrumentation and Controls] instrument technicians at the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Duke, L.L.

1989-09-01

To comply with Department of Energy Order 5480.XX (Draft), a job/task analysis was initiated by the Maintenance Management Department at Oak Ridge National Laboratory (ORNL). The analysis was applicable to instrument technicians working at the ORNL High Flux Isotope Reactor (HFIR). This document presents the procedures and results of that analysis. 2 refs., 2 figs

Production of high-specific activity radionuclides using SM high-flux reactor

International Nuclear Information System (INIS)

Karelin, Ye.A.; Toporov, Yu.G.; Filimonov, V.T.; Vakhetov, F.Z.; Tarasov, V.A.; Kuznetsov, R.A.; Lebedev, V.M.; Andreev, O.I.; Melnik, M.I.; Gavrilov, V.D.

1997-01-01

The development of High Specific Activity (HSA) radionuclides production technologies is one of the directions of RIAR activity, and the high flux research reactor SM, having neutron flux density up to 2.10 15 cm -2 s 1 in a wide range of neutron spectra hardness, plays the principal role in this development. The use of a high-flux reactor for radionuclide production provides the following advantages: production of radionuclides with extremely high specific activity, decreasing of impurities content in irradiated targets (both radioactive and non-radioactive), cost-effective use of expensive isotopically enriched target materials. The production technologies of P-33, Gd-153, W-188, Ni-63, Fe-55,59, Sn-113,117m,119m, Sr- 89, applied in industry, nuclear medicine, research, etc, were developed by RIAR during the last 5-10 years. The research work included the development of calculation procedures for radionuclide reactor accumulation forecast, experimental determination of neutron cross-sections, the development of irradiated materials reprocessing procedures, isolation and purification of radionuclides. The principal results are reviewed in the paper. (authors)

Short-lived radionuclides produced on the ORNL 86-inch cyclotron and High-Flux Isotope Reactor

International Nuclear Information System (INIS)

Lamb, E.

1985-01-01

The production of short-lived radionuclides at ORNL includes the preparation of target materials, irradiation on the 86-in. cyclotron and in the High Flux Isotope Reactor (HFIR), and chemical processing to recover and purify the product radionuclides. In some cases the target materials are highly enriched stable isotopes separated on the ORNL calutrons. High-purity 123 I has been produced on the 86-in. cyclotron by irradiating an enriched target of 123 Te in a proton beam. Research on calutron separations has led to a 123 Te product with lower concentrations of 124 Te and 126 Te and, consequently to lower concentrations of the unwanted radionuclides, 124 I and 126 I, in the 123 I product. The 86-in. cyclotron accelerates a beam of protons only but is unique in providing the highest available beam current of 1500 μA at 21 MeV. This beam current produces relatively large quantities of radionuclides such as 123 I and 67 Ga

The High Flux Isotope Reactor (HFIR) cold source project at ORNL

International Nuclear Information System (INIS)

Selby, D.L.; Lucas, A.T.; Chang, S.J.; Freels, J.D. . E-mail-yb2@ornl.gov

1998-01-01

Following the decision to cancel the Advanced Neutron Source (ANS) Project at Oak Ridge National Laboratory (ORNL), it was determined that a hydrogen cold source should be retrofitted into an existing beam tube of the High Flux Isotope Reactor (HFIR) at ORNL. The preliminary design of this system has been completed and an 'approval in principle' of the design has been obtained from the internal ORNL safety review committees and the U.S. Department of Energy (DOE) safety review committee. The cold source concept is basically a closed loop forced flow supercritical hydrogen system. The supercritical approach was chosen because of its enhanced stability in the proposed high heat flux regions. Neutron and gamma physics of the moderator have been analyzed using the 3D Monte Carlo code MCNP 1 A D structural analysis model of the moderator vessel, vacuum tube, and beam tube was completed to evaluate stress loadings and to examine the impact of hydrogen detonations in the beam tube. A detailed ATHENA 2 system model of the hydrogen system has been developed to simulate loop performance under normal and off-normal transient conditions. Semi-prototypic hydrogen loop tests of the system have been performed at the Arnold Engineering Design Center (AEDC) located in Tullahoma, Tennessee to verify the design and benchmark the analytical system model. A 3.5 kW refrigerator system has been ordered and is expected to be delivered to ORNL by the end of this calendar year. Our present schedule shows the assembling of the cold source loop on site during the fall of 1999 for final testing before insertion of the moderator plug assembly into the reactor beam tube during the end of the year 2000. (author)

Design Study for a Low-Enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2008

Energy Technology Data Exchange (ETDEWEB)

Primm, Trent [ORNL; Chandler, David [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL; Jolly, Brian C [ORNL

2009-03-01

This report documents progress made during FY 2008 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in reactor performance from the current level. Results of selected benchmark studies imply that calculations of LEU performance are accurate. Scoping experiments with various manufacturing methods for forming the LEU alloy profile are presented.

The method of life extension for the High Flux Isotope Reactor vessel

International Nuclear Information System (INIS)

Chang, Shib-Jung.

1995-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A hydrostatic pressure test (hydrotest) is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the reactor life from the safe hydrostatic pressure. The life extension calculation provides the following information on the remaining life of the reactor as a function of the nil ductility temperature increase: the probability of vessel fracture due to hydrotest vs vessel life at several hydrotest pressures; the hydrotest time interval vs the uncertainty of the nil ductility temperature increase rate; and the hydrotest pressure vs the uncertainty of the nil ductility temperature increase rate. It is understood that the use of a complete range of uncertainties of the nil ductility temperature increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85 degree F

Production of Thorium-229 at the ORNL High Flux Isotope Reactor

International Nuclear Information System (INIS)

Boll, Rose Ann; Garland, Marc A.; Mirzadeh, Saed

2008-01-01

The investigation of targeted cancer therapy using -emitters has developed considerably in recent years and clinical trials have generated promising results. In particular, the initial clinical trials for treatment of acute myeloid leukemia have demonstrated the effectiveness of the -emitter 213Bi in killing cancer cells. Pre-clinical studies have also shown the potential application of both 213Bi and its 225Ac parent radionuclide in a variety of cancer systems and targeted radiotherapy. Bismuth-213 is obtained from a radionuclide generator system from decay of the 10-d 225Ac parent, a member of the 7340-y 229Th chain. Currently, 233U is the only viable source for high purity 229Th; however, due to increasing difficulties associated with 233U safeguards, processing additional 233U is presently unfeasible. The recent decision to downblend and dispose of enriched 233U further diminished the prospects for extracting 229Th from 233U stock. Nevertheless, the anticipated growth in demand for 225Ac may soon exceed the levels of 229Th (∼40 g or ∼8 Ci; ∼80 times the current ORNL 229Th stock) present in the aged 233U stockpile. The alternative routes for the production of 229Th, 225Ra and 225Ac include both reactor and accelerator approaches. Here, we describe production of 229Th via neutron transmutation of 226Ra targets in the ORNL High Flux Isotope Reactor (HFIR).

Delivery of completed irradiation vehicles and the quality assurance document to the High Flux Isotope Reactor for irradiation

Energy Technology Data Exchange (ETDEWEB)

Petrie, Christian M. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); McDuffee, Joel Lee [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Katoh, Yutai [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

2015-10-01

This report details the initial fabrication and delivery of two Fuel Cycle Research and Development (FCRD) irradiation capsules (ATFSC01 and ATFSC02), with associated quality assurance documentation, to the High Flux Isotope Reactor (HFIR). The capsules and documentation were delivered by September 30, 2015, thus meeting the deadline for milestone M3FT-15OR0202268. These irradiation experiments are testing silicon carbide composite tubes in order to obtain experimental validation of thermo-mechanical models of stress states in SiC cladding irradiated under a prototypic high heat flux. This document contains a copy of the completed capsule fabrication request sheets, which detail all constituent components, pertinent drawings, etc., along with a detailed summary of the capsule assembly process performed by the Thermal Hydraulics and Irradiation Engineering Group (THIEG) in the Reactor and Nuclear Systems Division (RNSD). A complete fabrication package record is maintained by the THIEG and is available upon request.

Instrumentation for the advanced high-flux reactor workshop: proceedings

International Nuclear Information System (INIS)

Moon, R.M.

1984-01-01

The purpose of the Workshop on Instrumentation for the Advanced High-Flux Reactor, held on May 30, 1984, at the Oak Ridge National Laborattory, was two-fold: to announce to the scientific community that ORNL has begun a serious effort to design and construct the world's best research reactor, and to solicit help from the scientific community in planning the experimental facilities for this reactor. There were 93 participants at the workshop. We are grateful to the visiting scientists for their enthusiasm and interest in the reactor project. Our goal is to produce a reactor with a peak thermal flux in a large D 2 O reflector of 5 x 10 15 n/cm 2 s. This would allow the installation of unsurpassed facilities for neutron beam research. At the same time, the design will provide facilities for isotope production and materials irradiation which are significantly improved over those now available at ORNL. This workshop focussed on neutron beam facilities; the input from the isotope and materials irradiation communities will be solicited separately. The reactor project enjoys the full support of ORNL management; the present activities are financed by a grant of $663,000 from the Director's R and D Fund. However, we realize that the success of the project, both in realization and in use of the reactor, depends on the support and imagination of a broad segment of the scientific community. This is more a national project than an ORNL project. The reactor would be operated as a national user facility, open to any research proposal with high scientific merit. It is therefore important that we maintain a continuing dialogue with outside scientists who will be the eventual users of the reactor and the neutron beam facilities. The workshop was the first step in establishing this dialogue; we anticipate further workshops as the project continues

Total quality management for addressing suspect parts at the Oak Ridge High Flux Isotope Reactor

International Nuclear Information System (INIS)

Hendrix, K.A.; Tulay, M.P.

1993-01-01

Martin Marietta Energy System (MMES) Research Reactors Division (RRD), operator of the High Flux Isotope Reactor (HFIR) recently embarked on an aggressive Program to address the issue of suspect Parts and to enhance their procurement process. Through the application of TQM process improvement, RRD has already achieved improved efficiency in specifying, procuring, and accepting replacement items for its largest research reactor. These process improvements have significantly decreased the risk of installing suspect parts in the HFIR safety systems. To date, a systematic plan has been implemented, which includes the following elements: Process assessment and procedure review; Procedural enhancements; On-site training and technology transfer; Enhanced receiving inspections; Performance supplier evaluations and source verifications integrated processes for utilizing commercial grade products in nuclear safety-related applications. This paper will describe the above elements, how a partnership between MMES and Gilbert/Commonwealth facilitated the execution of the plan, and how process enhancements were applied. We will also present measures for improved efficiency and productivity, that MMES intends to continually address with Quality Action Teams

Establishing a Cost Basis for Converting the High Flux Isotope Reactor from High Enriched to Low Enriched Uranium Fuel

International Nuclear Information System (INIS)

Primm, Trent; Guida, Tracey

2010-01-01

Under the auspices of the Global Threat Reduction Initiative Reduced Enrichment for Research and Test Reactors Program, the National Nuclear Security Administration/Department of Energy (NNSA/DOE) has, as a goal, to convert research reactors worldwide from weapons grade to non-weapons grade uranium. The High Flux Isotope Reactor (HFIR) at Oak Ridge National Lab (ORNL) is one of the candidates for conversion of fuel from high enriched uranium (HEU) to low enriched uranium (LEU). A well documented business model, including tasks, costs, and schedules was developed to plan the conversion of HFIR. Using Microsoft Project, a detailed outline of the conversion program was established and consists of LEU fuel design activities, a fresh fuel shipping cask, improvements to the HFIR reactor building, and spent fuel operations. Current-value costs total $76 million dollars, include over 100 subtasks, and will take over 10 years to complete. The model and schedule follows the path of the fuel from receipt from fuel fabricator to delivery to spent fuel storage and illustrates the duration, start, and completion dates of each subtask to be completed. Assumptions that form the basis of the cost estimate have significant impact on cost and schedule.

Calculations for HFIR [High Flux Isotope Reactor] fuel plate non- bonding and fuel segregation uncertainty factors

International Nuclear Information System (INIS)

Kirkpatrick, J.R.

1990-10-01

The effects of non-bonds and of fuel segregation on the package factors of the heat flux in the High Flux Isotope Reactor (HFIR) are examined. The effects of the two defects are examined both separately and together. It is concluded that the peaking factors that are used in the present HFIR thermal analysis code are conservative and thus no changes in the peaking factors are necessary to continue to ensure that HFIR is safe. A study was made of the effect of the non-bond spot diameter on the peaking factor. The conclusion is that the spot can have diameter more than three times the maximum value allowed by the specifications before the peaking factor is greater than the maximum value specified in the present HFIR thermal analysis code. 6 refs., 7 figs., 8 tabs

Dynamic response of the high flux isotope reactor structure caused by nearby heavy load drop

International Nuclear Information System (INIS)

Chang, Shih-Jung.

1995-01-01

A heavy load of 50,000 lb is assumed to drop from 10 ft above the bottom of the High Flux Isotope Reactor (HFIR) pool at the loading station. The consequences of the dynamic impact to the bottom slab of the pool and to the nearby HFIR reactor vessel are analyzed by applying the ABAQUS computer code The results show that both the BM vessel structure and its supporting legs are subjected to elastic disturbances only and, therefore, will not be damaged. The bottom slab of the pool, however, will be damaged to about half of the slab thickness. The velocity response spectrum at the concrete floor next to the HFIR vessel as a result of the vibration caused by the impact is obtained. It is concluded, that the damage caused by heavy load drop at the loading station is controlled by the slab damage and the nearby HFIR vessel and the supporting legs will not be damaged

Design and use of the ORNL HFIR [High Flux Isotope Reactor] pneumatic tube irradiation systems

International Nuclear Information System (INIS)

Dyer, F.F.; Emery, J.F.; Robinson, L.; Teasley, N.A.

1987-01-01

A second pneumatic tube that was recently installed in the High Flux Isotope Reactor for neutron activation analysis is described. Although not yet tested, the system is expected to have a thermal neutron flux of about 1.5 x 10 14 cm -2 s -1 . A delayed neutron counter is an integral part of the pneumatic tube, and all of the hardware is present to enable automated use of the counter. The system is operated with a Gould programmable controller that is programmed with an IBM personal computer. Automation of any mode of operation, including the delayed neutron counter, will only require a nominal amount of software development. Except for the lack of a hot cell, the irradiation facility has all of the advantageous features of an older pneumatic tube that has been in operation for 17 years. The design of the system and some applications and methods of operation are described

High Flux Isotope Reactor cold neutron source reference design concept

International Nuclear Information System (INIS)

Selby, D.L.; Lucas, A.T.; Hyman, C.R.

1998-05-01

In February 1995, Oak Ridge National Laboratory's (ORNL's) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH 2 ) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH 2 cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

Energy Technology Data Exchange (ETDEWEB)

Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

2012-03-01

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Low-Enriched Uranium Fuel Conversion Activities for the High Flux Isotope Reactor, Annual Report for FY 2011

International Nuclear Information System (INIS)

Renfro, David G.; Cook, David Howard; Freels, James D.; Griffin, Frederick P.; Ilas, Germina; Sease, John D.; Chandler, David

2012-01-01

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

High Flux Isotope Reactor cold neutron source reference design concept

Energy Technology Data Exchange (ETDEWEB)

Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

1998-05-01

In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

Energy Technology Data Exchange (ETDEWEB)

Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

2014-09-01

Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

High Flux Isotopes Reactor (HFIR) Cooling Towers Demolition Waste Management

Energy Technology Data Exchange (ETDEWEB)

Pudelek, R. E.; Gilbert, W. C.

2002-02-26

This paper describes the results of a joint initiative between Oak Ridge National Laboratory, operated by UT-Battelle, and Bechtel Jacobs Company, LLC (BJC) to characterize, package, transport, treat, and dispose of demolition waste from the High Flux Isotope Reactor (HFIR), Cooling Tower. The demolition and removal of waste from the site was the first critical step in the planned HFIR beryllium reflector replacement outage scheduled. The outage was scheduled to last a maximum of six months. Demolition and removal of the waste was critical because a new tower was to be constructed over the old concrete water basin. A detailed sampling and analysis plan was developed to characterize the hazardous and radiological constituents of the components of the Cooling Tower. Analyses were performed for Resource Conservation and Recovery Act (RCRA) heavy metals and semi-volatile constituents as defined by 40 CFR 261 and radiological parameters including gross alpha, gross beta, gross gamma, alpha-emitting isotopes and beta-emitting isotopes. Analysis of metals and semi-volatile constituents indicated no exceedances of regulatory limits. Analysis of radionuclides identified uranium and thorium and associated daughters. In addition 60Co, 99Tc, 226Rm, and 228Rm were identified. Most of the tower materials were determined to be low level radioactive waste. A small quantity was determined not to be radioactive, or could be decontaminated. The tower was dismantled October 2000 to January 2001 using a detailed step-by-step process to aid waste segregation and container loading. The volume of waste as packaged for treatment was approximately 1982 cubic meters (70,000 cubic feet). This volume was comprised of plastic ({approx}47%), wood ({approx}38%) and asbestos transite ({approx}14%). The remaining {approx}1% consisted of the fire protection piping (contaminated with lead-based paint) and incidental metal from conduit, nails and braces/supports, and sludge from the basin. The waste

Report of the ANS Project Feasibility Workshop for a High Flux Isotope Reactor-Center for Neutron Research Facility

International Nuclear Information System (INIS)

Peretz, F.J.; Booth, R.S.

1995-07-01

The Advanced Neutron Source (ANS) Conceptual Design Report (CDR) and its subsequent updates provided definitive design, cost, and schedule estimates for the entire ANS Project. A recent update to this estimate of the total project cost for this facility was $2.9 billion, as specified in the FY 1996 Congressional data sheet, reflecting a line-item start in FY 1995. In December 1994, ANS management decided to prepare a significantly lower-cost option for a research facility based on ANS which could be considered during FY 1997 budget deliberations if DOE or Congressional planners wished. A cost reduction for ANS of about $1 billion was desired for this new option. It was decided that such a cost reduction could be achieved only by a significant reduction in the ANS research scope and by maximum, cost-effective use of existing High Flux Isotope Reactor (HFIR) and ORNL facilities to minimize the need for new buildings. However, two central missions of the ANS -- neutron scattering research and isotope production-were to be retained. The title selected for this new option was High Flux Isotope Reactor-Center for Neutron Research (HFIR-CNR) because of the project's maximum use of existing HFIR facilities and retention of selected, central ANS missions. Assuming this shared-facility requirement would necessitate construction work near HFIR, it was specified that HFIR-CNR construction should not disrupt normal operation of HFIR. Additional objectives of the study were that it be highly credible and that any material that might be needed for US Department of Energy (DOE) and Congressional deliberations be produced quickly using minimum project resources. This requirement made it necessary to rely heavily on the ANS design, cost, and schedule baselines. A workshop methodology was selected because assessment of each cost and/or scope-reduction idea required nearly continuous communication among project personnel to ensure that all ramifications of propsed changes

Production of medical radioisotopes in the ORNL High Flux Isotope Reactor (HFIR) for cancer treatment and arterial restenosis therapy after PTCA

International Nuclear Information System (INIS)

Knapp, F.F. Jr.; Beets, A.L.; Mirzadeh, S.; Alexander, C.W.; Hobbs, R.L.

1998-01-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) represents an important resource for the production of a wide variety of medical radioisotopes. In addition to serving as a key production site for californium-252 and other transuranic elements, important examples of therapeutic radioisotopes which are currently routinely produced in the HFIR for distribution include dysprosium-166 (parent of holmium-166), rhenium-186, tin-117m and tungsten-188 (parent of rhenium-188). The nine hydraulic tube (HT) positions in the central high flux region permit the insertion and removal of targets at any time during the operating cycle and have traditionally represented a major site for production of medical radioisotopes. To increase the irradiation capabilities of the HFIR, special target holders have recently been designed and fabricated which will be installed in the six Peripheral Target Positions (PTP), which are also located in the high flux region. These positions are only accessible during reactor refueling and will be used for long-term irradiations, such as required for the production of tin-117m and tungsten-188. Each of the PTP tubes will be capable of housing a maximum of eight HT targets, thus increasing the total maximum number of HT targets from the current nine, to a total of 57. In this paper the therapeutic use of reactor-produced radioisotopes for bone pain palliation and vascular brachytherapy and the therapeutic medical radioisotope production capabilities of the ORNL HFIR are briefly discussed

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

DESIGN STUDY FOR A LOW-ENRICHED URANIUM CORE FOR THE HIGH FLUX ISOTOPE REACTOR, ANNUAL REPORT FOR FY 2010

Energy Technology Data Exchange (ETDEWEB)

Cook, David Howard [ORNL; Freels, James D [ORNL; Ilas, Germina [ORNL; Jolly, Brian C [ORNL; Miller, James Henry [ORNL; Primm, Trent [ORNL; Renfro, David G [ORNL; Sease, John D [ORNL; Pinkston, Daniel [ORNL

2011-02-01

This report documents progress made during FY 2010 in studies of converting the High Flux Isotope Reactor (HFIR) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. With axial and radial grading of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current level. Studies are reported of support to a thermal hydraulic test loop design, the implementation of finite element, thermal hydraulic analysis capability, and infrastructure tasks at HFIR to upgrade the facility for operation at 100 MW. A discussion of difficulties with preparing a fuel specification for the uranium-molybdenum alloy is provided. Continuing development in the definition of the fuel fabrication process is described.

Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

Energy Technology Data Exchange (ETDEWEB)

Travis, Adam R [ORNL

2014-05-01

A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

Dissolution flowsheet for high flux isotope reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Foster, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U₃O₈) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H₂. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H₂) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H₂ and other permanent gases in the dissolution offgas allowing the development of H₂ generation rate profiles. The H₂ generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

High flux isotope reactor cold source preconceptual design study report

International Nuclear Information System (INIS)

Selby, D.L.; Bucholz, J.A.; Burnette, S.E.

1995-12-01

In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH 2 moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

Energy Technology Data Exchange (ETDEWEB)

Primm, R.T., III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N. (U. of Cincinnati)

2006-02-01

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U{sub 3}O{sub 8} mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties.

Assumptions and Criteria for Performing a Feasability Study of the Conversion of the High Flux Isotope Reactor Core to Use Low-Enriched Uranium Fuel

International Nuclear Information System (INIS)

Primm, R.T. III; Ellis, R.J.; Gehin, J.C.; Moses, D.L.; Binder, J.L.; Xoubi, N.

2006-01-01

A computational study will be initiated during fiscal year 2006 to examine the feasibility of converting the High Flux Isotope Reactor from highly enriched uranium fuel to low-enriched uranium. The study will be limited to steady-state, nominal operation, reactor physics and thermal-hydraulic analyses of a uranium-molybdenum alloy that would be substituted for the current fuel powder--U 3 O 8 mixed with aluminum. The purposes of this document are to (1) define the scope of studies to be conducted, (2) define the methodologies to be used to conduct the studies, (3) define the assumptions that serve as input to the methodologies, (4) provide an efficient means for communication with the Department of Energy and American research reactor operators, and (5) expedite review and commentary by those parties

Use of sup(233)U for high flux reactors

International Nuclear Information System (INIS)

Sekimoto, Hiroshi; Liem, P.H.

1991-01-01

The feasibility design study on the graphite moderated gas cooled reactor as a high flux reactor has been performed. The core of the reactor is equipped with two graphite reflectors, i.e., the inner reflector and the outer reflector. The highest value of the thermal neutron flux and moderately high thermal neutron flux are expected to be achieved in the inner reflector region and in the outer reflector region respectively. This reactor has many merits comparing to the conventional high flux reactors. It has the inherent safety features associated with the modular high temperature reactors. Since the core is composed with pebble bed, the on-power refueling can be performed and the experiment time can be chosen as long as necessary. Since the thermal-to-fast flux ratio is large, the background neutron level is low and material damage induced by fast neutrons are small. The calculation was performed using a four groups diffusion approximation in a one-dimensional spherical geometry and a two-dimensional cylindrical geometry. By choosing the optimal values of the core-reflector geometrical parameters and moderator-to-fuel atomic density, high thermal neutron flux can be obtained. Because of the thermal neutron flux can be obtained. Because of the thermal design constraint, however, this design will produce a relatively large core volume (about 10 7 cc) and consequently a higher reactor power (100 MWth). Preliminary calculational results show that with an average power density of only 10 W/cc, maximum thermal neutron flux of 10 15 cm -2 s -1 can be achieved in the inner reflector. The eta value of 233 U is larger than 235 U. By introducing 233 U as the fissile material for this reactor, the thermal neutron flux level can be increased by about 15%. (author). 3 refs., 2 figs., 4 tabs

Probability of fracture and life extension estimate of the high-flux isotope reactor vessel

International Nuclear Information System (INIS)

Chang, S.J.

1998-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in ductile-brittle transition temperature (DBTT) for fracture, often denoted by RT NDT for carbon steel. This transition temperature can be calibrated by the drop-weight test and, sometimes, by the Charpy impact test. The life extension for the high-flux isotope reactor (HFIR) vessel is calculated by using the method of fracture mechanics that is incorporated with the effect of the DBTT change. The failure probability of the HFIR vessel is limited as the life of the vessel by the reactor core melt probability of 10 -4 . The operating safety of the reactor is ensured by periodic hydrostatic pressure test (hydrotest). The hydrotest is performed in order to determine a safe vessel static pressure. The fracture probability as a result of the hydrostatic pressure test is calculated and is used to determine the life of the vessel. Failure to perform hydrotest imposes the limit on the life of the vessel. The conventional method of fracture probability calculations such as that used by the NRC-sponsored PRAISE CODE and the FAVOR CODE developed in this Laboratory are based on the Monte Carlo simulation. Heavy computations are required. An alternative method of fracture probability calculation by direct probability integration is developed in this paper. The present approach offers simple and expedient ways to obtain numerical results without losing any generality. In this paper, numerical results on (1) the probability of vessel fracture, (2) the hydrotest time interval, and (3) the hydrotest pressure as a result of the DBTT increase are obtained

Scientific upgrades at the high flux isotope reactor at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Selby, D.L.; Garrett, D.L.; Lucas, A.T.; Reeves, M.E.

2001-01-01

The United States Department of Energy is sponsoring a number of projects that will provide scientific upgrades to the neutron science facilities associated with the high flux isotope reactor (HFIR) located at Oak Ridge National Laboratory. Funding for the first upgrade project was initiated in 1996 and all presently identified upgrade projects are expected to be completed by the end of 2003. The upgrade projects include: 1) larger beam tubes, 2) a new monochromator drum for the HB-1 beam line, 3) a new HB-2 beam line system that includes one thermal guide and a new monochromator drum, 4) new instruments for the HB-2 beamline, 5) a new monochromator drum for the HB-3 beam line, 6) a supercritical hydrogen cold source system to be retrofitted into the HB-4 beam tube, 7) a 3.5 kW refrigeration system at 20 K to support the cold source and a new building to house it, 8) a new HB-4 beam line system composed of four cold neutron guides with various mirror coatings and associated shielding, 9) a number of new instruments for the cold beams including two new SANS instruments, and 10) construction of support buildings. This paper provides a short summary of these projects including their present status and schedule. (orig.)

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

Analysis and modeling of flow-blockage-induced steam explosion events in the high-flux isotope reactor

International Nuclear Information System (INIS)

Taleyarkhan, R.P.; Georgevich, V.; Nestor, C.W.; Gat, U.; Lepard, B.L.; Cook, D.H.; Freels, J.; Chang, S.J.; Luttrell, C.; Gwaltney, R.C.

1994-01-01

This article provides a perspective overview of the analysis and modeling work done to evaluate the threat from steam explosion loads in the High-Flux Isotope Reactor (HFIR) during flow blockage events. The overall work scope included modeling and analysis of core-melt initiation, melt propagation, bounding and best-estimate steam explosion energetics, vessel failure from fracture, bolts failure from exceedance of elastic limits, and, finally, missile evolution and transport. Aluminum ignition was neglected. Evaluations indicated that a thermally driven steam explosion with more than 65 MJ of energy insertion in the core region over several milliseconds would be needed to cause a sufficiently energetic missile with a capacity to cause early confinement failure. This amounts to about 65% of the HFIR core mass melting and participating in a steam explosion. Conservative melt propagation analyses have indicated that at most only 24% of the HFIR core mass could melt during flow blockage events under full-power conditions. 19 refs., 11 figs

Design Study for a Low-enriched Uranium Core for the High Flux Isotope Reactor, Annual Report for FY 2007

Energy Technology Data Exchange (ETDEWEB)

Primm, Trent [ORNL; Ellis, Ronald James [ORNL; Gehin, Jess C [ORNL; Ilas, Germina [ORNL; Miller, James Henry [ORNL; Sease, John D [ORNL

2007-11-01

This report documents progress made during fiscal year 2007 in studies of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) fuel to low enriched uranium fuel (LEU). Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum alloy. A high volume fraction U/Mo-in-Al fuel could attain the same neutron flux performance as with the current, HEU fuel but materials considerations appear to preclude production and irradiation of such a fuel. A diffusion barrier would be required if Al is to be retained as the interstitial medium and the additional volume required for this barrier would degrade performance. Attaining the high volume fraction (55 wt. %) of U/Mo assumed in the computational study while maintaining the current fuel plate acceptance level at the fuel manufacturer is unlikely, i.e. no increase in the percentage of plates rejected for non-compliance with the fuel specification. Substitution of a zirconium alloy for Al would significantly increase the weight of the fuel element, the cost of the fuel element, and introduce an as-yet untried manufacturing process. A monolithic U-10Mo foil is the choice of LEU fuel for HFIR. Preliminary calculations indicate that with a modest increase in reactor power, the flux performance of the reactor can be maintained at the current level. A linearly-graded, radial fuel thickness profile is preferred to the arched profile currently used in HEU fuel because the LEU fuel media is a metal alloy foil rather than a powder. Developments in analysis capability and nuclear data processing techniques are underway with the goal of verifying the preliminary calculations of LEU flux performance. A conceptual study of the operational cost of an LEU fuel fabrication facility yielded the conclusion that the annual fuel cost to the HFIR would increase significantly from the current, HEU fuel cycle. Though manufacturing can be accomplished with existing technology

Testing of research reactor fuel in the high flux reactor (Petten)

International Nuclear Information System (INIS)

Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

1999-01-01

The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

Probabilistic fracture mechanics analysis for the life extension estimate of the high flux isotope reactor vessel

International Nuclear Information System (INIS)

Chang, S.J.

1997-01-01

The state of the vessel steel embrittlement as a result of neutron irradiation can be measured by its increase in the nil ductility temperature (NDT). This temperature is sometimes referred to as the brittle-ductile transition temperature (DBT) for fracture. The life extension of the High Flux Isotope Reactor (HFIR) vessel is calculated by using the method of fracture mechanics. A new method of fracture probability calculation is presented in this paper. The fracture probability as a result of the hydrostatic pressure test (hydrotest) is used to determine the life of the vessel. The hydrotest is performed in order to determine a safe vessel static pressure. It is then followed by using fracture mechanics to project the safe reactor operation time from the time of the satisfactory hydrostatic test. The life extension calculation provides the following information on the remaining life of the reactor as a function of the NDT increase: (1) the life of the vessel is determined by the probability of vessel fracture as a result of hydrotest at several hydrotest pressures and vessel embrittlement conditions, (2) the hydrotest time interval vs the NDT increase rate, and (3) the hydrotest pressure vs the NDT increase rate. It is understood that the use of a complete range of uncertainties of the NDT increase is equivalent to the entire range of radiation damage that can be experienced by the vessel steel. From the numerical values for the probabilities of the vessel fracture as a result of hydrotest, it is estimated that the reactor vessel life can be extended up to 50 EFPY (100 MW) with the minimum vessel operating temperature equal to 85 degrees F

Isotopic alloying to tailor helium production rates in mixed spectrum reactors

International Nuclear Information System (INIS)

Mansur, L.K.; Rowcliffe, A.F.; Grossbeck, M.L.; Stoller, R.E.

1985-01-01

The purposes of this work are to increase the understanding of mechanisms by which helium affects microstructure and properties, to aid in the development of materials for fusion reactors, and to obtain data from fission reactors in regimes of direct interest for fusion reactor applications. Isotopic alloying is examined as a means of manipulating the ratio of helium transmutations to atom displacements in mixed spectrum reactors. The application explored is based on artificially altering the relative abundances of the stable isotopes of nickel to systematically vary the fraction of 58 Ni in nickel bearing alloys. The method of calculating helium production rates is described. Results of example calculations for proposed experiments in the High Flux Isotope Reactor are discussed

On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

International Nuclear Information System (INIS)

Freels, J.D.

1993-01-01

This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ''the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed

Selection of support structure materials for irradiation experiments in the HFIR [High Flux Isotope Reactor] at temperatures up to 500 degrees C

International Nuclear Information System (INIS)

Farrell, K.; Longest, A.W.

1990-01-01

The key factor in the design of capsules for irradiation of test specimens in the High Flux Isotope Reactor at preselected temperatures up to 500 degree C utilizing nuclear heating is a narrow gas-filled gap which surrounds the specimens and controls the transfer of heat from the specimens through the wall of a containment tube to the reactor cooling water. Maintenance of this gap to close tolerances is dependent on the characteristics of the materials used to support the specimens and isolate them from the water. These support structure materials must have low nuclear heating rates, high thermal conductivities, and good dimensional stabilities under irradiation. These conditions are satisfied by certain aluminum alloys. One of these alloys, a powder metallurgy product containing a fine dispersion of aluminum oxide, is no longer manufactured. A new alloys of this type, with the trade name DISPAL, is determined to be a suitable substitute. 23 refs., 13 figs., 3 tabs

Reactor production of 252Cf and transcurium isotopes

International Nuclear Information System (INIS)

Alexander, C.W.; Halperin, J.; Walker, R.L.; Bigelow, J.E.

1990-01-01

Berkelium, californium, einsteinium, and fermium are currently produced in the High Flux Isotope Reactor (HFIR) and recovered in the Radiochemical Engineering Development Center (REDC) at the Oak Ridge National Laboratory (ORNL). All the isotopes are used for research. In addition, 252 Cf, 253 Es, and 255 Fm have been considered or are used for industrial or medical applications. ORNL is the sole producer of these transcurium isotopes in the western world. A wide range of actinide samples were irradiated in special test assemblies at the Fast Flux Test Facility (FFTF) at Hanford, Washington. The purpose of the experiments was to evaluate the usefulness of the two-group flux model for transmutations in the special assemblies with an eventual goal of determining the feasibility of producing macro amounts of transcurium isotopes in the FFTF. Preliminary results from the production of 254g Es from 252 Cf will be discussed. 14 refs., 5 tabs

Reactor operations Brookhaven medical research reactor, Brookhaven high flux beam reactor informal monthly report

International Nuclear Information System (INIS)

Hauptman, H.M.; Petro, J.N.; Jacobi, O.

1995-04-01

This document is the April 1995 summary report on reactor operations at the Brookhaven Medical Research Reactor and the Brookhaven High Flux Beam Reactor. Ongoing experiments/irradiations in each are listed, and other significant operations functions are also noted. The HFBR surveillance testing schedule is also listed

Analysis and modeling of flow blockage-induced steam explosion events in the High-Flux Isotope Reactor

International Nuclear Information System (INIS)

Taleyarkhan, R.P.; Georgevich, V.; Lestor, C.W.; Gat, U.; Lepard, B.L.; Cook, D.H.; Freels, J.; Chang, S.J.; Luttrell, C.; Gwaltney, R.C.; Kirkpatrick, J.

1993-01-01

This paper provides a perspective overview of the analysis and modeling work done to evaluate the threat from steam explosion loads in the High-Flux Isotope Reactor during flow blockage events. The overall workscope included modeling and analysis of core melt initiation, melt propagation, bounding and best-estimate steam explosion energetics, vessel failure from fracture, bolts failure from exceedance of elastic limits, and finally, missile evolution and transport. Aluminum ignition was neglected. Evaluations indicated that a thermally driven steam explosion with more than 65 MJ of energy insertion in the core region over several miliseconds would be needed to cause a sufficiently energetic missile with a capacity to cause early confinement failure. This amounts to about 65% of the HFIR core mass melting and participating in a steam explosion. Conservative melt propagation analyses have indicated that at most only 24% of the HFIR core mass could melt during flow blockage events under full-power conditions. Therefore, it is judged that the HFIR vessel and top head structure will be able to withstand loads generated from thermally driven steam explosions initiated by any credible flow blockage event. A substantial margin to safety was demonstrated

Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

Energy Technology Data Exchange (ETDEWEB)

Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L., E-mail: mabs500@gmail.com, E-mail: narain@ufpe.br, E-mail: jairbezerra@gmail.com [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Centro de Tecnologia e Geociências. Departamento de Energia Nuclear

2017-07-01

Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

Analysis of the neutron flux in an annular pulsed reactor by using finite volume method

International Nuclear Information System (INIS)

Silva, Mário A.B. da; Narain, Rajendra; Bezerra, Jair de L.

2017-01-01

Production of very intense neutron sources is important for basic nuclear physics and for material testing and isotope production. Nuclear reactors have been used as sources of intense neutron fluxes, although the achievement of such levels is limited by the inability to remove fission heat. Periodic pulsed reactors provide very intense fluxes by a rotating modulator near a subcritical core. A concept for the production of very intense neutron fluxes that combines features of periodic pulsed reactors and steady state reactors was proposed by Narain (1997). Such a concept is known as Very Intense Continuous High Flux Pulsed Reactor (VICHFPR) and was analyzed by using diffusion equation with moving boundary conditions and Finite Difference Method with Crank-Nicolson formalism. This research aims to analyze the flux distribution in the Very Intense Continuous Flux High Pulsed Reactor (VICHFPR) by using the Finite Volume Method and compares its results with those obtained by the previous computational method. (author)

Calculation of the transmutation rates of Tc-99, I-129 and Cs-135 in the High Flux Reactor, in the Phenix Reactor and in a light water reactor

International Nuclear Information System (INIS)

Bultman, J.

1992-04-01

Transmutation of long-lived fission products is of interest for the reduction of the possible dose to the population resulting from long-term leakage of nuclear waste from waste disposals. Three isotopes are of special interest: Tc-99, I-129 and Cs-135. Therefore, experiments on transmutation of these isotopes in nuclear reactors are planned. In the present study, the possible transmutation rates and mass reductions are determined for experiments in High Flux Reactor (HFR) located in Petten (Netherlands) and in Phenix (France). Also, rates were determined for a standard Light Water Reactor (LWR). The transmutation rates of the 3 fission products will be much higher in HFR than in Phenix reactor, as both total flux and effective cross sections are higher. For thick targets the effective half lives are approximately 3, 2 and 7 years for Tc-99, I-129 and Cs-135 irradiation respectively in HFR and 22, 16 and 40 years for Tc-99, I-129 and Cs-135 irradiation in Phenix reactor. The transmutation rates in LWR are low. Only the relatively large power of LWR guarantees a large total mass reduction. Especially transmutation of Cs-135 will be very difficult in Phenix and LWR, clearly shown by the very long effective half lives of 40 and 100 years, respectively. (author). 7 refs.; 5 figs.; 7 tabs

Preliminary considerations of an intense slow positron facility based on a 78Kr loop in the high flux isotopes reactor

International Nuclear Information System (INIS)

Hulett, L.D. Jr.; Donohue, D.L.; Peretz, F.J.; Montgomery, B.H.; Hayter, J.B.

1990-01-01

Suggestions have been made to the National Steering Committee for the Advanced Neutron Source (ANS) by Mills that provisions be made to install a high intensity slow positron facility, based on a 78 Kr loop, that would be available to the general community of scientists interested in this field. The flux of thermal neutrons calculated for the ANS is E + 15 sec -1 m -2 , which Mills has estimated will produce 5 mm beam of slow positrons having a current of about 1 E + 12 sec -1 . The intensity of such a beam will be a least 3 orders of magnitude greater than those presently available. The construction of the ANS is not anticipated to be complete until the year 2000. In order to properly plan the design of the ANS, strong considerations are being given to a proof-of-principle experiment, using the presently available High Flux Isotopes Reactor, to test the 78 Kr loop technique. The positron current from the HFIR facility is expected to be about 1 E + 10 sec -1 , which is 2 orders of magnitude greater than any other available. If the experiment succeeds, a very valuable facility will be established, and important formation will be generated on how the ANS should be designed. 3 refs., 1 fig

Transmutation of technetium into stable ruthenium in high flux conceptual research reactor

International Nuclear Information System (INIS)

Amrani, N.; Boucenna, A.

2007-01-01

The effectiveness of transmutation for the long lived fission product technetium-99 in high flux research reactor, considering its large capture cross section in thermal and epithermal region is evaluated. The calculation of Ruthenium concentration evolution under irradiation was performed using Chain Solver 2.20 code. The approximation used for the transmutation calculation is the assumption that the influence of change in irradiated materials structures on the reactor operator mode characteristics is insignificant. The results on Technetium transmutation in high flux research reactor suggested an effective use of this kind of research reactors. The evaluation brings a new concept of multi-recycle Technetium transmutation using HFR T RAN (High Flux Research Reactor for Transmutation)

Calculation of radiation production of high specific activity isotopes 192Ir and 60Co

International Nuclear Information System (INIS)

Zhou Quan; Zhong Wenfa; Xu Xiaolin

1997-01-01

The high specific activity isotopes: 192 Ir and 60 Co in the high neutron flux reactor are calculated with the method of reactor physics. The results of calculation are analyzed in two aspects: the production of isotopes and the influence to parameters of the reactor, and hence a better case is proposed as a reference to the production

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

International Nuclear Information System (INIS)

Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F.; Vermeeren, L.; Lopez, A. Legrand

2011-01-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10 20 n/cm 2 . A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

Science.gov (United States)

Geslot, B.; Vermeeren, L.; Filliatre, P.; Lopez, A. Legrand; Barbot, L.; Jammes, C.; Bréaud, S.; Oriol, L.; Villard, J.-F.

2011-03-01

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 × 1020 n/cm2. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

New measurement system for on line in core high-energy neutron flux monitoring in materials testing reactor conditions

Energy Technology Data Exchange (ETDEWEB)

Geslot, B.; Filliatre, P.; Barbot, L.; Jammes, C.; Breaud, S.; Oriol, L.; Villard, J.-F. [CEA, DEN, Cadarache, SPEx/LDCI, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium); Lopez, A. Legrand [CEA, DEN, Saclay, SIREN/LECSI, F-91400 Saclay (France)

2011-03-15

Flux monitoring is of great interest for experimental studies in material testing reactors. Nowadays, only the thermal neutron flux can be monitored on line, e.g., using fission chambers or self-powered neutron detectors. In the framework of the Joint Instrumentation Laboratory between SCK-CEN and CEA, we have developed a fast neutron detector system (FNDS) capable of measuring on line the local high-energy neutron flux in fission reactor core and reflector locations. FNDS is based on fission chambers measurements in Campbelling mode. The system consists of two detectors, one detector being mainly sensitive to fast neutrons and the other one to thermal neutrons. On line data processing uses the CEA depletion code DARWIN in order to disentangle fast and thermal neutrons components, taking into account the isotopic evolution of the fissile deposit. The first results of FNDS experimental test in the BR2 reactor are presented in this paper. Several fission chambers have been irradiated up to a fluence of about 7 x 10{sup 20} n/cm{sup 2}. A good agreement (less than 10% discrepancy) was observed between FNDS fast flux estimation and reference flux measurement.

Optimization of neutron flux distribution in Isotope Production Reactor

International Nuclear Information System (INIS)

Valladares, G.L.

1988-01-01

In order to optimize the thermal neutrons flux distribution in a Radioisotope Production and Research Reactor, the influence of two reactor parameters was studied, namely the Vmod / Vcomb ratio and the core volume. The reactor core is built with uranium oxide pellets (UO 2 ) mounted in rod clusters, with an enrichment level of ∼3 %, similar to LIGHT WATER POWER REATOR (LWR) fuel elements. (author) [pt

Dissolution Flowsheet for High Flux Isotope Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

Daniel, W. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Karay, N. S [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-27

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U₃O₈) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H₂. The HFIR fuel cores will be dissolved and the recovered U will be down-blended into low-enriched U. HFIR fuel was previously processed in H-Canyon using a unique insert in both the 6.1D and 6.4D dissolvers. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The objective of this study was to identify flowsheet conditions through literature review and laboratory experimentation to safely and efficiently dissolve the HFIR fuel in H-Canyon. Laboratory-scale experiments were performed to evaluate the dissolution of HFIR fuel using both Al 1100 and Al 6061 T6 alloy coupons. The Al 1100 alloy was considered a representative surrogate which provided an upper bound on the generation of flammable (i.e., H₂) gas during the dissolution process. The dissolution of the Al 6061 T6 alloy proceeded at a slower rate than the Al 1100 alloy, and was used to verify that the target Al concentration in solution could be achieved for the selected Hg concentration. Mass spectrometry and Raman spectroscopy were used to provide continuous monitoring of the concentration of H₂ and other permanent gases in the dissolution offgas, allowing the development of H₂ generation rate profiles. The H₂ generation rates were subsequently used to evaluate if a full HFIR core could be dissolved in an H-Canyon dissolver without exceeding 60% of the

The High Flux Beam Reactor at Brookhaven National Laboratory

International Nuclear Information System (INIS)

Shapiro, S.M.

1994-01-01

Brookhaven National Laboratory's High Flux Beam Reactor (HFBR) was built because of the need of the scientist to always want 'more'. In the mid-50's the Brookhaven Graphite reactor was churning away producing a number of new results when the current generation of scientists, led by Donald Hughes, realized the need for a high flux reactor and started down the political, scientific and engineering path that led to the BFBR. The effort was joined by a number of engineers and scientists among them, Chemick, Hastings, Kouts, and Hendrie, who came up with the novel design of the HFBR. The two innovative features that have been incorporated in nearly all other research reactors built since are: (i) an under moderated core arrangement which enables the thermal flux to peak outside the core region where beam tubes can be placed, and (ii) beam tubes that are tangential to the core which decrease the fast neutron background without affecting the thermal beam intensity. Construction began in the fall of 1961 and four years later, at a cost of $12 Million, criticality was achieved on Halloween Night, 1965. Thus began 30 years of scientific accomplishments

High Flux Materials Testing Reactor (HFR), Petten

International Nuclear Information System (INIS)

1975-09-01

After conversion to burnable poison fuel elements, the High Flux Materials Testing Reactor (HFR) Petten (Netherlands), operated through 1974 for 280 days at 45 MW. Equipment for irradiation experiments has been replaced and extended. The average annual occupation by experiments was 55% as compared to 38% in 1973. Work continued on thirty irradiation projects and ten development activities

Annual report 1990. Operation of the high flux reactor

International Nuclear Information System (INIS)

Ahlf, J.; Gevers, A.

1990-01-01

In 1990 the operation of the High Flux Reactor was carried out as planned. The availability was 96% of scheduled operating time. The average utilization of the reactor was 71% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

Annual report 1989 operation of the high flux reactor

International Nuclear Information System (INIS)

Ahlf, J.; Gevers, A.

1989-01-01

In 1989 the operation of the High Flux Reactor Petten was carried out as planned. The availability was more than 100% of scheduled operating time. The average occupation of the reactor by experimental devices was 72% of the practical occupation limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons and for radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

Annual Report 1991. Operation of the high flux reactor

International Nuclear Information System (INIS)

Ahlf, J.; Gevers, A.

1992-01-01

In 1991 the operation of the High Flux Reactor was carried out as planned. The availability was more than 100% of scheduled operating time. The average utilization of the reactor was 69% of the practical limit. The reactor was utilized for research programmes in support of nuclear fission reactors and thermonuclear fusion, for fundamental research with neutrons, for radioisotope production, and for various smaller activities. Development activities addressed upgrading of irradiation devices, neutron capture therapy, neutron radiography and neutron transmutation doping of silicon. General activities in support of running irradiation programmes progressed in the normal way

Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

Energy Technology Data Exchange (ETDEWEB)

Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-10-01

An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

Very high flux steady state reactor and accelerator based sources

International Nuclear Information System (INIS)

Ludewig, H.; Todosow, M.; Simos, N.; Shapiro, S.; Hastings, J.

2004-01-01

With the number of steady state neutron sources in the US declining (including the demise of the Bnl HFBR) the remaining intense sources are now in Europe (i.e. reactors - ILL and FMR, accelerator - PSI). The intensity of the undisturbed thermal flux for sources currently in operation ranges from 10 14 n/cm 2 *s to 10 15 n/cm 2 *s. The proposed Advanced Neutron Source (ANS) was to be a high power reactor (about 350 MW) with a projected undisturbed thermal flux of 7*10 15 n/cm 2 *s but never materialized. The objective of the current study is to explore the requirements and implications of two source concepts with an undisturbed flux of 10 16 n/cm 2 *s. The first is a reactor based concept operating at high power density (10 MW/l - 15 MW/l) and a total power of 100 MW - 250 MW, depending on fissile enrichment. The second is an accelerator based concept relying on a 1 GeV - 1.5 GeV proton Linac with a total beam power of 40 MW and a liquid lead-bismuth eutectic target. In the reactor source study, the effects of fissile material enrichment, coolant temperature and pressure drop, and estimates of pressure vessel stress levels will be investigated. The fuel form for the reactor will be different from all other operating source reactors in that it is proposed to use an infiltrated graphitic structure, which has been developed for nuclear thermal propulsion reactor applications. In the accelerator based source the generation of spallation products and their activation levels, and the material damage sustained by the beam window will be investigated. (authors)

Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

International Nuclear Information System (INIS)

Bryant, Rebecca; Kszos, Lynn A.

2011-01-01

Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews

Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

2011-03-01

Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one

High flux-fluence measurements in fast reactors

International Nuclear Information System (INIS)

Lippincott, E.P.; Ulseth, J.A.

1977-01-01

Characterization of irradiation environments for fuels and materials tests in fast reactors requires determination of the neutron flux integrated over times as long as several years. An accurate integration requires, therefore, passive dosimetry monitors with long half-life or stable products which can be conveniently measured. In addition, burn-up, burn-in, and burn-out effects must be considered in high flux situations and use of minimum quantities of dosimeter materials is often desirable. These conditions force the use of dosimeter and dosimeter container designs, measured products, and techniques that are different from those that are used in critical facilities and other well-characterized benchmark fields. Recent measurements in EBR-II indicate that high-accuracy results can be attained and that tie-backs to benchmark field technique calibrations can be accomplished

Heat transfer for ultrahigh flux reactor

International Nuclear Information System (INIS)

Wadkins, R.P.; Lake, J.A.; Oh, C.H.

1987-01-01

The use of a uniquely designed nuclear reactor to supply neutrons for materials research is the focus of recent reactor design efforts. The biological, materials, and fundamental physics aspects of research require neutron fluxes much higher than present research and testing facilities can produce. The most advanced research using neutrons as probing detectors is being done in the High Flux Reactor at the Institut Laue Langeuin, France. The design of a reactor that can produce neutron fluxes of 1.0 x 10 16 n/cm 2 .s requires a relatively high power (300 MW range) and a small core volume (approximately 30 liters). This combination of power and volume leads to a high power density which places increased demands on thermal hydraulic margins

Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Sease, J.D.; Primm, R.T. III; Miller, J.H.

2007-01-01

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the

Conceptual Process for the Manufacture of Low-Enriched Uranium/Molybdenum Fuel for the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Sease, J.D.; Primm, R.T. III; Miller, J.H.

2007-09-30

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are

Reactor vessel dismantling at the high flux materials testing reactor Petten

International Nuclear Information System (INIS)

Tas, A.; Teunissen, G.

1986-01-01

The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

A conceptual high flux reactor design with scope for use in ADS ...

Indian Academy of Sciences (India)

By design the flux level in the seed fuel has been kept lower than in the high flux trap zones so that the burning rate of the seed is reduced. Another important objective of the design is to maximize the time interval of refueling. As against a typical refueling interval of a few weeks in such high flux reactor cores, it is desired to ...

Determination of integrated neutron flux by the measurement of the isotopic ratios of cadmium and gadolinium

International Nuclear Information System (INIS)

Tomiyoshi, Irene Akemy

1982-01-01

In this work, the possibility of the indirect determination of the integrated neutron flux, through the change of isotopic ratios of cadmium and gadolinium was investigated. The samples of cadmium we/e gadolinium were irradiated in the IEA-Rl reactor. These elements were chosen because they have high thermal neutron absorption cross section which permit the change in the isotopic composition during a short irradiation time to be measured accurately. The isotopic ratios were measured with a thermionic mass spectrometer the silica-gel technique and arrangement with single filament were used for the cadmium analysis, where as the oxi - reduction technique and arrangement with double filaments were used for gadolinium analysis. The mass fractionation effects for cadmium and gadolinium were corrected respectively by the exponential and potential expansion of the isotopic fractionation factor per atomic mass unit. The flux values supplied by the Centro de Operacao e Utilizacao do Reator de Pesquisas do IPEN were extrapolated. These values and the integrated flux values obtained experimentally were compared. (author)

Fuel management at the Petten high flux reactor

International Nuclear Information System (INIS)

Thijssen, P.J.M.

1999-01-01

Several years ago the shipment of spent fuel of the High Flux Reactor (HFR) at Petten has come to a standstill resulting in an ever growing stock of fuel elements that are labelled 'fully burnt up'. Examination of those elements showed that a reasonably number of them have a relatively high 235 U mass left. A reactor physics analysis showed that the use of such elements in the peripheral core zone allows the loading of four instead of five fresh fuel elements in many cycle cores. For the assessment of safety and performance parameters of HFR cores a new calculational tool is being developed. It is based on AEA Technology's Reactor physics code suite Winfrith Improved Multigroup Scheme (WIMS). NRG produced pre- and post-processing facilities to feed input data into WIMS's 2D transport code CACTUS and to extract relevant parameters from the output. The processing facilities can be used for many different types of application. (author)

Dosimetry issues for an ultra-high flux beam and multipurpose research reactor design

International Nuclear Information System (INIS)

West, C.D.

1993-01-01

The Advanced Neutron Source is a new user facility for all fields of neutron research, including neutron beam experiments, materials analysis, materials testing, and isotope production. The complement and layout of the experimental facilities have been determined sufficiently, at a conceptual design level, to make reliable cost and schedule estimates. The source of neutrons will be a heavy water reactor, constructed largely of aluminum, with an available thermal neutron flux 5--10 times higher than existing research reactors. Among the dosimetry issues to be faced are damage prediction and surveillance for component life attainment; measurement of fluence and spectra in regions where both change substantially over a distance of a few centimeters; and characterization and measurement of the radiation field in the research areas around the neutron beam experiments

KüFA safety testing of HTR fuel pebbles irradiated in the High Flux Reactor in Petten

Energy Technology Data Exchange (ETDEWEB)

Seeger, O., E-mail: oliver.seeger@rwth-aachen.de [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Laurie, M., E-mail: mathias.laurie@ec.europa.eu [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Abjani, A. El; Ejton, J.; Boudaud, D.; Freis, D.; Carbol, P.; Rondinella, V.V. [European Commission, Joint Research Centre (JRC), Institute for Transuranium Elements (ITU), Safety of Irradiated Nuclear Materials Unit, Postfach 2340, 76125 Karlsruhe (Germany); Fütterer, M. [European Commission, Joint Research Centre (JRC), Institute for Energy and Transport (IET), Nuclear Reactor Integrity Assessment and Knowledge Management Unit, PO Box 2, 1755 ZG Petten (Netherlands); Allelein, H.-J. [Lehrstuhl für Reaktorsicherheit und -technik an der RWTH Aachen, Kackertstraße 9, 52072 Aachen (Germany)

2016-09-15

The Cold Finger Apparatus (KühlFinger-Apparatur—KüFA) in operation at JRC-ITU is designed to experimentally scrutinize the effects of Depressurization LOss of Forced Circulation (D-LOFC) accident scenarios on irradiated High Temperature Reactor (HTR) fuel pebbles. Up to 1600 °C, the reference maximum temperature for these accidents, high-quality German HTR fuel pebbles have already demonstrated a small fission product release. This paper discusses and compares the releases obtained from KüFA-testing the pebbles HFR-K5/3 and HFR-EU1/3, which were both irradiated in the High Flux Reactor (HFR) in Petten. We present the time-dependent fractional release of the volatile fission product {sup 137}Cs as well as the fission gas {sup 85}Kr for both pebbles. For HFR-EU1/3 the isotopes {sup 134}Cs and {sup 154}Eu as well as the shorter-lived {sup 110m}Ag have also been measured. A detailed description of the experimental setup and its accuracy is given. The data for the recently tested pebbles is discussed in the context of previous results.

Modernization of the High Flux Isotope Reactor (HFIR) to Provide a Cold Neutron Source and Experimentation Facility

International Nuclear Information System (INIS)

Rothrock, Benjamin G.; Farrar, Mike B.

2009-01-01

In June 1961, construction was started on the High Flux Isotope Reactor (HFIR) facility inside the Oak Ridge National Laboratory (ORNL), at the recommendation of the U.S. Atomic Energy Commission (AEC) Division of Research. Construction was completed in early 1965 with criticality achieved on August 25, 19651. From the first full power operating cycle beginning in September 1966, the HFIR has achieved an outstanding record of service to the scientific community. In early 1995, the ORNL deputy director formed a group to examine the need for upgrades to the HFIR following the cancellation of the Advanced Neutron Source Project by DOE. This group indicated that there was an immediate need for the installation of a cold neutron source facility in the HFIR to produce cold neutrons for neutron scattering research uses. Cold neutrons have long wavelengths in the range of 4-12 angstroms. Cold neutrons are ideal for research applications with long length-scale molecular structures such as polymers, nanophase materials, and biological samples. These materials require large scale examination (and therefore require a longer wavelength neutron). These materials represent particular areas of science are at the forefront of current research initiatives that have a potentially significant impact on the materials we use in our everyday lives and our knowledge of biology and medicine. This paper discusses the installation of a cold neutron source at HFIR with respect to the project as a modernization of the facility. The paper focuses on why the project was required, the scope of the cold source project with specific emphasis on the design, and project management information.

Setup for polarized neutron imaging using in situ 3He cells at the Oak Ridge National Laboratory High Flux Isotope Reactor CG-1D beamline.

Science.gov (United States)

Dhiman, I; Ziesche, Ralf; Wang, Tianhao; Bilheux, Hassina; Santodonato, Lou; Tong, X; Jiang, C Y; Manke, Ingo; Treimer, Wolfgang; Chatterji, Tapan; Kardjilov, Nikolay

2017-09-01

In the present study, we report a new setup for polarized neutron imaging at the ORNL High Flux Isotope Reactor CG-1D beamline using an in situ 3 He polarizer and analyzer. This development is very important for extending the capabilities of the imaging instrument at ORNL providing a polarized beam with a large field-of-view, which can be further used in combination with optical devices like Wolter optics, focusing guides, or other lenses for the development of microscope arrangement. Such a setup can be of advantage for the existing and future imaging beamlines at the pulsed neutron sources. The first proof-of-concept experiment is performed to study the ferromagnetic phase transition in the Fe 3 Pt sample. We also demonstrate that the polychromatic neutron beam in combination with in situ 3 He cells can be used as the initial step for the rapid measurement and qualitative analysis of radiographs.

1982 Annual status report: operation of the high flux reactor

International Nuclear Information System (INIS)

1983-01-01

The high flux materials testing reactor has been operated in 1982 within a few percent of the pre-set schedule, attaining 73% overall availability. Its utilization reached another record figure in 20 years: 81% without, 92% with, the low enrichment test elements irradiated during the year

Operation of the High Flux Reactor. Annual report 1985

International Nuclear Information System (INIS)

1985-01-01

This year was characterized by the end of a major rebuilding of the installation during which the reactor vessel and its peripheral components were replaced by new and redesigned equipment. Both operational safety and experimental use were largely improved by the replacement. The reactor went back to routine operation on February 14, 1985, and has been operating without problem since then. All performance parameters were met. Other upgrading actions started during the year concerned new heat exchangers and improvements to the reactor building complex. The experimental load of the High Flux Reactor reached a satisfactory level with an average of 57%. New developments aimed at future safety related irradiation tests and at novel applications of neutrons from the horizontal beam tubes. A unique remote encapsulation hot cell facility became available adding new possibilities for fast breeder fuel testing and for intermediate specimen examination. The HFR Programme hosted an international meeting on development and use of reduced enrichment fuel for research reactors. All aspects of core physics, manufacture technology, and licensing of novel, proliferation-free, research reactor fuel were debated

High Flux Isotope Reactor quarterly report, July--September 1975

International Nuclear Information System (INIS)

McCord, R.V.; Corbett, B.L.

1975-01-01

The replacement of the permanent beryllium reflector was completed this quarter. The reactor was shut down for 87 days for this maintenance operation. Erosion of the sealing surface at the stainless steel adaptor flange on the HB-1 beam tube facility was confirmed. A soft metallic O-ring was used to effect a seal when this facility was reassembled. A comprehensive inspection of the normally inaccessible parts of the reactor pressure vessel was made. No abnormalities were found

Evolution of the Reactor Antineutrino Flux and Spectrum at Daya Bay.

Science.gov (United States)

An, F P; Balantekin, A B; Band, H R; Bishai, M; Blyth, S; Cao, D; Cao, G F; Cao, J; Chan, Y L; Chang, J F; Chang, Y; Chen, H S; Chen, Q Y; Chen, S M; Chen, Y X; Chen, Y; Cheng, J; Cheng, Z K; Cherwinka, J J; Chu, M C; Chukanov, A; Cummings, J P; Ding, Y Y; Diwan, M V; Dolgareva, M; Dove, J; Dwyer, D A; Edwards, W R; Gill, R; Gonchar, M; Gong, G H; Gong, H; Grassi, M; Gu, W Q; Guo, L; Guo, X H; Guo, Y H; Guo, Z; Hackenburg, R W; Hans, S; He, M; Heeger, K M; Heng, Y K; Higuera, A; Hsiung, Y B; Hu, B Z; Hu, T; Huang, E C; Huang, H X; Huang, X T; Huang, Y B; Huber, P; Huo, W; Hussain, G; Jaffe, D E; Jen, K L; Ji, X P; Ji, X L; Jiao, J B; Johnson, R A; Jones, D; Kang, L; Kettell, S H; Khan, A; Kohn, S; Kramer, M; Kwan, K K; Kwok, M W; Langford, T J; Lau, K; Lebanowski, L; Lee, J; Lee, J H C; Lei, R T; Leitner, R; Leung, J K C; Li, C; Li, D J; Li, F; Li, G S; Li, Q J; Li, S; Li, S C; Li, W D; Li, X N; Li, X Q; Li, Y F; Li, Z B; Liang, H; Lin, C J; Lin, G L; Lin, S; Lin, S K; Lin, Y-C; Ling, J J; Link, J M; Littenberg, L; Littlejohn, B R; Liu, J L; Liu, J C; Loh, C W; Lu, C; Lu, H Q; Lu, J S; Luk, K B; Ma, X Y; Ma, X B; Ma, Y Q; Malyshkin, Y; Martinez Caicedo, D A; McDonald, K T; McKeown, R D; Mitchell, I; Nakajima, Y; Napolitano, J; Naumov, D; Naumova, E; Ngai, H Y; Ochoa-Ricoux, J P; Olshevskiy, A; Pan, H-R; Park, J; Patton, S; Pec, V; Peng, J C; Pinsky, L; Pun, C S J; Qi, F Z; Qi, M; Qian, X; Qiu, R M; Raper, N; Ren, J; Rosero, R; Roskovec, B; Ruan, X C; Steiner, H; Stoler, P; Sun, J L; Tang, W; Taychenachev, D; Treskov, K; Tsang, K V; Tull, C E; Viaux, N; Viren, B; Vorobel, V; Wang, C H; Wang, M; Wang, N Y; Wang, R G; Wang, W; Wang, X; Wang, Y F; Wang, Z; Wang, Z; Wang, Z M; Wei, H Y; Wen, L J; Whisnant, K; White, C G; Whitehead, L; Wise, T; Wong, H L H; Wong, S C F; Worcester, E; Wu, C-H; Wu, Q; Wu, W J; Xia, D M; Xia, J K; Xing, Z Z; Xu, J L; Xu, Y; Xue, T; Yang, C G; Yang, H; Yang, L; Yang, M S; Yang, M T; Yang, Y Z; Ye, M; Ye, Z; Yeh, M; Young, B L; Yu, Z Y; Zeng, S; Zhan, L; Zhang, C; Zhang, C C; Zhang, H H; Zhang, J W; Zhang, Q M; Zhang, R; Zhang, X T; Zhang, Y M; Zhang, Y X; Zhang, Y M; Zhang, Z J; Zhang, Z Y; Zhang, Z P; Zhao, J; Zhou, L; Zhuang, H L; Zou, J H

2017-06-23

The Daya Bay experiment has observed correlations between reactor core fuel evolution and changes in the reactor antineutrino flux and energy spectrum. Four antineutrino detectors in two experimental halls were used to identify 2.2 million inverse beta decays (IBDs) over 1230 days spanning multiple fuel cycles for each of six 2.9 GW_{th} reactor cores at the Daya Bay and Ling Ao nuclear power plants. Using detector data spanning effective ^{239}Pu fission fractions F_{239} from 0.25 to 0.35, Daya Bay measures an average IBD yield σ[over ¯]_{f} of (5.90±0.13)×10^{-43}  cm^{2}/fission and a fuel-dependent variation in the IBD yield, dσ_{f}/dF_{239}, of (-1.86±0.18)×10^{-43}  cm^{2}/fission. This observation rejects the hypothesis of a constant antineutrino flux as a function of the ^{239}Pu fission fraction at 10 standard deviations. The variation in IBD yield is found to be energy dependent, rejecting the hypothesis of a constant antineutrino energy spectrum at 5.1 standard deviations. While measurements of the evolution in the IBD spectrum show general agreement with predictions from recent reactor models, the measured evolution in total IBD yield disagrees with recent predictions at 3.1σ. This discrepancy indicates that an overall deficit in the measured flux with respect to predictions does not result from equal fractional deficits from the primary fission isotopes ^{235}U, ^{239}Pu, ^{238}U, and ^{241}Pu. Based on measured IBD yield variations, yields of (6.17±0.17) and (4.27±0.26)×10^{-43}  cm^{2}/fission have been determined for the two dominant fission parent isotopes ^{235}U and ^{239}Pu. A 7.8% discrepancy between the observed and predicted ^{235}U yields suggests that this isotope may be the primary contributor to the reactor antineutrino anomaly.

A neutronic feasibility study for LEU conversion of the High Flux Beam Reactor (HFBR)

International Nuclear Information System (INIS)

Pond, R.B.; Hanan, N.A.; Matos, J.E.

1997-01-01

A neutronic feasibility study for converting the High Flux Beam Reactor at Brookhaven National Laboratory from HEU to LEU fuel was performed at Argonne National Laboratory. The purpose of this study is to determine what LEU fuel density would be needed to provide fuel lifetime and neutron flux performance similar to the current HEU fuel. The results indicate that it is not possible to convert the HFBR to LEU fuel with the current reactor core configuration. To use LEU fuel, either the core needs to be reconfigured to increase the neutron thermalization or a new LEU reactor design needs to be considered. This paper presents results of reactor calculations for a reference 28-assembly HEU-fuel core configuration and for an alternative 18-assembly LEU-fuel core configuration with increased neutron thermalization. Neutronic studies show that similar in-core and ex-core neutron fluxes, and fuel cycle length can be achieved using high-density LEU fuel with about 6.1 gU/cm 3 in an altered reactor core configuration. However, hydraulic and safety analyses of the altered HFBR core configuration needs to be performed in order to establish the feasibility of this concept. (author)

Presentation of the High-Flux Reactor of the Institut Laue-Langevin

International Nuclear Information System (INIS)

Guyon, H.

2006-01-01

Full text of publication follows: The High-Flux Reactor (HFR) of the Institut Laue-Langevin is the world's most intense source of neutrons for fundamental research. Thanks to its extremely compact core, which is made up of a single fuel element, the HFR is capable of producing a neutron flux of up to 1.5.10 15 n.cm -2 .s -1 with a moderate power output of 58 MW. Its heavy water reflector thermalizes these neutrons, giving them a wave length of the order of one angstrom. They then become an excellent tool for exploring the atomic structure of matter. In order to provide a full neutron spectrum, the reactor is also equipped with a hot source (a block of graphite heated to 2000 deg. C) and two cold sources (a volume of liquid deuterium at 25 K). All the reactor's components can be replaced and adapted in order to keep pace with both changing scientific needs and changing safety requirements. For example, in 1992 the reactor block was replaced, a second cold source was installed in 1985, and the beam tubes are replaced at regularly intervals and are also occasionally modified. In the same way, the reactor's civil engineering structures are currently being reinforced in order to comply with the reassessment of the reference earthquake spectra. Finally, the Institut Laue-Langevin's reactor is equipped with three solid containment barriers: - the fuel cladding: during the 35 years the reactor has been in operation, a cladding failure has never been detected; - the leak-tight primary cooling system: this is partly submerged in a pool which provides radiological shielding; - the double-wall containment: an overpressure of air is maintained between the inner reinforced concrete wall and the outer metal wall. The High-Flux Reactor is therefore all set to provide the scientific community with top quality service for the next 20 years to come, on a site which: - is home to the brightest synchrotron in the world (ESRF); - benefits from the microbiology expertise of the EMBL

Presentation of the High-Flux Reactor of the Institut Laue-Langevin

Energy Technology Data Exchange (ETDEWEB)

Guyon, H. [Institut Laue-Langevin, Grenoble (France)

2006-07-01

Full text of publication follows: The High-Flux Reactor (HFR) of the Institut Laue-Langevin is the world's most intense source of neutrons for fundamental research. Thanks to its extremely compact core, which is made up of a single fuel element, the HFR is capable of producing a neutron flux of up to 1.5.10{sup 15} n.cm{sup -2}.s{sup -1} with a moderate power output of 58 MW. Its heavy water reflector thermalizes these neutrons, giving them a wave length of the order of one angstrom. They then become an excellent tool for exploring the atomic structure of matter. In order to provide a full neutron spectrum, the reactor is also equipped with a hot source (a block of graphite heated to 2000 deg. C) and two cold sources (a volume of liquid deuterium at 25 K). All the reactor's components can be replaced and adapted in order to keep pace with both changing scientific needs and changing safety requirements. For example, in 1992 the reactor block was replaced, a second cold source was installed in 1985, and the beam tubes are replaced at regularly intervals and are also occasionally modified. In the same way, the reactor's civil engineering structures are currently being reinforced in order to comply with the reassessment of the reference earthquake spectra. Finally, the Institut Laue-Langevin's reactor is equipped with three solid containment barriers: - the fuel cladding: during the 35 years the reactor has been in operation, a cladding failure has never been detected; - the leak-tight primary cooling system: this is partly submerged in a pool which provides radiological shielding; - the double-wall containment: an overpressure of air is maintained between the inner reinforced concrete wall and the outer metal wall. The High-Flux Reactor is therefore all set to provide the scientific community with top quality service for the next 20 years to come, on a site which: - is home to the brightest synchrotron in the world (ESRF); - benefits from the

HFBR handbook, 1992: High flux beam reactor

International Nuclear Information System (INIS)

Axe, J.D.; Greenberg, R.

1992-10-01

Welcome to the High Flux Beam Reactor (HFBR), one of the world premier neutron research facilities. This manual is intended primarily to acquaint outside users (and new Brookhaven staff members) with (almost) everything they need to know to work at the HFBR and to help make the stay at Brookhaven pleasant as well as profitable. Safety Training Programs to comply with US Department of Energy (DOE) mandates are in progress at BNL. There are several safety training requirements which must be met before users can obtain unescorted access to the HFBR. The Reactor Division has prepared specific safety training manuals which are to be sent to experimenters well in advance of their expected arrival at BNL to conduct experiments. Please familiarize yourself with this material and carefully pay strict attention to all the safety and security procedures that are in force at the HFBR. Not only your safety, but the continued operation of the facility, depends upon compliance

Development of High Flux Isotope Reactor (HFIR) subcriticality monitoring methods

International Nuclear Information System (INIS)

Rothrock, R.B.

1991-01-01

Use of subcritical source multiplication measurements during refueling has been investigated as a possible replacement for out-of-reactor subcriticality measurements formerly made on fresh HFIR fuel elements at the ORNL Critical Experiment Facility. These measurements have been used in the past for preparation of estimated critical rod positions, and as a partial verification, prior to reactor startup, that the requirements for operational shutdown margin would be met. Results of subcritical count rate data collection during recent HFIR refuelings and supporting calculations are described illustrating the intended measurement method and its expected uncertainty. These results are compared to historical uses of the out-of-reactor core measurements and their accuracy requirements, and a planned in-reactor test is described which will establish the sensitivity of the method and calibrate it for future routine use during HFIR refueling. 2 refs., 1 fig., 2 tabs

COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

Directory of Open Access Journals (Sweden)

Phani Kumar Domalapally

2015-04-01

Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

Flux effect on neutron irradiation embrittlement of reactor pressure vessel steels irradiated to high fluences

International Nuclear Information System (INIS)

Soneda, N.; Dohi, K.; Nishida, K.; Nomoto, A.; Iwasaki, M.; Tsuno, S.; Akiyama, T.; Watanabe, S.; Ohta, T.

2011-01-01

Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is of great concern for the long term operation of light water reactors. In particular, the embrittlement of the RPV steels of pressurized water reactors (PWRs) at very high fluences beyond 6*10 19 n/cm 2 , E > 1 MeV, needs to be understood in more depth because materials irradiated in material test reactors (MTRs) to such high fluences show larger shifts than predicted by current embrittlement correlation equations available worldwide. The primary difference between the irradiation conditions of MTRs and surveillance capsules is the neutron flux. The neutron flux of MTR is typically more than one order of magnitude higher than that of surveillance capsule, but it is not necessarily clear if this difference in neutron flux causes difference in mechanical properties of RPV. In this paper, we perform direct comparison, in terms of mechanical property and microstructure, between the materials irradiated in surveillance capsules and MTRs to clarify the effect of flux at very high fluences and fluxes. We irradiate the archive materials of some of the commercial reactors in Japan in the MTR, LVR-15, of NRI Rez, Czech Republic. Charpy impact test results of the MTR-irradiated materials are compared with the data from surveillance tests. The comparison of the results of microstructural analyses by means of atom probe tomography is also described to demonstrate the similarity / differences in surveillance and MTR-irradiated materials in terms of solute atom behavior. It appears that high Cu material irradiated in a MTR presents larger shifts than those of surveillance data, while low Cu materials present similar embrittlement. The microstructural changes caused by MTR irradiation and surveillance irradiation are clearly different

High-flux first-wall design for a small reversed-field pinch reactor

International Nuclear Information System (INIS)

Cort, G.E.; Graham, A.L.; Christensen, K.E.

1982-01-01

To achieve the goal of a commercially economical fusion power reactor, small physical size and high power density should be combined with simplicity (minimized use of high-technology systems). The Reversed-Field Pinch (RFP) is a magnetic confinement device that promises to meet these requirements with power densities comparable to those in existing fission power plants. To establish feasibility of such an RFP reactor, a practical design for a first wall capable of withstanding high levels of cyclic neutron wall loadings is needed. Associated with the neutron flux in the proposed RFP reactor is a time-averaged heat flux of 4.5 MW/m 2 with a conservatively estimated transient peak approximately twice the average value. We present the design for a modular first wall made from a high-strength copper alloy that will meet these requirements of cyclic thermal loading. The heat removal from the wall is by subcooled water flowing in straight tubes at high linear velocities. We combined a thermal analysis with a structural fatigue analysis to design the heat transfer module to last 10 6 cycles or one year at 80% duty for a 26-s power cycle. This fatigue life is compatible with a radiation damage life of 14 MW/yr/m 2

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

Radiation dosimetry at the BNL High Flux Beam Reactor

International Nuclear Information System (INIS)

Holden, N.E.; Hu, J.P.; Reciniello, R.N.

1998-02-01

The HFBR is a heavy water, D 2 O, cooled and moderated reactor with twenty-eight fuel elements containing a maximum of 9.8 kilograms of 235 U. The core is 53 cm high and 48 cm in diameter and has an active volume of 97 liters. The HFBR, which was designed to operate at forty mega-watts, 40 NW, was upgraded to operate at 60 NW. Since 1991, it has operated at 30 MW. In a normal 30 MW operating cycle the HFBR operates 24 hours a day for thirty days, with a six to fourteen day shutdown period for refueling and maintenance work. While most reactors attempts to minimize the escape of neutrons from the core, the HFBR's D 2 O design allows the thermal neutron flux to peak in the reflector region and maximizes the number of thermal neutrons available to nine horizontal external beams, H-1 to H-9. The HFBR neutron dosimetry effort described here compares measured and calculated energy dependent neutron and gamma ray flux densities and/or dose rates at horizontal beam lines and vertical irradiation thimbles

The operating experience and incident analysis for High Flux Engineering Test Reactor

International Nuclear Information System (INIS)

Zhao Guang

1999-01-01

The paper describes the incidents analysis for High Flux Engineering test reactor (HFETR) and introduces operating experience. Some suggestion have been made to reduce the incidents of HFETR. It is necessary to adopt new improvements which enhance the safety and reliability of operation. (author)

Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

Directory of Open Access Journals (Sweden)

Noble Brooklyn

2012-01-01

Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Hurt, Christopher J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hobbs, Randy W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-08-01

There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 (²³⁸Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 (²³⁷Np) dioxide (NpO₂)/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs

Decommissioning of the High Flux Beam Reactor at Brookhaven Lab

Energy Technology Data Exchange (ETDEWEB)

Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2011-05-27

The High Flux Beam Reactor at the Brookhaven National Laboratory was a heavy water cooled and moderated reactor that achieved criticality on October 31, 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shutdown in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor’s spent fuel pool. The reactor remained shutdown for almost three years for safety and environmental reviews. In November 1999 the United States Department of Energy decided to permanently shutdown the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR cleanup conducted during 1999-2009 will be described in the paper.

The High Flux Reactor Petten, present status and prospects

Energy Technology Data Exchange (ETDEWEB)

Ahlf, J [Institute for Advanced Materials, Joint Research Centre, Petten (Netherlands)

1990-05-01

The High Flux Reactor (HFR) in Petten, The Netherlands, is a light water cooled and moderated multipurpose research reactor of the closed-tank in pool type. It is operated with highly enriched Uranium fuel at a power of 45 MW. The reactor is owned by the European Communities and operated under contract by the Dutch ECN. The HFR programme is funded by The Netherlands and Germany, a smaller share comes from the specific programmes of the Joint Research Centre (JRC) and from third party contract work. Since its first criticality in 1961 the reactor has been continuously upgraded by implementing developments in fuel element technology and increasing the power from 20 MW to the present 45 MV. In 1984 the reactor vessel was replaced by a new one with an improved accessibility for experiments. In the following years also other ageing equipment has been replaced (primary heat exchangers, pool heat exchanger, beryllium reflector elements, nuclear and process instrumentation, uninterruptable power supply). Control room upgrading is under preparation. A new safety analysis is near to completion and will form the basis for a renewed license. The reactor is used for nuclear energy related research (structural materials and fuel irradiations for LWR's, HTR's and FBR's, fusion materials irradiations). The beam tubes are used for nuclear physics as well as solid state and materials sciences. Radioisotope production at large scale, processing of gemstones and silicon with neutrons, neutron radiography and activation analysis are actively pursued. A clinical facility for boron neutron capture therapy is being designed at one of the large cross section beam tubes. It is foreseen to operate the reactor at least for a further decade. The exploitation pattern may undergo some changes depending on the requirements of the supporting countries and the JRC programmes. (author)

The High Flux Reactor Petten, present status and prospects

International Nuclear Information System (INIS)

Ahlf, J.

1990-01-01

The High Flux Reactor (HFR) in Petten, The Netherlands, is a light water cooled and moderated multipurpose research reactor of the closed-tank in pool type. It is operated with highly enriched Uranium fuel at a power of 45 MW. The reactor is owned by the European Communities and operated under contract by the Dutch ECN. The HFR programme is funded by The Netherlands and Germany, a smaller share comes from the specific programmes of the Joint Research Centre (JRC) and from third party contract work. Since its first criticality in 1961 the reactor has been continuously upgraded by implementing developments in fuel element technology and increasing the power from 20 MW to the present 45 MV. In 1984 the reactor vessel was replaced by a new one with an improved accessibility for experiments. In the following years also other ageing equipment has been replaced (primary heat exchangers, pool heat exchanger, beryllium reflector elements, nuclear and process instrumentation, uninterruptable power supply). Control room upgrading is under preparation. A new safety analysis is near to completion and will form the basis for a renewed license. The reactor is used for nuclear energy related research (structural materials and fuel irradiations for LWR's, HTR's and FBR's, fusion materials irradiations). The beam tubes are used for nuclear physics as well as solid state and materials sciences. Radioisotope production at large scale, processing of gemstones and silicon with neutrons, neutron radiography and activation analysis are actively pursued. A clinical facility for boron neutron capture therapy is being designed at one of the large cross section beam tubes. It is foreseen to operate the reactor at least for a further decade. The exploitation pattern may undergo some changes depending on the requirements of the supporting countries and the JRC programmes. (author)

Annual progress report 1988, operation of the high flux reactor

International Nuclear Information System (INIS)

1989-01-01

In 1988 the High Flux Reactor Petten was routinely operated without any unforeseen event. The availability was 99% of scheduled operation. Utilization of the irradiation positions amounted to 80% of the practical occupation limit. The exploitation pattern comprised nuclear energy deployment, fundamental research with neutrons, and radioisotope production. General activities in support of running irradiation programmes progressed in the normal way. Development activities addressed upgrading of irradiation devices, neutron radiography and neutron capture therapy

Research reactors - an overview

International Nuclear Information System (INIS)

West, C.D.

1997-01-01

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

Production of Sn-117m in the BR2 high-flux reactor.

Science.gov (United States)

Ponsard, B; Srivastava, S C; Mausner, L F; Russ Knapp, F F; Garland, M A; Mirzadeh, S

2009-01-01

The BR2 reactor is a 100MW(th) high-flux 'materials testing reactor', which produces a wide range of radioisotopes for various applications in nuclear medicine and industry. Tin-117m ((117m)Sn), a promising radionuclide for therapeutic applications, and its production have been validated in the BR2 reactor. In contrast to therapeutic beta emitters, (117m)Sn decays via isomeric transition with the emission of monoenergetic conversion electrons which are effective for metastatic bone pain palliation and radiosynovectomy with lesser damage to the bone marrow and the healthy tissues. Furthermore, the emitted gamma photons are ideal for imaging and dosimetry.

Surveillance programme and upgrading of the High Flux Reactor Petten

International Nuclear Information System (INIS)

Bieth, Michel

1995-01-01

The High Flux Reactor (HFR) at Petten (The Netherlands), a 45 MW light water cooled and moderated research reactor in operation during more than 30 years, has been kept up to date by replacing ageing components. In 1984, the HFR was shut down for replacement of the aluminium. reactor vessel which had been irradiated during more than 20 years. The demonstration that the new vessel contains no critical defect requires knowledge of the material properties of the aluminium alloy Al 5154 with and without neutron irradiation and of the likely defect presence through the periodic in-service inspections. An irradiation damage surveillance programme has been started in 1985 for the new vessel material to provide information on fracture mechanics properties. After the vessel replacement, the existing process of continuous upgrading and replacement of ageing components was accelerated. A stepwise upgrade of the control room is presently under realization. (author)

Surveillance programme and upgrading of the High Flux Reactor Petten

Energy Technology Data Exchange (ETDEWEB)

Bieth, Michel [Commission of the European Communities, Joint Research Centre, Institute for Advanced Materials, High Flux Reactor Unit, Petten (Netherlands)

1995-07-01

The High Flux Reactor (HFR) at Petten (The Netherlands), a 45 MW light water cooled and moderated research reactor in operation during more than 30 years, has been kept up to date by replacing ageing components. In 1984, the HFR was shut down for replacement of the aluminium. reactor vessel which had been irradiated during more than 20 years. The demonstration that the new vessel contains no critical defect requires knowledge of the material properties of the aluminium alloy Al 5154 with and without neutron irradiation and of the likely defect presence through the periodic in-service inspections. An irradiation damage surveillance programme has been started in 1985 for the new vessel material to provide information on fracture mechanics properties. After the vessel replacement, the existing process of continuous upgrading and replacement of ageing components was accelerated. A stepwise upgrade of the control room is presently under realization. (author)

Upgrading and modernization of the high flux reactor Petten

International Nuclear Information System (INIS)

Ahlf, J.

1992-01-01

The High Flux Reactor (HFR) at Petten, The Netherlands, owned by the European Communities and operated by the Netherlands Energy Research Foundation, is a water-cooled and moderated, multipurpose research reactor of the closed-tank in-pool type, operated at 45 MW. Performance upgrading comprised two power increases from 20 MW via 30 MW to 45 MW, providing more and higher rated irradiation positions in the tank. With the replacement of the original reactor vessel the experimental capabilities of the reactor were improved. Better pool side facilities and the introduction of a large cross-section, double, beam tube were implemented. Additional major installation upgrading activities consisted of the replacement of the primary and the pool heat exchangers, replacement of the beryllium reflector elements, extension of the overpower protection systems and upgrading of the nuclear instrumentation as well as the guaranteed power supply. Control room upgrading is in progress. A full new safety analysis, as well as the introduction of a comprehensive Quality Assurance system, are summarized under software upgrading. Continuous modernization and upgrading also takes place of equipment for fuel and structural materials irradiations for fission reactors and future fusion machines. In parallel, all supporting services, as well as the management structure for large irradiation programmes, have been developed. Presently the reactor is operating at about 275 full power days per year with an average utilization of the irradiation positions of 70 to 80%. (orig.)

Seismic upgrading of the Brookhaven High Flux Beam Research Reactor

International Nuclear Information System (INIS)

Subudhi, M.

1985-01-01

In recent years the High Flux Beam Research (HFBR) reactor facility at Brookhaven National Laboratory (BNL) was upgraded from 40 to 50 MW power level. The reactor plant was built in the early sixties to the seismic design requirements of the period, using the static load approach. While the plant power level was upgraded, the seismic design was also improved according to current design criteria. This included the development of new floor response spectra for the facility and an overall seismic analysis of those systems important to the safe shutdown of the reactor. Items included in the reanalysis are the containment building with its internal structure, the piping systems, tanks, equipment, and heat exchangers. This paper describes the procedure utilized in developing the floor response spectra for the existing facility. Also included in the paper are the findings and recommendations, based on the seismic analysis, regarding the seismic adequacy of structural and mechanical systems vital to achieving the safe shutdown of the reactor. 11 references, 4 figures, 1 table

New temperature monitoring devices for high-temperature irradiation experiments in the high flux reactor Petten

Energy Technology Data Exchange (ETDEWEB)

Laurie, M.; Futterer, M. A.; Lapetite, J. M. [European Commission Joint Research Centre, Institute for Energy, P.O. Box 2, NL-1755 ZG Petten (Netherlands); Fourrez, S. [THERMOCOAX SAS, BP 26, Planquivon, 61438 Flers Cedex (France); Morice, R. [Laboratoire National de Metrologie et d' Essais, 1 rue Gaston Boissier, 75724 Paris (France)

2009-07-01

Within the European High Temperature Reactor Technology Network (HTR-TN) and related projects a number of HTR fuel irradiations are planned in the High Flux Reactor Petten (HFR), The Netherlands, with the objective to explore the potential of recently produced fuel for even higher temperature and burn-up. Irradiating fuel under defined conditions to extremely high burn-ups will provide a better understanding of fission product release and failure mechanisms if particle failure occurs. After an overview of the irradiation rigs used in the HFR, this paper sums up data collected from previous irradiation tests in terms of thermocouple data. Some research and development work for further improvement of thermocouples and other on-line instrumentation will be outlined. (authors)

Evaluation of the performance of high temperature conversion reactors for compound-specific oxygen stable isotope analysis.

Science.gov (United States)

Hitzfeld, Kristina L; Gehre, Matthias; Richnow, Hans-Hermann

2017-05-01

In this study conversion conditions for oxygen gas chromatography high temperature conversion (HTC) isotope ratio mass spectrometry (IRMS) are characterised using qualitative mass spectrometry (IonTrap). It is shown that physical and chemical properties of a given reactor design impact HTC and thus the ability to accurately measure oxygen isotope ratios. Commercially available and custom-built tube-in-tube reactors were used to elucidate (i) by-product formation (carbon dioxide, water, small organic molecules), (ii) 2nd sources of oxygen (leakage, metal oxides, ceramic material), and (iii) required reactor conditions (conditioning, reduction, stability). The suitability of the available HTC approach for compound-specific isotope analysis of oxygen in volatile organic molecules like methyl tert-butyl ether is assessed. Main problems impeding accurate analysis are non-quantitative HTC and significant carbon dioxide by-product formation. An evaluation strategy combining mass spectrometric analysis of HTC products and IRMS 18 O/ 16 O monitoring for future method development is proposed.

Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

International Nuclear Information System (INIS)

Wulff, W.

1990-01-01

A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs

Status in 1998 of the high flux reactor fuel cycle

International Nuclear Information System (INIS)

Guidez, J.; Gevers, A.; Wijtsma, F.J.; Thijssen, P.M.J.

1998-01-01

The High Flux Reactor located at Petten (The Netherlands), is owned by the European Commission and is operated under contract by ECN (Netherlands Energy Research Foundation). This plant is in operation since 1962 using HEU enriched at 90%. Conversion studies were conducted several years ago with the hypothesis of a global conversion of the entire core. The results of these studies have shown a costly operation with a dramatic decrease of the thermal flux which is necessary for the medical use of the plant (Molybdene 99 production). Some tests with low enriched elements were also conducted with several companies, several geometrical configurations and several enrichments. They are described in this paper. Explanations are also given on future possibilities for new fuel testing. (author)

Application of the successive linear programming technique to the optimum design of a high flux reactor using LEU fuel

International Nuclear Information System (INIS)

Mo, S.C.

1991-01-01

The successive linear programming technique is applied to obtain the optimum thermal flux in the reflector region of a high flux reactor using LEU fuel. The design variables are the reactor power, core radius and coolant channel thickness. The constraints are the cycle length, average heat flux and peak/average power density ratio. The characteristics of the optimum solutions with various constraints are discussed

Measurements of neutron flux in the RA reactor

International Nuclear Information System (INIS)

Raisic, N.

1961-12-01

This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire [sr

High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

International Nuclear Information System (INIS)

Chrysochoides, N.G.; Cundy, M.R.; Von der Hardt, P.; Husmann, K.; Swanenburg de Veye, R.J.; Tas, A.

1985-01-01

The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

Decommissioning of the High Flux Beam Reactor at Brookhaven National Laboratory.

Science.gov (United States)

Hu, Jih-Perng; Reciniello, Richard N; Holden, Norman E

2012-08-01

The High Flux Beam Reactor (HFBR) at the Brookhaven National Laboratory was a heavy-water cooled and moderated reactor that achieved criticality on 31 October 1965. It operated at a power level of 40 mega-watts. An equipment upgrade in 1982 allowed operations at 60 mega-watts. After a 1989 reactor shutdown to reanalyze safety impact of a hypothetical loss of coolant accident, the reactor was restarted in 1991 at 30 mega-watts. The HFBR was shut down in December 1996 for routine maintenance and refueling. At that time, a leak of tritiated water was identified by routine sampling of ground water from wells located adjacent to the reactor's spent fuel pool. The reactor remained shut down for almost 3 y for safety and environmental reviews. In November 1999, the United States Department of Energy decided to permanently shut down the HFBR. The decontamination and decommissioning of the HFBR complex, consisting of multiple structures and systems to operate and maintain the reactor, were complete in 2009 after removing and shipping off all the control rod blades. The emptied and cleaned HFBR dome, which still contains the irradiated reactor vessel is presently under 24/7 surveillance for safety. Details of the HFBR's cleanup performed during 1999-2009, to allow the BNL facilities to be re-accessed by the public, will be described in the paper.

The possible transmutation of radioactive waste from nuclear reactors

International Nuclear Information System (INIS)

Harries, J.R.

1974-01-01

A nuclear reactor power program produces high level and long lived radioactive wastes. The high level activity is associated with fission products, but beyond 400 years the principal waste hazard is from transuranic elements produced in the reactor. Several schemes have been proposed for the transmutation of the problem isotopes into more easily handled isotopes. The neutron flux in a thermal reactor is not high enough to significantly reduce the longer lived fission product isotopes 90 Sr and 132 Gs, but the transuranic elements can be reduced by recycling through power reactors. The limitation on recycling of the transuranic elements is the separation process to remove trace quantities from the waste stream. In fast reactors the transuranic elements are the principal fuel and fast reactor waste contains only half as much 90 Sr as thermal reactors. However, the overall waste hazard is similar to thermal reactors. A sufficiently intense neutron flux for fission product transmutation could perhaps be produced by a spallation reactor driven by a proton linear accelerator or a controlled thermonuclear reactor. However, both concepts are still some years in the future. Transmutation by accelerator sources of protons, electrons of gammas tend to require more energy than neutron transmutation. (author)

Radiation Dosimetry of the Pressure Vessel Internals of the High Flux Beam Reactor

Science.gov (United States)

Holden, Norman E.; Reciniello, Richard N.; Hu, Jih-Perng; Rorer, David C.

2003-06-01

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. The measurements were made using Red Perspex™ polymethyl methacrylate high-level film dosimeters, a Radcal "peanut" ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rates, the Monte Carlo MCNP code and geometric progressive MicroShield code were used to model the gamma-ray transport and dose buildup.

RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR

International Nuclear Information System (INIS)

HOLDEN, N.E.; RECINIELLO, R.N.; HU, J.P.; RORER, D.C.

2002-01-01

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex(trademark) polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup

Rebuilding the Brookhaven high flux beam reactor: A feasibility study

International Nuclear Information System (INIS)

Brynda, W.J.; Passell, L.; Rorer, D.C.

1995-01-01

After nearly thirty years of operation, Brookhaven's High Flux Beam Reactor (HFBR) is still one of the world's premier steady-state neutron sources. A major center for condensed matter studies, it currently supports fifteen separate beamlines conducting research in fields as diverse as crystallography, solid-state, nuclear and surface physics, polymer physics and structural biology and will very likely be able to do so for perhaps another decade. But beyond that point the HFBR will be running on borrowed time. Unless appropriate remedial action is taken, progressive radiation-induced embrittlement problems will eventually shut it down. Recognizing the HFBR's value as a national scientific resource, members of the Laboratory's scientific and reactor operations staffs began earlier this year to consider what could be done both to extend its useful life and to assure that it continues to provide state-of-the-art research facilities for the scientific community. This report summarizes the findings of that study. It addresses two basic issues: (i) identification and replacement of lifetime-limiting components and (ii) modifications and additions that could expand and enhance the reactor's research capabilities

Thermal neutron flux distribution in ET-RR-2 reactor thermal column

Directory of Open Access Journals (Sweden)

Imam Mahmoud M.

2002-01-01

Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

Numeric modeling of HfO2 neutron flux sensor parameters during sensor burnup in the RBMK-1500 reactor

International Nuclear Information System (INIS)

Jurkevicius, A.; Remeikis, V.

2001-01-01

The isotopic composition of hafnium in the radial neutron flux sensor of the RBMK-1500 reactor, the rates of the neutron absorption on Hf isotopes and the neutron spectrum in the sensor were numerically modeled. The sequence SAS2 (Shielding Analysis Sequence) from the package SCALE 4.3 was used for calculations. It has been obtained that the main neutron absorber 167 Er isotope practically burns up completely at the 18 MW d/kgU burnup depth, and at that time the capture rate of thermal neutrons in erbium decreases ten-fold. The average neutron flux density was calculated 7.6*10 13 neutrons. Cm -2 S -1 in the RBMK-1500 reactor grating, when the nuclear fuel enriched with 235 U by 2.4% and with Er by 0.4% is used in a fuel assembly. When the sensor burnup reaches 28 MW d/kgU, the neutron absorption rate of 178 Hf exceeds the rate of 177 Hf. The overall neutron absorption rate in hafnium decreases 2.53 times due to the sensor burnup to 56 MW d/kgU. The corrective factors ξ d (I) at different integral flux I of the sensors were calculated. The obtained dependence ξ d (I) calculated numerically was compared to the experimental one determined by processing repeated calibration results of Hf sensors in RBMK-1500 reactors, as well as compared to the theoretical one currently used in the Ignalina NPP special mathematical algorithms. (author)

Conditional CO2 flux analysis of a managed grassland with the aid of stable isotopes

Science.gov (United States)

Zeeman, M. J.; Tuzson, B.; Emmenegger, L.; Knohl, A.; Buchmann, N.; Eugster, W.

2009-04-01

Short statured managed ecosystems, such as agricultural grasslands, exhibit high temporal changes in carbon dioxide assimilation and respiration fluxes for which measurements of the net CO2 flux, e.g. by using the eddy covariance (EC) method, give only limited insight. We have therefore adopted a recently proposed concept for conditional EC flux analysis of forest to grasslands, in order to identify and quantify daytime sub-canopy respiration fluxes. To validate the concept, high frequency (≈5 Hz) stable carbon isotope analyis of CO2 was used. We made eddy covariance measurements of CO2 and its isotopologues during four days in August 2007, using a novel quantum cascade laser absorption spectrometer, capable of high time resolution stable isotope analysis. The effects of a grass cut during the measurement period could be detected and resulted in a sub-canopy source conditional flux classification, for which the isotope composition of the CO2 could be confirmed to be of a respiration source. However, the conditional flux method did not work for an undisturbed grassland canopy. We attribute this to the flux measurement height that was chosen well above the roughness sublayer, where the natural isotopic tracer (δ13C) of respiration was too well mixed with background air.

The feature of high flux engineering test reactor and its role in nuclear power development

International Nuclear Information System (INIS)

Lu Guangquan

1987-01-01

The High Flux Engineering Test Reactor (HFETR) designed and built by Chinese own efforts reached to its initial criticality on Dec. 27, 1979, and then achieved high power operation on Dec. 16, 1980. Until Nov. 11. 1986, the reactor had been operated for thirteen cycles. The paper presents briefly main feature of HFETR and its utilization during past years. The paper also deals with its role in nuclear power development. Finally, author gives his opinion on comprehensive utilization of HFETR. (author)

High flux Particle Bed Reactor systems for rapid transmutation of actinides and long lived fission products

International Nuclear Information System (INIS)

Powell, J.; Ludewig, H.; Maise, G.; Steinberg, M.; Todosow, M.

1993-01-01

An initial assessment of several actinide/LLFP burner concepts based on the Particle Bed Reactor (PBR) is described. The high power density/flux level achievable with the PBR make it an attractive candidate for this application. The PBR based actinide burner concept also possesses a number of safety and economic benefits relative to other reactor based transmutation approaches including a low inventory of radionuclides, and high integrity, coated fuel particles which can withstand extremely high in temperatures while retaining virtually all fission products. In addition the reactor also posesses a number of ''engineered safety features,'' which, along with the use of high temperature capable materials further enhance its safety characteristics

Monitoring the fast neutrons in a high flux: The case for 242Pu fission chambers

International Nuclear Information System (INIS)

Filliatre, P.; Jammes, C.; Oriol, L.; Geslot, B.; Vermeeren, L.

2009-01-01

Fission chambers are widely used for on-line monitoring of neutron fluxes in irradiation reactors. A selective measurement of a component of interest of the neutron flux is possible in principle thanks to a careful choice of the deposit material. However, measuring the fast component is challenging when the flux is high (up to 10 15 n/cm 2 /s) with a significant thermal component. The main problem is that the isotopic content of a material selected for its good response to fast neutrons evolves with irradiation, so that the material is more and more sensitive to thermal neutrons. Within the framework of the FNDS (Fast Neutron Detector System) project, we design tools that simulate the evolution of the isotopic composition and fission rate for several deposits under any given flux. In the case of a high flux with a significant thermal component, 242 Pu is shown after a comprehensive study of all possibilities to be the best choice for measuring the fast component, as long as its purity is sufficient. If an estimate of the thermal flux is independently available, one can correct the signal for that component. This suggests a system of two detectors, one of which being used for such a correction. It is of very high interest when the detectors must be operated up to a high neutron fluence. (authors)

EL-2 reactor: Thermal neutron flux distribution

International Nuclear Information System (INIS)

Rousseau, A.; Genthon, J.P.

1958-01-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

A high intensity positron beam at the Brookhaven reactor

International Nuclear Information System (INIS)

Weber, M.; Lynn, K.G.; Roellig, L.O.; Mills, A.P. Jr.; Moodenbaugh, A.R.

1987-01-01

We describe a high intensity, low energy positron beam utilizing high specific activity /sup 64/Cu sources (870 Ci/g) produced in a reactor with high thermal neutron flux. Fast-to-slow moderation can be performed in a self moderation mode or with a transmission moderator. Slow positron rates up to 1.6 x 10/sup 8/ e/sup +//s with a half life of 12.8 h are calculated. Up to 1.0 x 10/sup 8/ e/sup +//s have been observed. New developments including a Ne moderator and an on-line isotope separation process are discussed. 21 refs., 9 figs

Analysis of high burnup pressurized water reactor fuel using uranium, plutonium, neodymium, and cesium isotope correlations with burnup

International Nuclear Information System (INIS)

Kim, Jung Suk; Jeon, Young Shin; Park, Soon Dal; Ha, Yeong Keong; Song, Kyu Seok

2015-01-01

The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional 235 U burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using 233 U, 242 Pu, 150 Nd, and 133 Cs as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code

Isotopic exchange reactions. Kinetics and efficiency of the reactors using them in isotopic separation

International Nuclear Information System (INIS)

Ravoire, Jean

1979-11-01

In the first part, some definitions and the thermodynamic and kinetic isotopic effect concepts are recalled. In the second part the kinetic laws are established, in homogeneous and heterogeneous medium (one component being on occasions present in both phases), without and with isotopic effects. Emphasis is put on application to separation of isotopes, the separation factor α being close to 1, one isotope being in large excess with respect to the other one. Isotopic transfer is then given by: J = Ka (x - y/α) where x and y are the (isotopic) mole fractions in both phases, Ka may be either the rate of exchange or a transfer coefficient which can be considered as the 'same in both ways' if α-1 is small compared to the relative error on the measure of Ka. The third part is devoted to isotopic exchange reactors. Relationships between their efficiency and kinetics are established in some simple cases: plug cocurrent flow reactors, perfectly mixed reactors, countercurrent reactors without axial mixing. We treat only cases where α and the up flow to down flow ratio is close to 1 so that Murphee efficiency approximately overall efficiency (discrete stage contactors). HTU (phase 1) approximately HTU (phase 2) approximately HETP (columns). In a fourth part, an expression of the isotopic separative power of reactors is proposed and discussed [fr

Bayesian statistics applied to neutron activation data for reactor flux spectrum analysis

International Nuclear Information System (INIS)

Chiesa, Davide; Previtali, Ezio; Sisti, Monica

2014-01-01

Highlights: • Bayesian statistics to analyze the neutron flux spectrum from activation data. • Rigorous statistical approach for accurate evaluation of the neutron flux groups. • Cross section and activation data uncertainties included for the problem solution. • Flexible methodology applied to analyze different nuclear reactor flux spectra. • The results are in good agreement with the MCNP simulations of neutron fluxes. - Abstract: In this paper, we present a statistical method, based on Bayesian statistics, to analyze the neutron flux spectrum from the activation data of different isotopes. The experimental data were acquired during a neutron activation experiment performed at the TRIGA Mark II reactor of Pavia University (Italy) in four irradiation positions characterized by different neutron spectra. In order to evaluate the neutron flux spectrum, subdivided in energy groups, a system of linear equations, containing the group effective cross sections and the activation rate data, has to be solved. However, since the system’s coefficients are experimental data affected by uncertainties, a rigorous statistical approach is fundamental for an accurate evaluation of the neutron flux groups. For this purpose, we applied the Bayesian statistical analysis, that allows to include the uncertainties of the coefficients and the a priori information about the neutron flux. A program for the analysis of Bayesian hierarchical models, based on Markov Chain Monte Carlo (MCMC) simulations, was used to define the problem statistical model and solve it. The first analysis involved the determination of the thermal, resonance-intermediate and fast flux components and the dependence of the results on the Prior distribution choice was investigated to confirm the reliability of the Bayesian analysis. After that, the main resonances of the activation cross sections were analyzed to implement multi-group models with finer energy subdivisions that would allow to determine the

Neutron flux distribution forecasting device of reactor

International Nuclear Information System (INIS)

Uematsu, Hitoshi

1991-01-01

A neutron flux distribution is forecast by using current data obtained from a reactor. That is, the device of the present invention comprises (1) a neutron flux monitor disposed in various positions in the reactor, (2) a forecasting means for calculating and forecasting a one-dimensional neutron flux distribution relative to imaginable events by using data obtained from the neutron flux monitor and physical models, and (3) a display means for displaying the results forecast in the forecasting means to a reactor operation console. Since the forecast values for the one-dimensional neutron flux distribution relative to the imaginable events are calculated in the device of the present invention by using data obtained from the neutron flux monitor and the physical models, the data as a base of the calculation are new and the period for calculating the forecast values can be shortened. Accordingly, although there is a worry of providing some errors in the forecast values, they can be utilized sufficiently as reference data. As a result, the reactor can be operated more appropriately. (I.N.)

Irradiation Tests Supporting LEU Conversion of Very High Power Research Reactors in the US

Energy Technology Data Exchange (ETDEWEB)

Woolstenhulme, N. E.; Cole, J. I.; Glagolenko, I.; Holdaway, K. K.; Housley, G. K.; Rabin, B. H.

2016-10-01

The US fuel development team is developing a high density uranium-molybdenum alloy monolithic fuel to enable conversion of five high-power research reactors. Previous irradiation tests have demonstrated promising behavior for this fuel design. A series of future irradiation tests will enable selection of final fuel fabrication process and provide data to qualify the fuel at moderately-high power conditions for use in three of these five reactors. The remaining two reactors, namely the Advanced Test Reactor and High Flux Isotope Reactor, require additional irradiation tests to develop and demonstrate the fuel’s performance with even higher power conditions, complex design features, and other unique conditions. This paper reviews the program’s current irradiation testing plans for these moderately-high irradiation conditions and presents conceptual testing strategies to illustrate how subsequent irradiation tests will build upon this initial data package to enable conversion of these two very-high power research reactors.

Monitoring the fast neutrons in a high flux: The case for {sup 242}Pu fission chambers

Energy Technology Data Exchange (ETDEWEB)

Filliatre, P.; Jammes, C.; Oriol, L.; Geslot, B. [Commissariat a l' Energie Atomique, DEN/SPEX/LDCI, Centre de Cadarache, F-13108 Saint-Paul-lez-Durance (France); Vermeeren, L. [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

2009-07-01

Fission chambers are widely used for on-line monitoring of neutron fluxes in irradiation reactors. A selective measurement of a component of interest of the neutron flux is possible in principle thanks to a careful choice of the deposit material. However, measuring the fast component is challenging when the flux is high (up to 10{sup 15} n/cm{sup 2}/s) with a significant thermal component. The main problem is that the isotopic content of a material selected for its good response to fast neutrons evolves with irradiation, so that the material is more and more sensitive to thermal neutrons. Within the framework of the FNDS (Fast Neutron Detector System) project, we design tools that simulate the evolution of the isotopic composition and fission rate for several deposits under any given flux. In the case of a high flux with a significant thermal component, {sup 242}Pu is shown after a comprehensive study of all possibilities to be the best choice for measuring the fast component, as long as its purity is sufficient. If an estimate of the thermal flux is independently available, one can correct the signal for that component. This suggests a system of two detectors, one of which being used for such a correction. It is of very high interest when the detectors must be operated up to a high neutron fluence. (authors)

Evolution of the hafnium isotopic composition in the RBMK reactor

International Nuclear Information System (INIS)

Jurkevicius, A.; Remeikis, V.

2002-01-01

The isotopic composition of hafnium in the radial neutron flux sensor of the RBMK-1500 reactor, the rates of the neutron absorption on Hf isotopes and the neutron spectrum in the sensor were numerically modeled. The sequence SAS2 (Shielding Analysis Sequence) program from the package SCALE 4.4A and the HELIOS code system were used for calculations. It has been obtained that the overall neutron absorption rates in hafnium for the sensors located in the 2.4 % and 2.6 % enrichment uranium-erbium nuclear fuel assemblies are by 16 % and 19 % lower than in the 2.0 % enrichment uranium nuclear fuel assemblies. The overall neutron absorption rate in hafnium decreases 2.70-2.75 times due to the sensor burnup to 5800 MW d. The sensitivity of the Hf sensors to the thermal neutron flux increases twice due to the nuclear fuel assembly burnup to 3000 MW d. The corrective factors ξ d (I) at the different integral current I of the sensors and ξ td (E) at the different burnup E of the nuclear fuel assemblies were calculated. The obtained dependence ξ d (I) calculated numerically was compared to the experimental one determined by comparing signals of the fresh sensor and the sensor with the integral current I and by processing repeated calibration results of Hf sensors in RBMK-1500 reactors. The relative relationship coefficients K T (T FA ) were found for all RBMK-1500 nuclear fuel types. (author)

Reactor calculations in aid of isotope production at SAFARI-1

International Nuclear Information System (INIS)

Ball, G.

2003-01-01

Varying levels of reactor physics support is given to the isotope production industry. As the pressures on both the safety limits and economical production of reactor produced isotopes mount, reactor physics calculational support is playing an ever increasing role. Detailed modelling of the reactor, irradiation rigs and target material enables isotope production in reactors to be maximised with respect to yields and quality. NECSA's methodology in this field is described and some examples are given. (author)

Risk analysis of environmental hazards at the High Flux Beam Reactor

International Nuclear Information System (INIS)

Boccio, J.L.; Ho, V.S.; Johnson, D.H.

1994-01-01

In the late 1980s, a Level 1 internal event probabilistic risk assessment (PRA) was performed for the High-Flux Beam Reactor (HFBR), a US Department of Energy research reactor located at Brookhaven National Laboratory. Prior to the completion of that study, a level 1 PRA for external events was initiated, including environmental hazards such as fire, internal flooding, etc. Although this paper provides a brief summary of the risks from environmental hazards, emphasis will be placed on the methodology employed in utilizing industrial event databases for event frequency determination for the HFBR complex. Since the equipment in the HFBR is different from that of, say, a commercial nuclear power plant, the current approach is to categorize the industrial events according to the hazard initiators instead of categorizing by initiator location. But first a general overview of the analysis

Markets for reactor-produced non-fission radioisotopes

International Nuclear Information System (INIS)

Bennett, R.G.

1995-01-01

Current market segments for reactor produced radioisotopes are developed and reported from a review of current literature. Specific radioisotopes studied in is report are the primarily selected from those with major medical or industrial markets, or those expected to have strongly emerging markets. Relative market sizes are indicated. Special emphasis is given to those radioisotopes that are best matched to production in high flux reactors such as the Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory or the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory. A general bibliography of medical and industrial radioisotope applications, trends, and historical notes is included

Considerations in the design of a high power medical isotope production reactor

International Nuclear Information System (INIS)

Ball, Russell M.; Nordyke, William H.; Brown, Roy

2002-01-01

For the low enriched aqueous homogeneous reactor to be economic in the production of medical isotopes, such as Mo-99 and Sr-89, the power level should be of the order of 100 kWth. This is double the earlier designs and this paper discusses the design changes which must be considered to meet this goal. The topics considered are: 1. Heat removal from the reactor solution; 2. Recombination of radiolytic gases; 3. Adequate radiation shielding; 4. Stability of reactor power with fluctuating reactivity; 5. Adequate cooling of the reflector; 6. Independent shutdown mechanisms; 7. Required volume of the reactor; 8. Economic implementation. (author)

Determination of Unknown Neutron Cross Sections for the Production of Medical Isotopes

Energy Technology Data Exchange (ETDEWEB)

Stephen E. Binney

2004-04-09

Calculational assessment and experimental verification of certain neutron cross sections that are related to widely needed new medical isotopes. Experiments were performed at the Oregon State University TRIGA Reactor and the High Flux Irradiation Reactor at Oak Ridge National Laboratory.

Evaluation and Compilation of Neutron Activation Cross Sections for Medical Isotope Production

International Nuclear Information System (INIS)

Binney, Stephen E.

2004-01-01

Calculational assessment and experimental verification of certain neutron cross sections that are related to widely needed new medical isotopes. Experiments were performed at the Oregon State University TRIGA Reactor and the High Flux Irradiation Reactor at Oak Ridge National Laboratory

The development of ex-core neutron flux monitoring system for integral reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S

2004-12-01

Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying

The development of ex-core neutron flux monitoring system for integral reactor

International Nuclear Information System (INIS)

Lee, J. K.; Kwon, H. J.; Park, H. Y.; Koo, I. S.

2004-12-01

Due to the arrangement of major components within the reactor vessel, the integral reactor has relatively long distance between the core support barrel and the reactor vessel when compared with the currently operating plants. So, a neutron flux leakage at the ex-vessel represents a relatively low flux level which may generate some difficulties in obtaining a wide range of neutron flux information including the source range one. This fact may have an impact upon the design and fabrication of an ex-core neutron flux detector. Therefore, it is required to study neutron flux detectors that are suitable for the installation location and characteristics of an integral reactor. The physical constraints of an integral reactor should be considered when one designs and develops the ex-core neutron flux monitoring detectors and their systems. As a possible installation location of the integral reactor ex-core neutron flux detector assembly, two candidate locations are considered, that is, one is between the core support barrel and the reactor vessel and the other is within the Internal Shielding Tank(IST). And, for these locations, some factors such as the environmental requirements and geometrical restrictions are investigated In the case of considering the inside of the IST as a ex-core neutron flux detector installation position, an electrical insulation problem and a low neutron flux measurement problem arose and when considering the inside of the reactor vessel, a detector's sensitivity variation problem, an electrical insulation problem, a detector's insertion and withdrawal problem, and a high neutron flux measurement problem were encountered. Through a survey of the detector installation of the currently operating plants and detector manufacturer's products, the proposed structure and specifications of an ex-core neutron flux detector are suggested. And, the joint ownership strategy for a proposed detector model is also depicted. At the end, by studying the ex

Grasland Stable Isotope Flux Measurements: Three Isotopomers of Carbon Dioxide Measured by QCL Spectroscopy

Science.gov (United States)

Zeeman, M. J.; Tuzson, B.; Eugster, W.; Werner, R. A.; Buchmann, N.; Emmenegger, L.

2007-12-01

To improve our understanding of greenhouse gas dynamics of managed ecosystems such as grasslands, we not only need to investigate the effects of management (e.g., grass cuts) and weather events (e.g., rainy days) on carbon dioxide fluxes, but also need to increase the time resolution of our measurements. Thus, for the first time, we assessed respiration and assimilation fluxes with high time resolution (5Hz) stable isotope measurements at an intensively managed farmland in Switzerland (Chamau, 400m ASL). Two different methods were used to quantify fluxes of carbon dioxide and associated fluxes of stable carbon isotopes: (1) the flux gradient method, and (2) the eddy covariance method. During a week long intensive measurement campaign, we (1) measured mixing ratios of carbon dioxide isotopomers (12C16O2, 12C16O18O, 13C16O2) with a Quantum Cascade Laser (QCL, Aerodyne Inc.) spectroscope and (2) collected air samples for isotope analyses (13C/12C) and (18O/16O) of carbon dioxide by Isotope Ratio Mass Spectrometry (IRMS, Finnigan) every two hours, concurrently along a height profile (z = 0.05; 0.10; 0.31; 2.15m). In the following week, the QCL setup was used for closed-path eddy covariance flux measurement of the carbon dioxide isotopomers, with the air inlet located next to an open-path Infra Red Gas Analyzers (IRGA, LiCor 7500) used simultaneously for carbon dioxide measurements. During this second week, an area of grass inside the footprint was cut and harvested after several days. The first results of in-field continuous QCL measurements of carbon dioxide mixing ratios and their stable isotopic ratios show good agreement with IRGA measurements and isotope analysis of flask samples by IRMS. Thus, QCL spectroscopy is a very promising tool for stable isotope flux investigations.

Prediction of Flow and Temperature Distributions in a High Flux Research Reactor Using the Porous Media Approach

Directory of Open Access Journals (Sweden)

Shanfang Huang

2017-01-01

Full Text Available High thermal neutron fluxes are needed in some research reactors and for irradiation tests of materials. A High Flux Research Reactor (HFRR with an inverse flux trap-converter target structure is being developed by the Reactor Engineering Analysis Lab (REAL at Tsinghua University. This paper studies the safety of the HFRR core by full core flow and temperature calculations using the porous media approach. The thermal nonequilibrium model is used in the porous media energy equation to calculate coolant and fuel assembly temperatures separately. The calculation results show that the coolant temperature keeps increasing along the flow direction, while the fuel temperature increases first and decreases afterwards. As long as the inlet coolant mass flow rate is greater than 450 kg/s, the peak cladding temperatures in the fuel assemblies are lower than the local saturation temperatures and no boiling exists. The flow distribution in the core is homogeneous with a small flow rate variation less than 5% for different assemblies. A large recirculation zone is observed in the outlet region. Moreover, the porous media model is compared with the exact model and found to be much more efficient than a detailed simulation of all the core components.

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

International Nuclear Information System (INIS)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J.P.

2011-01-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 x 10 21 m -2 s -1 , ion fluence: 4 x 10 25 m -2 ) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

Science.gov (United States)

Shimada, Masashi; Hatano, Y.; Calderoni, P.; Oda, T.; Oya, Y.; Sokolov, M.; Zhang, K.; Cao, G.; Kolasinski, R.; Sharpe, J. P.

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 × 1021 m-2 s-1, ion fluence: 4 × 1025 m-2) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

Energy Technology Data Exchange (ETDEWEB)

Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

2011-02-15

The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

Irradiation effects in fused quartz 'Suprasil' as a detector of fission fragments under high flux of reactor neutrons

International Nuclear Information System (INIS)

Moraes, O.M.G. de.

1984-01-01

A systematic study about the registration characteristics of synthetic fused quartz 'Suprasil I' use as a detector of fission fragments under high flux of reactor neutrons and the effects of irradiation on it was performed. Fission fragments of 252 Cf, gamma radiation doses of of 60 Co up to 150 MGy, and integrated neutrons fluxes up to 10 20 n/cm 2 were used. A model to explain the effects on track registration and development characteristics of 'Suprasil I' irradiated on reactors were proposed, based on the obtained results for efficiency an for annealing. (C.G.C.) [pt

Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Bodey, Isaac T [ORNL

2014-05-01

Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

Analytical evaluation of neutron diffusion equation for the geometry of very intense continuous high flux pulsed reactor

International Nuclear Information System (INIS)

Narain, Rajendra

1995-01-01

Using the concept of Very Intense Continuous High Flux Pulsed Reactor to obtain a rotating high flux pulse in an annular core an analytical treatment for the quasi-static solution with a moving reflector is presented. Under quasi-static situation, time averaged values for important parameters like multiplication factor, flux, leakage do not change with time. As a result the instantaneous solution can be considered to be separable in time and space after correcting for the coordinates for the motion of the pulser. The space behaviour of the pulser is considered as exp(-αx 2 ). Movement of delayed neutron precursors is also taken into account. (author). 4 refs

Why does the need of HEU for high flux research reactors remain?

International Nuclear Information System (INIS)

Glaeser, W.

1991-01-01

It has shown that high performance high flux reactors need an ongoing supply of highly enriched uranium. The new fuel materials in their highly enriched version offer prospective for advanced and better neutron sources vital for the future of neutron research. This is another very attractive result of the RERTR programme. One-sided restriction would only provide marginal or no values for research. If we adopt the sometimes expressed views that high enriched RERTR developed fuel should only be made available when unique benefits to mankind could be obtained, then certainly basic research at the forefront belongs to this category. HEU would only pose theoretical difficulties, if it would remain under proper safeguards and obviously this is the way to be pursued. (orig.)

High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Guillen, Donna; Greenwood, Lawrence R.; Parry, James

2014-06-22

A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

Target-fueled nuclear reactor for medical isotope production

Science.gov (United States)

Coats, Richard L.; Parma, Edward J.

2017-06-27

A small, low-enriched, passively safe, low-power nuclear reactor comprises a core of target and fuel pins that can be processed to produce the medical isotope .sup.99Mo and other fission product isotopes. The fuel for the reactor and the targets for the .sup.99Mo production are the same. The fuel can be low enriched uranium oxide, enriched to less than 20% .sup.235U. The reactor power level can be 1 to 2 MW. The reactor is passively safe and maintains negative reactivity coefficients. The total radionuclide inventory in the reactor core is minimized since the fuel/target pins are removed and processed after 7 to 21 days.

Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

International Nuclear Information System (INIS)

Bodey, Isaac T.; Curtis, Franklin G.; Arimilli, Rao V.; Ekici, Kivanc; Freels, James D.

2015-01-01

The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

Research and Development of Multiphysics Models in Support of the Conversion of the High Flux Isotope Reactor to Low Enriched Uranium Fuel

Energy Technology Data Exchange (ETDEWEB)

Bodey, Isaac T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Curtis, Franklin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Arimilli, Rao V. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ekici, Kivanc [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Freels, James D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-11-01

The findings presented in this report are results of a five year effort led by the RRD Division of the ORNL, which is focused on research and development toward the conversion of the High Flux Isotope Reactor (HFIR) fuel from high-enriched uranium (HEU) to low-enriched uranium (LEU). This report focuses on the tasks accomplished by the University of Tennessee Knoxville (UTK) team from the Department of Mechanical, Aerospace, and Biomedical Engineering (MABE) that provided expert support in multiphysics modeling of complex problems associated with the LEU conversion of the HFIR reactor. The COMSOL software was used as the main computational modeling tool, whereas Solidworks was also used in support of computer-aided-design (CAD) modeling of the proposed LEU fuel design. The UTK research has been governed by a statement of work (SOW), which was updated annually to clearly define the specific tasks reported herein. Ph.D. student Isaac T. Bodey has focused on heat transfer and fluid flow modeling issues and has been aided by his major professor Dr. Rao V. Arimilli. Ph.D. student Franklin G. Curtis has been focusing on modeling the fluid-structure interaction (FSI) phenomena caused by the mechanical forces acting on the fuel plates, which in turn affect the fluid flow in between the fuel plates, and ultimately the heat transfer, is also affected by the FSI changes. Franklin Curtis has been aided by his major professor Dr. Kivanc Ekici. M.Sc. student Adam R. Travis has focused two major areas of research: (1) on accurate CAD modeling of the proposed LEU plate design, and (2) reduction of the model complexity and dimensionality through interdimensional coupling of the fluid flow and heat transfer for the HFIR plate geometry. Adam Travis is also aided by his major professor, Dr. Kivanc Ekici. We must note that the UTK team, and particularly the graduate students, have been in very close collaboration with Dr. James D. Freels (ORNL technical monitor and mentor) and have

Production of transplutonium elements in the high flux isotope reactor (HFIR)

International Nuclear Information System (INIS)

Bigelow, J.E.; Corbett, B.L.; King, L.J.; McGuire, S.C.; Sims, T.M.

1980-01-01

The techniques described have been demonstrated to be adequate to predict the contents of transplutonium element production targets which have been irradiated in the HFIR. The deviations, at least for isotopes of mass 253 or less, are generally within the usual analytical uncertainties, or else are for isiotopes which are of little overall import to the program. Work is especially needed to get a better picture of the production of 250 Cm, 254 Es, 255 Es, and ultimately 257 Fm, since researchers are frequently stating their interest in obtaining larger quantities of these rare and difficult-to-produce nuclides

''Sleeping reactor'' irradiations: Shutdown reactor determination of short-lived activation products

International Nuclear Information System (INIS)

Jerde, E.A.; Glasgow, D.C.

1998-01-01

At the High-Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory, the principal irradiation system has a thermal neutron flux (φ) of ∼ 4 x 10 14 n/cm 2 · s, permitting the detection of elements via irradiation of 60 s or less. Irradiations of 6 or 7 s are acceptable for detection of elements with half-lives of as little as 30 min. However, important elements such as Al, Mg, Ti, and V have half-lives of only a few minutes. At HFIR, these can be determined with irradiation times of ∼ 6 s, but the requirement of immediate counting leads to increased exposure to the high activity produced by irradiation in the high flux. In addition, pneumatic system timing uncertainties (about ± 0.5 s) make irradiations of 9 Be(γ,n) 8 Be, the gamma rays principally originating in the spent fuel. Upon reactor SCRAM, the flux drops to ∼ 1 x 10 10 n/cm 2 · s within 1 h. By the time the fuel elements are removed, the flux has dropped to ∼ 6 x 10 8 . Such fluxes are ideal for the determination of short-lived elements such as Al, Ti, Mg, and V. An important feature of the sleeping reactor is a flux that is not constant

Fast flux fluid fuel reactor: A concept for the next generation of nuclear power production

International Nuclear Information System (INIS)

Palmiotti, G.; Feldman, E.E.

1999-01-01

Nuclear energy has not become the preferred method of electrical energy production largely because of economic, safety, and proliferation concerns and challenges posed by nuclear waste disposal. Economies is the most important factor. To reduce the capital costs, the authors propose a compact configuration with a very high power density and correspondingly reduced reactor component sizes. Enhanced efficiency made possible by higher operating temperatures will also improve the economics of the design, and design simplicity will keep capital, operational, and maintenance costs down. The most direct solution to the nuclear waste problem is to eliminate waste production or, at least, minimize its amount and long-term radiotoxicity. This can be achieved by very high burnups, ideally 100%, and by the eventual transmutation of the long-lived fission products in situ. Very high burnups also improve the economics by optimal exploitation of the fuel. Safety concerns can be addressed by an inherently safe reactor design. Because of the intrinsic nature of nuclear materials, there probably is no definitive answer to proliferation concerns for systems that generate neutrons; however, it is important to minimize proliferation risks. The thorium cycle is a promising option because (a) plutonium is produced only in very small quantities, (b) the presence of 232 U makes handling the fuel very difficult and therefore proliferation resistant, and (c) 233 U is a fissile isotope that is less suitable than 239 Pu for making weapons and can be diluted with other uranium isotopes. An additional benefit of the thorium cycle is that it increases nuclear fuel resources by one order of magnitude. A fast flux fluid fuel reactor is a concept that can satisfy all the foregoing requirements. The fluid fuel systems have a very simple structure. Because integrity of the fuel is not an issue, these systems can operate at very high temperatures, can have high power densities, and can achieve very

iMS2Flux – a high–throughput processing tool for stable isotope labeled mass spectrometric data used for metabolic flux analysis

Directory of Open Access Journals (Sweden)

Poskar C Hart

2012-11-01

Full Text Available Abstract Background Metabolic flux analysis has become an established method in systems biology and functional genomics. The most common approach for determining intracellular metabolic fluxes is to utilize mass spectrometry in combination with stable isotope labeling experiments. However, before the mass spectrometric data can be used it has to be corrected for biases caused by naturally occurring stable isotopes, by the analytical technique(s employed, or by the biological sample itself. Finally the MS data and the labeling information it contains have to be assembled into a data format usable by flux analysis software (of which several dedicated packages exist. Currently the processing of mass spectrometric data is time-consuming and error-prone requiring peak by peak cut-and-paste analysis and manual curation. In order to facilitate high-throughput metabolic flux analysis, the automation of multiple steps in the analytical workflow is necessary. Results Here we describe iMS2Flux, software developed to automate, standardize and connect the data flow between mass spectrometric measurements and flux analysis programs. This tool streamlines the transfer of data from extraction via correction tools to 13C-Flux software by processing MS data from stable isotope labeling experiments. It allows the correction of large and heterogeneous MS datasets for the presence of naturally occurring stable isotopes, initial biomass and several mass spectrometry effects. Before and after data correction, several checks can be performed to ensure accurate data. The corrected data may be returned in a variety of formats including those used by metabolic flux analysis software such as 13CFLUX, OpenFLUX and 13CFLUX2. Conclusion iMS2Flux is a versatile, easy to use tool for the automated processing of mass spectrometric data containing isotope labeling information. It represents the core framework for a standardized workflow and data processing. Due to its flexibility

Irradiation experiments and materials testing capabilities in High Flux Reactor in Petten

International Nuclear Information System (INIS)

Luzginova, N.; Blagoeva, D.; Hegeman, H.; Van der Laan, J.

2011-01-01

The text of publication follows: The High Flux Reactor (HFR) in Petten is a powerful multi-purpose research and materials testing reactor operating for about 280 Full Power Days per year. In combination with hot cells facilities, HFR provides irradiation and post-irradiation examination services requested by nuclear energy research and development programs, as well as by industry and research organizations. Using a variety of the custom developed irradiation devices and a large experience in executing irradiation experiments, the HFR is suitable for fuel, materials and components testing for different reactor types. Irradiation experiments carried out at the HFR are mainly focused on the understanding of the irradiation effects on materials; and providing databases for irradiation behavior of materials to feed into safety cases. The irradiation experiments and materials testing at the HFR include the following issues. First, materials irradiation to support the nuclear plant life extensions, for instance, characterization of the reactor pressure vessel stainless steel claddings to insure structural integrity of the vessel, as well as irradiation of the weld material coupons to neutron fluence levels that are representative for Light Water Reactors (LWR) internals applications. Secondly, development and qualification of the structural materials for next generation nuclear fission reactors as well as thermo-nuclear fusion machines. The main areas of interest are in both conventional stainless steel and advanced reduced activation steels and special alloys such as Ni-base alloys. For instance safety-relevant aspects of High Temperature Reactors (HTR) such as the integrity of fuel and structural materials with increasing neutron fluence at typical HTR operating conditions has been recently assessed. Thirdly, support of the fuel safety through several fuel irradiation experiments including testing of pre-irradiated LWR fuel rods containing UO 2 or MOX fuel. Fourthly

First result of deuterium retention in neutron-irradiated tungsten exposed to high flux plasma in TPE

Energy Technology Data Exchange (ETDEWEB)

Shimada, Masashi, E-mail: Masashi.Shimada@inl.gov [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Hatano, Y. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan); Calderoni, P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Oda, T. [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo 113-8656 (Japan); Oya, Y. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Sokolov, M. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Zhang, K. [Hydrogen Isotope Research Center, University of Toyama, Toyama 930-8555 (Japan); Cao, G. [Department of Engineering Physics, University of Wisconsin-Madison, Madison, WI 53706 (United States); Kolasinski, R. [Hydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, CA 94551 (United States); Sharpe, J.P. [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

2011-08-01

With the Japan-US joint research project Tritium, Irradiations, and Thermofluids for America and Nippon (TITAN), an initial set of tungsten samples (99.99% purity, A.L.M.T. Co.) were irradiated by high flux neutrons at 323 K to 0.025 dpa in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Subsequently, one of the neutron-irradiated tungsten samples was exposed to a high-flux deuterium plasma (ion flux: 5 x 10{sup 21} m{sup -2} s{sup -1}, ion fluence: 4 x 10{sup 25} m{sup -2}) in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory (INL). The deuterium retention in the neutron-irradiated tungsten was 40% higher in comparison to the unirradiated tungsten. The observed broad desorption spectrum from neutron-irradiated tungsten and associated TMAP modeling of the deuterium release suggest that trapping occurs in the bulk material at more than three different energy sites.

EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

Energy Technology Data Exchange (ETDEWEB)

Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1958-07-01

The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

The proposed use of low enriched uranium fuel in the High Flux Australian Reactor (HIFAR)

International Nuclear Information System (INIS)

Vittorio, D.; Durance, G.

2002-01-01

The Australian Nuclear Science and Technology Organisation (ANSTO) operates the High Flux Australian Reactor (HIFAR). HIFAR commenced operation in the late 1950's with fuel elements containing uranium enriched to 93%. From that time the level of enrichment has gradually decreased to the current level of 60%. It is now proposed to further reduce the enrichment of HIFAR fuel to <20% by utilising LEU fuel assemblies manufactured by RISO National Laboratory, that were originally intended for use in the DR-3 reactor. Minor modifications have been made to the assemblies to adapt them for use in HIFAR. A detailed design review has been performed and initial safety analysis and reactor physics calculations are to be submitted to ARPANSA as part of a four-stage approval process. (author)

Safety and quality management at the high flux reactor Petten

International Nuclear Information System (INIS)

Zurita, A.; Ahlf, J.

1995-01-01

The High Flux Reactor (HFR) is one high power multi-purpose materials testing research reactor of the tank-in-pool type, cooled and moderated by light-water. It is operated at 45 MW at a prescribed schedule of 11 cycles per year, each comprising 25 operation days and three shut-down days. Since the licence for the operation of HFR was granted in 1962, a total of 14 amendments to the original licence have been made following different modifications in the installations. In the meantime, international nuclear standards were developed, especially in the framework of the NUSS programme of the IAEA, which were adopted by the Dutch Licensing Authorities. In order to implement the new standards, the situation at the HFR was comprehensively reviewed in the course of an audit performed by the Dutch Licensing Authorities in 1988. This also resulted in formulating the task of setting-up an 'HFR - Integral Quality Assurance Handbook' (HFR-IQAD) involving both organizations JRCIAM and ECN, which had the unique framework and basic guideline to assure the safe and efficient operation and exploitation of the HFR and to promote safety and quality in all aspects of HFR related activities. The assurance of safe and efficient operation and exploitation of the HFR is condensed together under the concepts of safety and quality of services and is achieved through the safety and quality management. (orig.)

13C metabolic flux analysis: optimal design of isotopic labeling experiments.

Science.gov (United States)

Antoniewicz, Maciek R

2013-12-01

Measuring fluxes by 13C metabolic flux analysis (13C-MFA) has become a key activity in chemical and pharmaceutical biotechnology. Optimal design of isotopic labeling experiments is of central importance to 13C-MFA as it determines the precision with which fluxes can be estimated. Traditional methods for selecting isotopic tracers and labeling measurements did not fully utilize the power of 13C-MFA. Recently, new approaches were developed for optimal design of isotopic labeling experiments based on parallel labeling experiments and algorithms for rational selection of tracers. In addition, advanced isotopic labeling measurements were developed based on tandem mass spectrometry. Combined, these approaches can dramatically improve the quality of 13C-MFA results with important applications in metabolic engineering and biotechnology. Copyright © 2013 Elsevier Ltd. All rights reserved.

Experimental studies on mitigation of LOCA for a high flux research reactor

International Nuclear Information System (INIS)

Saxena, A.K.

2006-01-01

Experimental studies on the rewetting behaviour of hot vertical annular channels were performed to study the mitigation of consequences of loss of coolant accident (LOCA) for a high flux research reactor. Studies were carried out to study the rewetting behaviour with hot inner tube, for bottom flooding and top flow rewetting conditions. The tube was made of stainless steel. Experiments were conducted for water flow rates in the annulus upto 7 litres per minute (l pm) (11.7 x 10 -5 m 3 s -1 ). The initial surface temperature of the inner tube was varied from 200 to 500 degC. (author)

maximum neutron flux at thermal nuclear reactors

International Nuclear Information System (INIS)

Strugar, P.

1968-10-01

Since actual research reactors are technically complicated and expensive facilities it is important to achieve savings by appropriate reactor lattice configurations. There is a number of papers, and practical examples of reactors with central reflector, dealing with spatial distribution of fuel elements which would result in higher neutron flux. Common disadvantage of all the solutions is that the choice of best solution is done starting from the anticipated spatial distributions of fuel elements. The weakness of these approaches is lack of defined optimization criteria. Direct approach is defined as follows: determine the spatial distribution of fuel concentration starting from the condition of maximum neutron flux by fulfilling the thermal constraints. Thus the problem of determining the maximum neutron flux is solving a variational problem which is beyond the possibilities of classical variational calculation. This variational problem has been successfully solved by applying the maximum principle of Pontrjagin. Optimum distribution of fuel concentration was obtained in explicit analytical form. Thus, spatial distribution of the neutron flux and critical dimensions of quite complex reactor system are calculated in a relatively simple way. In addition to the fact that the results are innovative this approach is interesting because of the optimization procedure itself [sr

Neutron flux measurements in PUSPATI Triga Reactor

International Nuclear Information System (INIS)

Gui Ah Auu; Mohamad Amin Sharifuldin Salleh; Mohamad Ali Sufi.

1983-01-01

Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm 2 sec and 1.10E+12n/cm 2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

2008-10-29

The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

Flux distribution measurements in the Bruce B Unit 6 reactor using a transportable traveling flux detector system

International Nuclear Information System (INIS)

Leung, T.C.; Drewell, N.H.; Hall, D.S.; Lopez, A.M.

1987-01-01

A transportable traveling flux detector (TFD) system for use in power reactors has been developed and tested at Chalk River Nuclear Labs. in Canada. It consists of a miniature fission chamber, a motor drive mechanism, a computerized control unit, and a data acquisition subsystem. The TFD system was initially designed for the in situ calibration of fixed self-powered detectors in operating power reactors and for flux measurements to verify reactor physics calculations. However, this system can also be used as a general diagnostic tool for the investigation of apparent detector failures and flux anomalies and to determine the movement of reactor internal components. This paper describes the first successful use of the computerized TFD system in an operating Canada deuterium uranium (CANDU) power reactor and the results obtained from the flux distribution measurements. An attempt is made to correlate minima in the flux profile with the locations of fuel channels so that future measurements can be used to determine the sag of the channels. Twenty-seven in-core flux detector assemblies in the 855-MW (electric) Unit 6 reactor of the Ontario Hydro Bruce B Generating Station were scanned

Low helium flux from the mantle inferred from simulations of oceanic helium isotope data

Science.gov (United States)

Bianchi, Daniele; Sarmiento, Jorge L.; Gnanadesikan, Anand; Key, Robert M.; Schlosser, Peter; Newton, Robert

2010-09-01

The high 3He/ 4He isotopic ratio of oceanic helium relative to the atmosphere has long been recognized as the signature of mantle 3He outgassing from the Earth's interior. The outgassing flux of helium is frequently used to normalize estimates of chemical fluxes of elements from the solid Earth, and provides a strong constraint to models of mantle degassing. Here we use a suite of ocean general circulation models and helium isotope data obtained by the World Ocean Circulation Experiment to constrain the flux of helium from the mantle to the oceans. Our results suggest that the currently accepted flux is overestimated by a factor of 2. We show that a flux of 527 ± 102 mol year - 1 is required for ocean general circulation models that produce distributions of ocean ventilation tracers such as radiocarbon and chlorofluorocarbons that match observations. This new estimate calls for a reevaluation of the degassing fluxes of elements that are currently tied to the helium fluxes, including noble gases and carbon dioxide.

HTR fuel research in the HTR-TN network on the high flux reactor

Energy Technology Data Exchange (ETDEWEB)

Guidez, J.; Conrad, R.; Sevini, P.; Burghartz, M. [HFR Unit, Institute for Advanced Materials, European Commission, Joint Research Centre, Petten (Netherlands); Languille, A. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Guillermier, P. [FRAMATOME ANP, 69 - Lyon (France); Bakker, K. [Nuclear Research and Consultancy Group, Petten (Netherlands); Nabielek, H. [Forschungszentrum Juelich (Germany)

2001-07-01

Foremost, this paper explains the economic and strategic reasons for the comeback of the HTR reactor as one of the most promising reactors in the future. To study all the points related to HTR technology, a European network called HTR-TN was created in April 2000, with actually twenty European companies involved. This paper explains the organisation of the network and the related task-groups. In the field of fuel, one of these task-groups works on the fuel cycle and another works on the fuel itself in order to validate by testing HTR fuel possibilities. To this aim, an experimental loop is under construction in the HFR reactor to test full-size pebble type fuel elements and another under study to test compact fuel possibilities. These loops are based on all the experience accumulated by the High Flux Reactor in the years 70-90, when a lot of test were performed for fuel and material for the HTR technology and the facility design uses all the existing HFR knowledge. In conclusion, a host of research work, co-ordinated in the frame of a European network HTR-TN has begun. and should allow in the near future a substantial progress in the knowledge of this very promising fuel. (author)

HTR fuel research in the HTR-TN network on the high flux reactor

International Nuclear Information System (INIS)

Guidez, J.; Conrad, R.; Sevini, P.; Burghartz, M.; Languille, A.; Guillermier, P.; Bakker, K.; Nabielek, H.

2001-01-01

Foremost, this paper explains the economic and strategic reasons for the comeback of the HTR reactor as one of the most promising reactors in the future. To study all the points related to HTR technology, a European network called HTR-TN was created in April 2000, with actually twenty European companies involved. This paper explains the organisation of the network and the related task-groups. In the field of fuel, one of these task-groups works on the fuel cycle and another works on the fuel itself in order to validate by testing HTR fuel possibilities. To this aim, an experimental loop is under construction in the HFR reactor to test full-size pebble type fuel elements and another under study to test compact fuel possibilities. These loops are based on all the experience accumulated by the High Flux Reactor in the years 70-90, when a lot of test were performed for fuel and material for the HTR technology and the facility design uses all the existing HFR knowledge. In conclusion, a host of research work, co-ordinated in the frame of a European network HTR-TN has begun. and should allow in the near future a substantial progress in the knowledge of this very promising fuel. (author)

Utilization of cold neutron beams at intermediate flux reactors

International Nuclear Information System (INIS)

Clark, D.D.

1992-01-01

With the advent of cold neutron beam (CNB) facilities at U.S. reactors [National Institute of Standards and Technology (NIST) in 1991; Cornell University and the University of Texas at Austin, anticipated in 1992], it is appropriate to reexamine the types of research for which they are likely to be best suited or uniquely suited. With the exception of a small-angle neutron scattering facility at Brookhaven National Laboratory, there has been no prior experience in the United States with such beams, but they have been extensively used at European reactors where cold neutron sources and neutron guides were developed some years age. This paper does not discuss specialized cases such as ultracold neutrons or very high flux facilities such as the Institute Laue-Langevin ractor and the proposed advanced neutron source. Instead, it concentrates on potential utilization of CNBs at intermediate-flux reactors such as at Cornell and Texas, i.e., in the 1-MW range and operated <24 h a day

Experimental Study of Thermal Crisis in Connection with Tokamak Reactor High Heat Flux Components

International Nuclear Information System (INIS)

Gallo, D.; Giardina, M.; Castiglia, F.; Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

2000-01-01

The results of an experimental research on high heat flux thermal crisis in forced convective subcooled water flow, under operative conditions of interest to the thermal-hydraulic design of TOKAMAK fusion reactors, are here reported. These experiments, carried out in the framework of a collaboration between the Nuclear Engineering Department of Palermo University and the National Institute of Thermal - Fluid Dynamics of the ENEA - Casaccia (Rome), were performed on the STAF (Scambio Termico Alti Flussi) water loop and consisted, essentially, in a high speed photographic study which enabled focusing several information on bubble characteristics and flow patterns taking place during the burnout phenomenology

Impact induced response spectrum for the safety evaluation of the high flux isotope reactor

International Nuclear Information System (INIS)

Chang, S.J.

1997-01-01

The dynamic impact to the nearby HFIR reactor vessel caused by heavy load drop is analyzed. The impact calculation is carried out by applying the ABAQUS computer code. An impact-induced response spectrum is constructed in order to evaluate whether the HFIR vessel and the shutdown mechanism may be disabled. For the frequency range less than 10 Hz, the maximum spectral velocity of impact is approximately equal to that of the HFIR seismic design-basis spectrum. For the frequency range greater than 10 Hz, the impact-induced response spectrum is shown to cause no effect to the control rod and the shutdown mechanism. An earlier seismic safety assessment for the HFIR control and shutdown mechanism was made by EQE. Based on EQE modal solution that is combined with the impact-induced spectrum, it is concluded that the impact will not cause any damage to the shutdown mechanism, even while the reactor is in operation. The present method suggests a general approach for evaluating the impact induced damage to the reactor by applying the existing finite element modal solution that has been carried out for the seismic evaluation of the reactor

Measurements of flux and isotopic composition of soil carbon dioxide

International Nuclear Information System (INIS)

Gorczyca, Z.; Rozanski, K.; Kuc, T.

2002-01-01

The flux and isotope composition of soil CO 2 has been regularly measured at three sites located in the southern Poland, during the time period: January 1998 - October 2000. They represent typical ecosystems appearing in central Europe: (i) mixed forest; (ii) cultivated agricultural field; (iii) grassland. To monitor the flux and isotopic composition of soil CO 2 , a method based on the inverted cup principle was adopted. The flux of soil CO 2 reveals distinct seasonal fluctuations, with maximum values up to ca. 25 mmol/m 2 /h during sommer months and around ten times lower values during winter time. Also significant differences among the monitored sites were detected, the flux density of this gas being highest for the mixed forest site and ca. two times lower for the cultivated grassland. Carbon-13 content of the soil CO 2 reveals little seasonal variability, with δ 13 C values essentially reflecting the isotopic composition of the soil organic matter and the vegetation type. The carbon-14 content of soil CO 2 flux also reveals slight seasonality, with lower δ 14 C values recorded during winter time. Significantly lower δ 14 C values recorded during winter time. Significantly lower δ 14 C values were recorded at depth. (author)

High Fluency Low Flux Embrittlement Models of LWR Reactor Pressure Vessel Embrittlement and a Supporting Database from the UCSB ATR-2 Irradiation Experiment

Energy Technology Data Exchange (ETDEWEB)

Odette, G. Robert [Univ. of California, Santa Barbara, CA (United States)

2017-01-24

Reactor pressure vessel embrittlement may limit the lifetime of light water reactors (LWR). Embrittlement is primarily caused by formation of nano-scale precipitates, which cause hardening and a subsequent increase in the ductile-to-brittle transition temperature of the steel. While the effect of Cu has historically been the largest research focus of RPV embrittlement, there is increasing evidence that Mn, Ni and Si are likely to have a large effect at higher fluence, where Mn-Ni-Si precipitates can form, even in the absence of Cu. Therefore, extending RPV lifetimes will require a thorough understanding of both precipitation and embrittlement at higher fluences than have ever been observed in a power reactor. To address this issue, test reactors that irradiate materials at higher neutron fluxes than power reactors are used. These experiments at high neutron flux can reach extended life neutron fluences in only months or several years. The drawback of these test irradiations is that they add additional complexity to interpreting the data, as the irradiation flux also plays a role into both precipitate formation and irradiation hardening and embrittlement. This report focuses on developing a database of both microstructure and mechanical property data to better understand the effect of flux. In addition, a previously developed model that enables the comparison of data taken over a range of neutron flux is discussed.

Absolute measurement of neutron fluxes inside the reactor core

International Nuclear Information System (INIS)

Ajdacic, S. V.

1964-10-01

The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li 6 -semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li 6 -semiconductor spectrometer with plane geometry is given. A new type of Li 6 -semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li 6 -spectrometer made (author)

Seismic hazard studies for the high flux beam reactor at Brookhaven National Laboratory

International Nuclear Information System (INIS)

Costantino, C.J.; Heymsfield, E.; Park, Y.J.; Hofmayer, C.H.

1991-01-01

This paper presents the results of a calculation to determine the site specific seismic hazard appropriate for the deep soil site at Brookhaven National Laboratory (BNL) which is to be used in the risk assessment studies being conducted for the High Flux Beam Reactor (HFBR). The calculations use as input the seismic hazard defined for the bedrock outcrop by a study conducted at Lawrence Livermore National Laboratory (LLNL). Variability in site soil properties were included in the calculations to obtain the seismic hazard at the ground surface and compare these results with those using the generic amplification factors from the LLNL study

Evaluation of selected ex-reactor accidents related to the tritium and medical isotope production mission at the FFTF

Energy Technology Data Exchange (ETDEWEB)

Himes, D.A.

1997-11-17

The Fast Flux Test Facility (FFTF) has been proposed as a production facility for tritium and medical isotopes. A range of postulated accidents related to ex-reactor irradiated fuel and target handling were identified and evaluated using new source terms for the higher fuel enrichment and for the tritium and medical isotope targets. In addition, two in-containment sodium spill accidents were re-evaluated to estimate effects of increased fuel enrichment and the presence of the Rapid Retrieval System. Radiological and toxicological consequences of the analyzed accidents were found to be well within applicable risk guidelines.

Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

International Nuclear Information System (INIS)

Fischer, U.; Herring, S.; Hogenbirk, A.; Leichtle, D.; Nagao, Y.; Pijlgroms, B.J.; Ying, A.

2000-01-01

Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2 O, Li 4 SiO 4 and Li 2 TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6 Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W(T) can be met when taking into account these prerequisites. Irradiation times in the order of 2-3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation

Neutron flux enhancement in the NRAD reactor

International Nuclear Information System (INIS)

Weeks, A.A.; Heidel, C.C.; Imel, G.R.

1988-01-01

In 1987 a series of experiments were conducted at the NRAD reactor facility at Argonne National Laboratory - West (ANL-W) to investigate the possibility of increasing the thermal neutron content at the end of the reactor's east beam tube through the use of hydrogenous flux traps. It was desired to increase the thermal flux for a series of experiments to be performed in the east radiography cell, in which the enhanced flux was required in a relatively small volume. Hence, it was feasible to attempt to focus the cross section of the beam to a smaller area. Two flux traps were constructed from unborated polypropylene and tested to determine their effectiveness. Both traps were open to the entire cross-sectional area of the neutron beam (as it emerges from the wall and enters the beam room). The sides then converged such that at the end of the trap the beam would be 'focused' to a greater intensity. The differences in the two flux traps were primarily in length, and hence angle to the beam as the inlet and outlet cross-sectional areas were held constant. The experiments have contributed to the design of a flux trap in which a thermal flux of nearly 10 9 was obtained, with an enhancement of 6.61

Alize 3 - first critical experiment for the franco-german high flux reactor - calculations

International Nuclear Information System (INIS)

Scharmer, K.

1969-01-01

The results of experiments in the light water cooled D 2 O reflected critical assembly ALIZE III have been compared to calculations. A diffusion model was used with 3 fast and epithermal groups and two overlapping thermal groups, which leads to good agreement of calculated and measured power maps, even in the case of strong variations of the neutron spectrum in the core. The difference of calculated and measured k eff was smaller than 0.5 per cent δk/k. Calculations of void and structure material coefficients of the reactivity of 'black' rods in the reflector, of spectrum variations (Cd-ratio, Pu-U-ratio) and to the delayed photoneutron fraction in the D 2 O reflector were made. Measurements of the influence of beam tubes on reactivity and flux distribution in the reflector were interpreted with regard to an optimum beam tube arrangement for the Franco- German High Flux Reactor. (author) [fr

Neutron-antineutron transition search at HFIR reactor

International Nuclear Information System (INIS)

Kamyshkov, Yuri A.

1997-01-01

A new experiment to search for neutron-antineutron transitions was recently proposed for High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). In this paper the physics motivation of a new search, the scheme and the discovery potential of the proposed HFIR-based experiment are discussed

Neutron-antineutron transition search at HFIR Reactor

International Nuclear Information System (INIS)

Kamyshkov, Y.A.

1997-01-01

A new experiment to search for neutron-antineutron transitions was recently proposed for High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). In this paper the physics motivation of a new search, the scheme and the discovery potential of the proposed HFIR-based experiment are discussed

Practical course on reactor instrumentation

International Nuclear Information System (INIS)

Boeck, H.; Villa, M.

2004-06-01

This course is based on the description of the instrumentation of the TRIGA-reactor Vienna, which is used for training research and isotope production. It comprises the following chapters: 1. instrumentation, 2. calibration of the nuclear channels, 3. rod drop time of the control rods, 4. neutron flux density measurements using compensated ionization, 5. neutron flux density measurement with fission chambers (FC), 6. neutron flux density measurement with self-powered neutron detectors (SPND), 7. pressurized water reactor simulator, 8. verification of the radiation level during reactor operation. There is one appendix about neutron-sensitive thermocouples. (nevyjel)

δ2H isotopic flux partitioning of evapotranspiration over a grass field following a water pulse and subsequent dry down

Science.gov (United States)

Good, Stephen P.; Soderberg, Keir; Guan, Kaiyu; King, Elizabeth G.; Scanlon, Todd M.; Caylor, Kelly K.

2014-02-01

The partitioning of surface vapor flux (FET) into evaporation (FE) and transpiration (FT) is theoretically possible because of distinct differences in end-member stable isotope composition. In this study, we combine high-frequency laser spectroscopy with eddy covariance techniques to critically evaluate isotope flux partitioning of FET over a grass field during a 15 day experiment. Following the application of a 30 mm water pulse, green grass coverage at the study site increased from 0 to 10% of ground surface area after 6 days and then began to senesce. Using isotope flux partitioning, transpiration increased as a fraction of total vapor flux from 0% to 40% during the green-up phase, after which this ratio decreased while exhibiting hysteresis with respect to green grass coverage. Daily daytime leaf-level gas exchange measurements compare well with daily isotope flux partitioning averages (RMSE = 0.0018 g m-2 s-1). Overall the average ratio of FT to FET was 29%, where uncertainties in Keeling plot intercepts and transpiration composition resulted in an average of uncertainty of ˜5% in our isotopic partitioning of FET. Flux-variance similarity partitioning was partially consistent with the isotope-based approach, with divergence occurring after rainfall and when the grass was stressed. Over the average diurnal cycle, local meteorological conditions, particularly net radiation and relative humidity, are shown to control partitioning. At longer time scales, green leaf area and available soil water control FT/FET. Finally, we demonstrate the feasibility of combining isotope flux partitioning and flux-variance similarity theory to estimate water use efficiency at the landscape scale.

Excitation of neutron flux waves in reactor core transients

International Nuclear Information System (INIS)

Carew, J.F.; Neogy, P.

1983-01-01

An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

International Nuclear Information System (INIS)

Cartier, J.; Casoli, P.; Chappert, F.

2013-01-01

In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

International Nuclear Information System (INIS)

Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

2010-01-01

Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

Epithermal neutron flux characterization of the TRIGA Mark III reactor, Salazar, Mexico, for use in Internal Neutron Activation Analysis

International Nuclear Information System (INIS)

Diaz Rizo, O.; Herrera Peraza, E.

1996-01-01

The non ideality of the epithermal neutron flux distribution at a reactor site parameter (made, using Chloramine-T method. Radiochemical purity and stability of the labelled product were determined by radiochromatography. The labelled Melagenine-II showed two radioactive fractions thermal-to-epithermal neutron ratio (f) were determined in the 3 typical irradiations positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using the Cd-ratio for multi monitor and bare bi-isotopic monitor methods respectively. This characterization is of use in the K o - method of neutron activation analysis, recently introduced at the Institute

The PALLAS research and isotope reactor project status

International Nuclear Information System (INIS)

Van Der Schaaf, B.; De Jong, P.

2010-01-01

In the European Union the first generation research reactors is nearing their end of life condition. Several committees recommend a comprehensive set of reactors in the EU, amongst them the replacement for the HFR research and isotope reactor in Petten: PALLAS. The business case for PALLAS supports a future for a research and isotope reactor in Petten as a perfect fit for the future EU set of test reactors. The tender for PALLAS started in 2007, following the EU rules for tendering complex objects with the competitive dialogue. This procedure involved an extensive consultation phase between individual tendering companies and NRG, resulting in definitive specifications in summer 2008. The evaluation of offers, including conceptual designs, took place in summer 2009. At present NRG is still active in the acquisition of the funding for the project. The licensing path has been started in autumn 2009 with a initiation note on the environmental impact assessment, EIA. The public hearings held in the lead to the advice from the national EIA committee for the approach of the assessment. The PALLAS project team in Petten will guide the design and build processes. It is also responsible for the licensing of the building and operation of PALLAS. The team also manages the design and construction for the infrastructure, such as cooling devices, including remnant heat utilization, and utility provisions. A particular responsibility for the team is the design and construction of experimental and isotope capsules, based on launch customer requirements. (author)

Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

Absolute measurement of neutron fluxes inside the reactor core

Energy Technology Data Exchange (ETDEWEB)

Ajdacic, S V [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1964-10-15

The subject of this work is the development and study of two methods of neutron measurements in nuclear reactors, the new method of high neutron flux measurements and the Li{sup 6}-semiconductor neutron spectrometer. This work is presented in four sections: Section I. The introduction explains the need for neutron measurements in reactors. A critical survey is given of the existing methods of high neutron flux measurement and methods of fast neutron spectrum determination. Section II. Theoretical basis of the work of semiconductor counters and their most important characteristics are given. Section III. The main point of this section is in presenting the basis of the new method which the author developed, i.e., the long-tube method, and the results obtained by it, with particular emphasis on absolute measurement of high neutron fluxes. Advantages and limitations of this method are discussed in details at the end of this section. Section IV. A comparison of the existing semiconductor neutron spectrometers is made and their advantages and shortcomings underlined. A critical analysis of the obtained results with the Li{sup 6}-semiconductor spectrometer with plane geometry is given. A new type of Li{sup 6}-semiconductor spectrometer is described, its characteristics experimentally determined, and a comparison of it with a classical Li{sup 6}-spectrometer made (author)

Transport of spent nuclear fuel from the High Flux Beam Reactor

International Nuclear Information System (INIS)

Holland, Michael; Carelli, Joseph; Shelton, Thomas

1997-01-01

The shipment of more than 1000 elements of spent nuclear fuel (SNF) from the Department of Energy's Brookhaven National Laboratory (BNL) High Flux Beam Reactor (HFBR) to the Department's Savannah River Site (SRS) for long term interim storage required overcoming several significant obstacles. The project management team was comprised of DOE, BNL and NAC International personnel. This achievement involved coordinating the efforts of numerous government and contractor organizations such as the U.S. Coast Guard, the U.S. Nuclear Regulatory Commission, state and local governments, marine and motor carriers, and carrier inspectors. Unique experience was gained during development and execution of the project in the following areas: dry transfer of SNF to shipping casks; inter-modal transfers; logistics; cask licensing by the Nuclear Regulatory Commission (NRC); compliance with environmental regulations; transportation plan development, and stakeholder outreach and coordination

Scoping assessment on medical isotope production at the Fast Flux Test Facility

International Nuclear Information System (INIS)

Scott, S.W.

1997-01-01

The Scoping Assessment addresses the need for medical isotope production and the capability of the Fast Flux Test Facility to provide such isotopes. Included in the discussion are types of isotopes used in radiopharmaceuticals, which types of cancers are targets, and in what way isotopes provide treatment and/or pain relief for patients

Scoping assessment on medical isotope production at the Fast Flux Test Facility

Energy Technology Data Exchange (ETDEWEB)

Scott, S.W.

1997-08-29

The Scoping Assessment addresses the need for medical isotope production and the capability of the Fast Flux Test Facility to provide such isotopes. Included in the discussion are types of isotopes used in radiopharmaceuticals, which types of cancers are targets, and in what way isotopes provide treatment and/or pain relief for patients.

Partitioning evapotranspiration fluxes with water stable isotopic measurements: from the lab to the field

Science.gov (United States)

Quade, M. E.; Brueggemann, N.; Graf, A.; Rothfuss, Y.

2017-12-01

Water stable isotopes are powerful tools for partitioning net into raw water fluxes such as evapotranspiration (ET) into soil evaporation (E) and plant transpiration (T). The isotopic methodology for ET partitioning is based on the fact that E and T have distinct water stable isotopic compositions, which in turn relies on the fact that each flux is differently affected by isotopic kinetic effects. An important work to be performed in parallel to field measurements is to better characterize these kinetic effects in the laboratory under controlled conditions. A soil evaporation laboratory experiment was conducted to retrieve characteristic values of the kinetic fractionation factor (αK) under varying soil and atmospheric water conditions. For this we used a combined soil and atmosphere column to monitor the soil and atmospheric water isotopic composition profiles at a high temporal and vertical resolution in a nondestructive manner by combining micro-porous membranes and laser spectroscopy. αK was calculated by using a well-known isotopic evaporation model in an inverse mode with the isotopic composition of E as one input variable, which was determined using a micro-Keeling regression plot. Knowledge on αK was further used in the field (Selhausen, North Rhine-Westphalia, Germany) to partition ET of catch crops and sugar beet (Beta vulgaris) during one growing season. Soil and atmospheric water isotopic profiles were measured automatically across depths and heights following a similar modus operandi as in the laboratory experiment. Additionally, a newly developed continuously moving elevator was used to obtain water vapor isotopic composition profiles with a high vertical resolution between soil surface, plant canopy and atmosphere. Finally, soil and plant samples were collected destructively to provide a comparison with the traditional isotopic methods. Our results illustrate the changing proportions of T and E along the growing season and demonstrate the

Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

International Nuclear Information System (INIS)

Knight, R.W.; Morin, R.A.

1999-01-01

The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U 3 O 8 powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated

Fabrication procedures for manufacturing High Flux Isotope Reactor fuel elements - 2

Energy Technology Data Exchange (ETDEWEB)

Knight, R.W.; Morin, R.A.

1999-12-01

The original fabrication procedures written in 1968 delineated the manufacturing procedures at that time. Since 1968, there have been a number of procedural changes. This rewrite of the fabrication procedures incorporates these changes. The entire fuel core of this reactor is made up of two fuel elements. Each element consists of one annular array of fuel plates. These annuli are identified as the inner and outer fuel elements, since one fits inside the other. The inner element consists of 171 identical fuel plates, and the outer element contains 369 identical fuel plates differing slightly from those in the inner element. Both sets of fuel plates contain U{sub 3}O{sub 8} powder as the fuel, dispersed in an aluminum powder matrix and clad with aluminum. Procedures for manufacturing and inspection of the fuel elements are described and illustrated.

Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

Energy Technology Data Exchange (ETDEWEB)

Cartier, J. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France); Casoli, P. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, Valduc, F-21120 Is sur Tille (France); Chappert, F. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France)

2013-07-01

In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

Reactor Fuel Isotopics and Code Validation for Nuclear Applications

Energy Technology Data Exchange (ETDEWEB)

Francis, Matthew W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Weber, Charles F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pigni, Marco T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauld, Ian C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-02-01

Experimentally measured isotopic concentrations of well characterized spent nuclear fuel (SNF) samples have been collected and analyzed by previous researchers. These sets of experimental data have been used extensively to validate the accuracy of depletion code predictions for given sets of burnups, initial enrichments, and varying power histories for different reactor types. The purpose of this report is to present the diversity of data in a concise manner and summarize the current accuracy of depletion modeling. All calculations performed for this report were done using the Oak Ridge Isotope GENeration (ORIGEN) code, an internationally used irradiation and decay code solver within the SCALE comprehensive modeling and simulation code. The diversity of data given in this report includes key actinides, stable fission products, and radioactive fission products. In general, when using the current ENDF/B-VII.0 nuclear data libraries in SCALE, the major actinides are predicted to within 5% of the measured values. Large improvements were seen for several of the curium isotopes when using improved cross section data found in evaluated nuclear data file ENDF/B-VII.0 as compared to ENDF/B-V-based results. The impact of the flux spectrum on the plutonium isotope concentrations as a function of burnup was also shown. The general accuracy noted for the actinide samples for reactor types with burnups greater than 5,000 MWd/MTU was not observed for the low-burnup Hanford B samples. More work is needed in understanding these large discrepancies. The stable neodymium and samarium isotopes were predicted to within a few percent of the measured values. Large improvements were seen in prediction for a few of the samarium isotopes when using the ENDF/B-VII.0 libraries compared to results obtained with ENDF/B-V libraries. Very accurate predictions were obtained for ¹³³Cs and ¹⁵³Eu. However, the predicted values for the stable ruthenium and rhodium isotopes varied

Effect of spectral characterization of gaseous fuel reactors on transmutation and burning of actinides

Energy Technology Data Exchange (ETDEWEB)

Fung, C.; Anghaie, S. [Florida Univ., Wilmington, NC (United States)

2007-07-01

Gaseous Core Reactors (GCR) are fueled with stable uranium compounds in a reflected cavity. The spectral characteristics of neutrons in GCR systems could shift from one end of the spectrum to the other end by changing design parameters such as reflector material and thickness, uranium enrichment, and the average operational temperature and pressure. The rate of actinide generation, transmutation, and burnup is highly influenced by the average neutron energy in reactor core. In particular, the production rate and isotopic mix of plutonium are highly dependent on the neutron spectrum in the reactor. Other actinides of primary interest to this work are neptunium-237 and americium-241 due to their pivotal impact on high-level nuclear waste disposal. In all cavity reactors including GCR's, the reflector material and thickness are the most important design parameters in determining the core spectrum. The increase in the gaseous fuel pressure and enrichment results in relative shift of neutron population toward energies greater than 2 eV. Reflector materials considered in this study are beryllium oxide, lithium hydride, lithium deuteride, zirconium carbide, graphite, lead, and tungsten. Results of the study suggest that the beryllium oxide and tungsten reflected GCR systems set the lower (softest) and upper (hardest) limits of neutron spectra, respectively. The inventory of actinides with half-lives greater than 1000 years can be minimized by increasing neutron flux level in the reactor core. The higher the neutron flux, the lower the inventory of these actinides. The majority of the GCR designs maintained a flux level on the order of 10{sup 15} cm{sup -2}*s{sup -1} while the PWR flux is one order of magnitude lower. The inventory of the feeder isotopes to Np{sup 237} including U{sup 237}, Pu{sup 241}, and Am{sup 241} decreases with relative shift of neutron spectrum toward higher energies. This is due to increased resonance absorption in these isotopes due to higher

Chronology of the beryllium replacement shutdown at the High Flux Isotope Reactor (HFIR), 1983

International Nuclear Information System (INIS)

Kohring, M.W.

1984-04-01

In addition to the permanent beryllium reflector, several other components were replaced. The outer shroud and lower tracks were replaced. The new control rod access plugs and the upper tracks were installed. Replacement of collimator tubes for HB-1 and -2 are tentatively slated for the next permanent beryllium changeout. Inspection of the reactor vessel, the vessel-to-nozzle welds, core support structure, and vessel internal cladding showed them to be in acceptable condition. The highest, accumulative radiation doses received by Reactor Operations personnel during the shutdown, in mrem, were 665, 606, and 560; the highest for P and E personnel were 520, 505, and 475

Design and Thermal Analysis for Irradiation of Pyrolytic Carbon/Silicon Carbide Diffusion Couples in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Gerczak, Tyler J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

Tristructural-isotropic (TRISO)–coated particle fuel is a promising advanced fuel concept consisting of a spherical fuel kernel made of uranium oxide and uranium carbide, surrounded by a porous carbonaceous buffer layer and successive layers of dense inner pyrolytic carbon (IPyC), silicon carbide (SiC) deposited by chemical vapor , and dense outer pyrolytic carbon (OPyC). This fuel concept is being considered for advanced reactor applications such as high temperature gas-cooled reactors (HTGRs) and molten salt reactors (MSRs), as well as for accident-tolerant fuel for light water reactors (LWRs). Development and implementation of TRISO fuel for these reactor concepts support the US Department of Energy (DOE) Office of Nuclear Energy mission to promote safe, reliable nuclear energy that is sustainable and environmentally friendly. During operation, the SiC layer serves as the primary barrier to metallic fission products and actinides not retained in the kernel. It has been observed that certain fission products are released from TRISO fuel during operation, notably, Ag, Eu, and Sr [1]. Release of these radioisotopes causes safety and maintenance concerns.

Development of high flux thermal neutron generator for neutron activation analysis

Energy Technology Data Exchange (ETDEWEB)

Vainionpaa, Jaakko H., E-mail: hannes@adelphitech.com [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Chen, Allan X.; Piestrup, Melvin A.; Gary, Charles K. [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Jones, Glenn [G& J Jones Enterprice, 7486 Brighton Ct, Dublin, CA 94568 (United States); Pantell, Richard H. [Department of Electrical Engineering, Stanford University, Stanford, CA (United States)

2015-05-01

The new model DD110MB neutron generator from Adelphi Technology produces thermal (<0.5 eV) neutron flux that is normally achieved in a nuclear reactor or larger accelerator based systems. Thermal neutron fluxes of 3–5 · 10{sup 7} n/cm{sup 2}/s are measured. This flux is achieved using four ion beams arranged concentrically around a target chamber containing a compact moderator with a central sample cylinder. Fast neutron yield of ∼2 · 10{sup 10} n/s is created at the titanium surface of the target chamber. The thickness and material of the moderator is selected to maximize the thermal neutron flux at the center. The 2.5 MeV neutrons are quickly thermalized to energies below 0.5 eV and concentrated at the sample cylinder. The maximum flux of thermal neutrons at the target is achieved when approximately half of the neutrons at the sample area are thermalized. In this paper we present simulation results used to characterize performance of the neutron generator. The neutron flux can be used for neutron activation analysis (NAA) prompt gamma neutron activation analysis (PGNAA) for determining the concentrations of elements in many materials. Another envisioned use of the generator is production of radioactive isotopes. DD110MB is small enough for modest-sized laboratories and universities. Compared to nuclear reactors the DD110MB produces comparable thermal flux but provides reduced administrative and safety requirements and it can be run in pulsed mode, which is beneficial in many neutron activation techniques.

High flux transmutation of fission products and actinides

International Nuclear Information System (INIS)

Gerasimov, A.; Kiselev, G.; Myrtsymova, L.

2001-01-01

Long-lived fission products and minor actinides accumulated in spent nuclear fuel of power reactors comprise the major part of high level radwaste. Their incineration is important from the point of view of radwaste management. Transmutation of these nuclides by means of neutron irradiation can be performed either in conventional nuclear reactors, or in specialized transmutation reactors, or in ADS facilities with subcritical reactor and neutron source with application of proton accelerator. Different types of transmutation nuclear facilities can be used in order to insure optimal incineration conditions for radwaste. The choice of facility type for optimal transmutation should be based on the fundamental data in the physics of nuclide transformations. Transmutation of minor actinides leads to the increase of radiotoxicity during irradiation. It takes significant time compared to the lifetime of reactor facility to achieve equilibrium without effective transmutation. High flux nuclear facilities allow to minimize these draw-backs of conventional facilities with both thermal and fast neutron spectrum. They provide fast approach to equilibrium and low level of equilibrium mass and radiotoxicity of transmuted actinides. High flux facilities are advantageous also for transmutation of long-lived fission products as they provide short incineration time

Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

Energy Technology Data Exchange (ETDEWEB)

Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2016-11-15

Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

International Nuclear Information System (INIS)

Rosenthal, Murray Wilford

2009-01-01

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

The importance of using the mixed neutron flux in activation analysis of D-3He fueled reactors

International Nuclear Information System (INIS)

Khater, H.Y.; Sawan, M.E.

1992-01-01

This paper reports on the D-D and D-T secondary reactions in D- 3 He reactors which provide the neutron source term for most of the radioactivity produced in the structure of the reactor. radionuclides are produced as a result of neutron interactions with their parent nuclides. The amount of activity produced by any radionuclide depends on the number of its parent atoms present at any given time. One approach to account for the activity induced by both neutron sources in any activation analysis is to add their individual contributions. Performing two separate calculations for the D-D and D-T neutron flux components and adding their contributions yields conservative results due to underestimating the destruction of the parent atoms. The overestimation is more pronounced for short and intermediate lived nuclides, long operation time, large neutron flux and large destruction cross section for the parent atoms. In the steel first wall of a typical d- 3 He reactor, adding the individual contributions of the tow neutron sources results in overestimating the activities produced by most of the radioactive isotopes of Ag, Lu, Ta, W and Re. After 30 years of reactor operation, the activity of 187 W, which is a major source of safety concern in case of an accident, is more than an order of magnitude higher than its value if the mixed neutron flux is used. The activity of 188 Re, which is an important source of offsite does in case of accidental release, is overestimated by more than a factor of two

A proposed standard on medical isotope production in fission reactors

International Nuclear Information System (INIS)

Schenter, R. E.; Brown, G. J.; Holden, C. S.

2006-01-01

Authors Robert E. Sehenter, Garry Brown and Charles S. Holden argue that a Standard for 'Medical Isotope Production' is needed. Medical isotopes are becoming major components of application for the diagnosis and treatment of all the major diseases including all forms of cancer, heart disease, arthritis, Alzheimer's, among others. Current nuclear data to perform calculations is incomplete, dated or imprecise or otherwise flawed for many isotopes that could have significant applications in medicine. Improved data files will assist computational analyses to design means and methods for improved isotope production techniques in the fission reactor systems. Initial focus of the Standard is expected to be on neutron cross section and branching data for both fast and thermal reactor systems. Evaluated and reviewed tables giving thermal capture cross sections and resonance integrals for the major target and product medical isotopes would be the expected 'first start' for the 'Standard Working Group'. (authors)

Characteristics and uses of a 250 kW TRIGA reactor

International Nuclear Information System (INIS)

Dimic, V.

1985-01-01

The 250 kW TRIGA Mark II reactor is a light water reactor with solid fuel elements in which the zirconium hydride moderator is homogeneously distributed between enriched uranium. Therefore the reactor has the large prompt negative temperature coefficient of reactivity, the fuel also has very high retention of radioactive fission products. The reactor core is a cylindrical configuration with an annular graphite reflector. The experimental facilities include a rotary specimen rack, a central incore radiation thimble, a pneumatic transfer system, and pulsing capability. Other experimental facilities include two radial and two tangential beam tubes, a graphite thermal column, and a graphite thermalizing column. At the steady state power of 250 kW the peak flux is 1x10 13 n/cm 2 s in the central test position. In addition, pulsing to about 2000 MW is usually provided giving peak fluxes of about 2x10 16 n/cm 2 sec. All TRIGA reactors produce a core-average thermal neutron flux of about 10 7 n.v per watt. Only with very large accelerators could such a high neutron flux be achieved. In order to give an appreciation for the research conducted at research reactors, the types of research could be summarized as follows: thermal neutron scattering, neutron radiography, neutron and nuclear physics, activation analysis, radiochemistry, biology and medicine, and teaching and training. Typical applied research with a 250 kW reactor has been conducted in medicine in biology, archeology, metallurgy and materials science, engineering and criminology. It is well known that research reactors have been used routinely to produce isotopes for industry and medicine. In some instances, reactors are the preferred method of isotope production. We can conclude that the 250 kW TRIGA research reactor is a useful and wide ranging source of radiation for basic and applied research. The operation cost for this instrument is relatively low. (author)

Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

Energy Technology Data Exchange (ETDEWEB)

Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

1994-12-31

The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

Three new nondestructive evaluation tools based on high flux neutron sources

International Nuclear Information System (INIS)

Hubbard, C.R.; Raine, D.; Peascoe, R.; Wright, M.

1997-01-01

Nondestructive evaluation methods and systems based on specific attributes of neutron interactions with materials are being developed. The special attributes of neutrons are low attenuation in most engineering materials, strong interaction with low Z elements, and epithermal neutron absorption resonances. The three methods under development at ORNL include neutron based tomography and radiography; through thickness, nondestructive texture mapping; and internal, noninvasive temperature measurement. All three techniques require high flux sources such as the High Flux Isotope Reactor, a steady state source, or the Oak Ridge Electron Linear Accelerator, a pulsed neutron source. Neutrons are quite penetrating in most engineering materials and thus can be useful to detect internal flaws and features. Hydrogen atoms, such as in a hydrocarbon fuel, lubricant, or a metal hydride, are relatively opaque to neutron transmission and thus neutron based tomography/radiography is ideal to image their presence. Texture, the nonrandom orientation of crystalline grains within materials, can be mapped nondestructively using neutron diffraction methods. Epithermal neutron resonance absorption is being studied as a noncontacting temperature sensor. This paper highlights the underlying physics of the methods, progress in development, and the potential benefits for science and industry of the three facilities

Neutron flux enhancement in the NRAD reactor

International Nuclear Information System (INIS)

Weeks, A.A.; Heidel, C.C.; Imel, G.R.

1988-01-01

In 1987 a series of experiments were conducted at the NRAD reactor facility at Argonne National Laboratory - West (ANL-W) to investigate the possibility of increasing the thermal neutron content at the end of the reactor's east beam tube through the use of hydrogenous flux traps. It was desired to increase the thermal flux for a series of experiments to be performed in the east radiography cell, in which the enhanced flux was required in a relatively small volume. Hence, it was feasible to attempt to focus the cross section of the beam to a smaller area. Two flux traps were constructed from unborated polypropylene and tested to determine their effectiveness. Both traps were open to the entire cross-sectional area of the neutron beam (as it emerges from the wall and enters the beam room). The sides then converged such that at the end of the trap the beam would be 'focused' to a greater intensity. The differences in the two flux traps were primarily in length, and hence angle to the beam as the inlet and outlet cross-sectional areas were held constant. It should be noted that merely placing a slab of polypropylene in the beam will not yield significant multiplication as neutrons are primarily scattered away

Core management, operational limits and conditions and safety aspects of the Australian High Flux Reactor (HIFAR)

International Nuclear Information System (INIS)

Town, S.L.

1997-01-01

HIFAR is a DIDO class reactor which commenced routine operation at approximately 10 MW in 1960. It is principally used for production of medical radio-isotopes, scientific research using neutron scattering facilities and irradiation of silicon ingots for the electronics industry. A detailed description of the core, including fuel types, is presented. Details are given of the current fuel management program HIFUEL and the experimental measurements associated with reactor physics analysis of HIFAR are discussed. (author)

Stable water isotope and surface heat flux simulation using ISOLSM: Evaluation against in-situ measurements

KAUST Repository

Cai, Mick Y.; Wang, Lixin; Parkes, Stephen; Strauss, Josiah; McCabe, Matthew; Evans, Jason P.; Griffiths, Alan D.

2015-01-01

The stable isotopes of water are useful tracers of water sources and hydrological processes. Stable water isotope-enabled land surface modeling is a relatively new approach for characterizing the hydrological cycle, providing spatial and temporal variability for a number of hydrological processes. At the land surface, the integration of stable water isotopes with other meteorological measurements can assist in constraining surface heat flux estimates and discriminate between evaporation (E) and transpiration (T). However, research in this area has traditionally been limited by a lack of continuous in-situ isotopic observations. Here, the National Centre for Atmospheric Research stable isotope-enabled Land Surface Model (ISOLSM) is used to simulate the water and energy fluxes and stable water isotope variations. The model was run for a period of one month with meteorological data collected from a coastal sub-tropical site near Sydney, Australia. The modeled energy fluxes (latent heat and sensible heat) agreed reasonably well with eddy covariance observations, indicating that ISOLSM has the capacity to reproduce observed flux behavior. Comparison of modeled isotopic compositions of evapotranspiration (ET) against in-situ Fourier Transform Infrared spectroscopy (FTIR) measured bulk water vapor isotopic data (10. m above the ground), however, showed differences in magnitude and temporal patterns. The disparity is due to a small contribution from local ET fluxes to atmospheric boundary layer water vapor (~1% based on calculations using ideal gas law) relative to that advected from the ocean for this particular site. Using ISOLSM simulation, the ET was partitioned into E and T with 70% being T. We also identified that soil water from different soil layers affected T and E differently based on the simulated soil isotopic patterns, which reflects the internal working of ISOLSM. These results highlighted the capacity of using the isotope-enabled models to discriminate

Stable water isotope and surface heat flux simulation using ISOLSM: Evaluation against in-situ measurements

KAUST Repository

Cai, Mick Y.

2015-04-01

The stable isotopes of water are useful tracers of water sources and hydrological processes. Stable water isotope-enabled land surface modeling is a relatively new approach for characterizing the hydrological cycle, providing spatial and temporal variability for a number of hydrological processes. At the land surface, the integration of stable water isotopes with other meteorological measurements can assist in constraining surface heat flux estimates and discriminate between evaporation (E) and transpiration (T). However, research in this area has traditionally been limited by a lack of continuous in-situ isotopic observations. Here, the National Centre for Atmospheric Research stable isotope-enabled Land Surface Model (ISOLSM) is used to simulate the water and energy fluxes and stable water isotope variations. The model was run for a period of one month with meteorological data collected from a coastal sub-tropical site near Sydney, Australia. The modeled energy fluxes (latent heat and sensible heat) agreed reasonably well with eddy covariance observations, indicating that ISOLSM has the capacity to reproduce observed flux behavior. Comparison of modeled isotopic compositions of evapotranspiration (ET) against in-situ Fourier Transform Infrared spectroscopy (FTIR) measured bulk water vapor isotopic data (10. m above the ground), however, showed differences in magnitude and temporal patterns. The disparity is due to a small contribution from local ET fluxes to atmospheric boundary layer water vapor (~1% based on calculations using ideal gas law) relative to that advected from the ocean for this particular site. Using ISOLSM simulation, the ET was partitioned into E and T with 70% being T. We also identified that soil water from different soil layers affected T and E differently based on the simulated soil isotopic patterns, which reflects the internal working of ISOLSM. These results highlighted the capacity of using the isotope-enabled models to discriminate

Elaboration of mini plates with U-Mo for irradiation in a high flux reactor

International Nuclear Information System (INIS)

Pasqualini, Enrique E.

2005-01-01

Full text: International new efforts for the reconversion of HEU in research, testing and radioisotopes production reactors, have greatly incremented U-Mo fuels qualification activities. These qualifications require the resolution of undesired interaction at high fluxes between UMo particles and the aluminum matrix in the case of dispersed fuels and the development of U-Mo monolithic fuels. These efforts are being manifested in the planning and execution of additional series of irradiation tests of mini plates and full size plates. Recently, CNEA has elaborated mini plates with different proposals for the irradiation at the ATR reactor (250 MWTH, maximum thermal neutron flux 10 15 n.cm -2 .seg -1 ) at Idaho National Laboratory, USA. Uranium 7% (w/w) molybdenum (U-7Mo) particles were coated with silicon. Chemical vapour deposition (CVD) of silane and high temperature diffusion of silicon were used. Hydrided, milled and dehydrated (HMD) particles heat treated at 1000 C degrees during four hours and centrifugal atomized powder were coated and the results compared. Mini plates were elaborated with both kinds of particles. Mini plates were also elaborated with U-7Mo and silicon particles dispersed in the aluminium matrix. Monolithic mini plates were also developed by co lamination of U-7Mo with a Zircaloy-4 cladding. The different steps of this process are detailed and the method is shown to be versatile, can be easily scaled up and is performed with small modifications of usual equipment in fuel plants. The irradiation experiment is called RERTR-7A, includes a total of 32 mini plates and it is planed to finalize by mid 2006. (author) [es

Maximum neutron flux in thermal reactors

International Nuclear Information System (INIS)

Strugar, P.V.

1968-12-01

Direct approach to the problem is to calculate spatial distribution of fuel concentration if the reactor core directly using the condition of maximum neutron flux and comply with thermal limitations. This paper proved that the problem can be solved by applying the variational calculus, i.e. by using the maximum principle of Pontryagin. Mathematical model of reactor core is based on the two-group neutron diffusion theory with some simplifications which make it appropriate from maximum principle point of view. Here applied theory of maximum principle are suitable for application. The solution of optimum distribution of fuel concentration in the reactor core is obtained in explicit analytical form. The reactor critical dimensions are roots of a system of nonlinear equations and verification of optimum conditions can be done only for specific examples

High Flux Isotope Reactor technical specifications

International Nuclear Information System (INIS)

1977-03-01

Included are sections covering safety limits and limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. An appendix deals with accidents and anticipated transients

The effective management of medical isotope production in research reactors

International Nuclear Information System (INIS)

Drummond, D.T.

1993-01-01

During the 50-yr history of the use of radioisotopes for medical applications, research reactors have played a pivotal role in the production of many if not most of the key products. The marriage between research reactors and production operations is subject to significant challenges on two fronts. The medical applications of the radioisotope products impose some unique constraints and requirements on the production process. In addition, the mandates and priorities of a research reactor are not always congruent with the demands of a production environment. This paper briefly reviews the historical development of medical isotope production, identifies the unique challenges facing this endeavor, and discusses the management of the relationship between the isotope producer and the research reactor operator. Finally, the key elements of a successful relationship are identified

Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor.

Science.gov (United States)

Holden, Norman E; Reciniello, Richard N; Hu, Jih-Perng

2004-08-01

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions.

Determination of neutron flux densities in WWR-S reactor core

International Nuclear Information System (INIS)

Tomasek, F.

1989-04-01

The method is described of determining neutron flux densities and neutron fluences using activation detectors. The basic definitions and relations for determining reaction rates, fluence and neutron flux as well as the characteristics of some reactions and of sitable activation detectors are reported. The flux densities were determined of thermal and fast neutrons and of gamma quanta in the WWR-S reactor core. The data measured in the period 1984-1987 are tabulated. Cross sections for the individual reactions were determined from spectra measurements processed using program SAND-II and cross section library ENDF-B IV. Neutron flux densities were also measured for the WWR-S reactor vertical channels. (E.J.). 10 figs., 8 tabs., 111 refs

Measurement of the thermal flux distribution in the IEA-R1 reactor

International Nuclear Information System (INIS)

Tangari, C.M.; Moreira, J.M.L.; Jerez, R.

1986-01-01

The knowledge of the neutron flux distribution in research reactors is important because it gives the power distribution over the core, and it provides better conditions to perform experiments and sample irradiations. The measured neutron flux distribution can also be of interest as a means of comparison for the calculational methods of reactor analysis currently in use at this institute. The thermal neutron flux distribution of the IEA-R1 reactor has been measured with the miniature chamber WL-23292. For carrying out the measurements, it was buit a guide system that permit the insertion of the mini-chamber i between the fuel of the fuel elements. It can be introduced in two diferent positions a fuel element and in each it spans 26 axial positions. With this guide system the thermal neutron flux distribution of the IEA-R1 nuclear reactor can be obtained in a fast and efficient manner. The element measured flux distribution shows clearly the effects of control rods and reflectors in the IEA-R1 reactor. The difficulties encountered during the measurements are mentioned with detail as well as the procedures adopteed to overcome them. (Author) [pt

A computationally simple model for determining the time dependent spectral neutron flux in a nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Schneider, E.A. [Department of Mechanical Engineering, University of Texas, Austin, TX (United States); Deinert, M.R. [Theoretical and Applied Mechanics, Cornell University, 219 Kimball Hall, Ithaca, NY 14853 (United States)]. E-mail: mrd6@cornell.edu; Cady, K.B. [Theoretical and Applied Mechanics, Cornell University, 219 Kimball Hall, Ithaca, NY 14853 (United States)

2006-10-15

The balance of isotopes in a nuclear reactor core is key to understanding the overall performance of a given fuel cycle. This balance is in turn most strongly affected by the time and energy-dependent neutron flux. While many large and involved computer packages exist for determining this spectrum, a simplified approach amenable to rapid computation is missing from the literature. We present such a model, which accepts as inputs the fuel element/moderator geometry and composition, reactor geometry, fuel residence time and target burnup and we compare it to OECD/NEA benchmarks for homogeneous MOX and UOX LWR cores. Collision probability approximations to the neutron transport equation are used to decouple the spatial and energy variables. The lethargy dependent neutron flux, governed by coupled integral equations for the fuel and moderator/coolant regions is treated by multigroup thermalization methods, and the transport of neutrons through space is modeled by fuel to moderator transport and escape probabilities. Reactivity control is achieved through use of a burnable poison or adjustable control medium. The model calculates the buildup of 24 actinides, as well as fission products, along with the lethargy dependent neutron flux and the results of several simulations are compared with benchmarked standards.

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Rosenthal, Murray Wilford [ORNL

2009-08-01

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Determination flux in the Reactor JEN-1

International Nuclear Information System (INIS)

Manas Diaz, L.; Montes Ponce de leon, J.

1960-01-01

This report summarized several irradiations that have been made to determine the neutron flux distributions in the core of the JEN-1 reactor. Gold foils of 380 μ gr and Mn-Ni (12% de Ni) of 30 mg have been employed. the epithermal flux has been determined by mean of the Cd radio. The resonance integral values given by Macklin and Pomerance have been used. (Author) 9 refs

Safety assessment of Department of Energy nuclear reactors

International Nuclear Information System (INIS)

1981-03-01

One of the first tasks of the NFPQT Committee was to determine which DOE reactors would be assessed. The Committee determined that in view of the limited time available to conduct the assessment, 13 DOE reactors were of such size (physical, power or fission product inventory) to warrant review. This determination was approved by the Under Secretary. A decision was also made in the cases of three weapons material production reactors, C, K and P, to concentrate on the K reactor only, since all three are of the same basic design, have the same operating features, are all at the same site, and are all operated by the same contractor. The assessment was accomplished in the following ways: reviewing the results of assessments conducted by the DOE organizations with reactor safety responsibilities, which were undertaken in compliance with the request of the various program directors; reviewing selected documents that were requested by the Committee and assembled at DOE Headquarters; interviewing DOE Headquarters and Field Office personnel; and conducting on-site reviews of four reactors located at four different sites. The four reactors for on-site reviews were: Advanced Test Reactor (ATR); K Production Reactor; High Flux Beam Reactor (HFBR); and High Flux Isotope Reactor (HFIR). Specific findings and recommendations from the assessment are presented

The reactor and the production of isotopes

International Nuclear Information System (INIS)

Hevesy, G. de

1962-01-01

The construction of the cyclotron immensely advanced the availability of radioactive tracers, a few of which even today can be produced only with the aid of this device. But even this great advance was overshadowed by the fabulous production of isotopes by the reactors. Isotopes of almost any element and of almost unlimited activity became available. It now became possible to apply H 3 - discovered already in the 'thirties by Rutherford and Oliphant - and C 14 , and these were used in thousands of investigations

Calculation with MCNP of capture photon flux in VVER-1000 experimental reactor.

Science.gov (United States)

Töre, Candan; Ortego, Pedro

2005-01-01

The aim of this study is to obtain by Monte Carlo method the high energy photon flux due to neutron capture in the internals and vessel layers of the experimental reactor LR-0 located in REZ, Czech Republic, and loaded with VVER-1000 fuel. The calclated neutron, photon and photon to neutron flux ratio are compared with experimental measurements performed with a multi-parameter stilbene detector. The results show clear underestimation of photon flux in downcomer and some overestimation at vessel surface and 1/4 thickness but a good fitting for deeper points in vessel.

Generating the flux map of Nigeria Research Reactor-1 for efficient ...

African Journals Online (AJOL)

One of the main uses to which the Nigeria Research Reactor-1 (NIRR-1) will be put is neutron activation analysis. The activation analyst requires information about the flux level at various points within and around the reactor core to enable him identify the point of optimum flux (at a given operating power) for any irradiation ...

Near-Continuous Isotopic Characterization of Soil N2O Fluxes from Maize Production

Science.gov (United States)

Anex, R. P.; Francis Clar, J.

2015-12-01

Isotopomer ratios of N2O and especially intramolecular 15N site preference (SP) have been proposed as indicators of the sources of N2O and for providing insight into the contributions of different microbial processes. Current knowledge, however, is mainly based on pure culture studies and laboratory flask studies using mass spectrometric analysis. Recent development of laser spectroscopic methods has made possible high-precision, in situ measurements. We present results from a maize production field in Columbia County, Wisconsin, USA. Data were collected from the fertilized maize phase of a maize-soybean rotation. N2O mole fractions and isotopic composition were determined using an automatic gas flux measurement system comprising a set of custom-designed automatic chambers, circulating gas paths and an OA-ICOS N2O Isotope Analyzer (Los Gatos Research, Inc., Model 914-0027). The instrument system allows for up to 15 user programmable soil gas chambers. Wide dynamic range and parts-per-billion precision of OA-ICOS laser absorption instrument allows for extremely rapid estimation of N2O fluxes. Current operational settings provide measurements of N2O and its isotopes every 20 seconds with a precision of 0.1 ± 0.050 PPB. Comparison of measurements from four chambers (two between row and two in-row) show very different aggregate N2O flux, but SP values suggest similar sources from nitrifier denitrification and incomplete bacterial denitrification. SP values reported are being measured throughout the current growing season. To date, the majority of values are consistent with an origin from bacterial denitrification and coincide with periods of high water filled pore space.

A high-speed data acquisition system to measure low-level current from self-powered flux detectors in CANDU nuclear reactors

International Nuclear Information System (INIS)

Lawrence, C.B.; Hall, D.S.

1982-05-01

Self-powered flux detectors are used in CANDU nuclear power reactors to determine the spatial neutron flux distribution in the reactor core for use by both the reactor control and safety systems. To establish the dynamic response of different types of flux detectors, the Chalk River Nuclear Laboratories have an ongoing experimental irradiation program in the NRU research reactor for which a data acquistion system has been developed. The system described in this paper is used to measure the currents from the detectors both at a slow, regular logging interval, and at a rapid, adaptive rate following a reactor shutdown. Currents that range from 100 pA to 1 mA full scale can be measured from up to 38 detectors and stored at sampling rates of up to 20 samples per second. The dynamic characteristics of the detectors can be computed from the stored records. The data acquisition system comprises a DEC LSI-11/23 microcomputer, dual cartridge disks, floppy disks, a hard copy and a video display terminal. The RT-11 operating system is used and all application programs are written in FORTRAN

Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell.

Modelling deuterium release from tungsten after high flux high temperature deuterium plasma exposure

Energy Technology Data Exchange (ETDEWEB)

Grigorev, Petr, E-mail: grigorievpit@gmail.com [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Ghent University, Applied Physics EA17 FUSION-DC, St. Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Department of Experimental Nuclear Physics K-89, Institute of Physics, Nanotechnologies, and Telecommunications, Peter the Great St. Petersburg Polytechnic University, St. Petersburg (Russian Federation); Matveev, Dmitry [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425, Jülich (Germany); Bakaeva, Anastasiia [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Department of Applied Physics, Ghent University (Belgium); Terentyev, Dmitry [SCK-CEN, Nuclear Materials Science Institute, Boeretang 200, Mol, 2400 (Belgium); Zhurkin, Evgeny E. [Department of Experimental Nuclear Physics K-89, Institute of Physics, Nanotechnologies, and Telecommunications, Peter the Great St. Petersburg Polytechnic University, St. Petersburg (Russian Federation); Van Oost, Guido [Ghent University, Applied Physics EA17 FUSION-DC, St. Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Noterdaeme, Jean-Marie [Ghent University, Applied Physics EA17 FUSION-DC, St. Pietersnieuwstraat, 41 B4, B-9000, Gent (Belgium); Max-Planck-Institut für Plasmaphysik, Garching (Germany)

2016-12-01

Tungsten is a primary candidate for plasma facing materials for future fusion devices. An important safety concern in the design of plasma facing components is the retention of hydrogen isotopes. Available experimental data is vast and scattered, and a consistent physical model of retention of hydrogen isotopes in tungsten is still missing. In this work we propose a model of non-equilibrium hydrogen isotopes trapping under fusion relevant plasma exposure conditions. The model is coupled to a diffusion-trapping simulation tool and is used to interpret recent experiments involving high plasma flux exposures. From the computational analysis performed, it is concluded that high flux high temperature exposures (T = 1000 K, flux = 10{sup 24} D/m{sup 2}/s and fluence of 10{sup 26} D/m{sup 2}) result in generation of sub-surface damage and bulk diffusion, so that the retention is driven by both sub-surface plasma-induced defects (bubbles) and trapping at natural defects. On the basis of the non-equilibrium trapping model we have estimated the amount of H stored in the sub-surface region to be ∼10{sup −5} at{sup −1}, while the bulk retention is about 4 × 10{sup −7} at{sup −1}, calculated by assuming the sub-surface layer thickness of about 10 μm and adjusting the trap concentration to comply with the experimental results for the integral retention.

Type B investigation of the iridium contamination event at the High Flux Isotope Reactor on September 7, 1993

International Nuclear Information System (INIS)

1994-03-01

On the title date, at ORNL, area radiation alarms sounded during a routine transfer of a shielding cask (containing 60 Ci 192 Ir) from the HFIR pool side to a transport truck. Small amounts of Ir were released from the cask onto the reactor bay floor. The floor was cleaned, and the cask was shipped to a hot cell at Building 3047 on Oct. 3, 1993. The event was caused by rupture of one of the Ir target rods after it was loaded into the cask for normal transport operations; the rupture was the result of steam generation in the target rod soon after it was placed in the cask (water had entered the target rod through a tiny defect in a weld while it was in the reactor under pressure). While the target rods were in the reactor and reactor pool, there was sufficient cooling to prevent steam generation; when the target rod was loaded into the dry transport cask, the temperature increased enough to result in boiling of the trapped water and produced high enough pressure to result in rupture. The escaping steam ejected some of the Ir pellets. The event was reported as Occurrence Report Number ORO--MMES-X10HFIR-1993-0030, dated Sept. 8, 1993. Analysis indicated that the following conditions were probable causes: less than adequate welding procedures, practices, or techniques, material controls, or inspection methods, or combination thereof, could have led to weld defects, affecting the integrity of target rod IR-75; less than adequate secondary containment in the cask allowed Ir pellets to escape

Anti-neutrino flux in a research reactor for non-proliferation application

Energy Technology Data Exchange (ETDEWEB)

Khakshournia, Samad; Foroughi, Shokoufeh [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Atomic Energy Organization of Iran (AEOI)

2017-11-15

Owing to growing interest in the study of emitted antineutrinos from nuclear reactors to test the Atomic Energy Agency safeguards, antineutrino flux was studied in the Tehran Research Reactor (TRR) using ORIGEN code. According to our prediction, antineutrino rate was obtained 2.6 x 10{sup 17} (v{sub e}/sec) in the core No. 57F of the TRR. Calculations indicated that evolution of antineutrino flux was very slow with time and the performed refueling had not an observable effect on antineutrino flux curve for a 5 MW reactor with the conventional refueling program. It is seen that for non-proliferation applications the measurement of the contribution of {sup 239}Pu to the fission using an antineutrino detector is not viable in the TRR.

The epithermal neutron-flux distribution in the reactor RA - Vinca

International Nuclear Information System (INIS)

Marinkov, V.; Bikit, I.; Martinc, R.; Veskovic, M.; Slivka, J.; Vaderna, S.

1987-01-01

The distribution of the epithermal neutron flux in the reactor RA - Vinca has been measured by means of Zr - activation detectors. In the channel VK-8 non-homogeneous flux distribution was observed (author) [sr

Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

International Nuclear Information System (INIS)

Nguyen Hoang Hai; Do Quang Binh

2011-01-01

Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

CO and H2 uptake and emissions by soil: variability of fluxes and their isotopic signatures

Science.gov (United States)

Popa, Maria Elena; Chen, Qianjie; Ferrero Lopez, Noelia; Röckmann, Thomas

2017-04-01

In order to study the uptake and release of H2 and CO by soil, we performed long term, high frequency measurements with an automatic soil chamber at two sites in the Netherlands (Cabauw - grassland, and Speuld - forest). The measurements were performed over different seasons and cover in total a cumulated interval of about one year. These measurements allow determining separately, for each species, the two distinct fluxes i.e. uptake and release, and investigating their temporal variability and dependencies on environmental variables. Additional experiments were performed for determining the isotopic signatures of the H2 and CO uptake and release by soil. Flask samples were filled from the soil chamber, and then analyzed in the laboratory for the stable isotopic composition of H2 (δD) and CO (δ13C and δ18O). We find that both uptake and release are present at all times, regardless of the direction of the net flux. The emissions are significant for both species and at Cabauw, there are times and places where emissions outweigh the soil uptake. For each species, the two fluxes have different behavior and dependence on external variables, which indicates that they have different origins. The isotope results also support that, for both H2 and CO, uptake and emission occur simultaneously. We were able to determine separately the isotopic effects of the two fluxes. For both H2 and CO, soil uptake is associated with a small positive fractionation (the lighter molecule is taken up faster). The soil uptake fractionation (α = kheavy/klight) was 0.945 ± 0.004 for H2; for CO, the fractionation was 0.992 for 13C and 0.985 for 18O. The isotopic composition of the H2 emitted from the grassland was -530 ± 40 ‰, less depleted that what is expected from the isotopic equilibrium of H2 with water. For CO, the isotopic composition of the soil emission is depleted in 13C compared to atmospheric CO, and lower than the average isotopic composition of plant or soil organic matter.

Analysis of Neutron Flux Distribution in Rsg-Gas Reactor With U-Mo Fuels

Directory of Open Access Journals (Sweden)

Taswanda Taryo

2004-01-01

Full Text Available The use of U-Mo fuels in research reactors seems to be promising and, recently, world researchers have carried out these such activities actively. The National Nuclear Energy Agency (BATAN which owns RSG-GAS reactor available in Serpong Research Center for Atomic Energy should anticipate this trend. It is, therefore, this research work on the use of U-Mo fuels in RSG-GAS reactor should be carried out. The work was focused on the analysis of neutron flux distribution in the RSG-GAS reactor using different content of molybdenum in U-Mo fuels. To begin with, RSG-GAS reactor core model was developed and simulated into X, Y and Z dimensions. Cross section of materials based on the developed cells of standard and control fuels was then generated using WIMS-D5-B. The criticality calculations were finally carried out applying BATAN-2DIFF code. The results showed that the neutron flux distribution obtained in U-Mo-fuel-based RSG-GAS core is very similar to those achieved in the 300-gram sillicide-fuel-based RSG-GAS reactor core. Indeed, the utilization of the U-Mo RSG-GAS core can be very similar to that of the high-density sillicide reactor core and even could be better in the future.

Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

International Nuclear Information System (INIS)

Ashoub, N.; Saleh, H.

1995-01-01

The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

Candidate processes for diluting the 235U isotope in weapons-capable highly enriched uranium

International Nuclear Information System (INIS)

Snider, J.D.

1996-02-01

The United States Department of Energy (DOE) is evaluating options for rendering its surplus inventories of highly enriched uranium (HEU) incapable of being used to produce nuclear weapons. Weapons-capable HEU was earlier produced by enriching uranium in the fissile 235 U isotope from its natural occurring 0.71 percent isotopic concentration to at least 20 percent isotopic concentration. Now, by diluting its concentration of the fissile 235 U isotope in a uranium blending process, the weapons capability of HEU can be eliminated in a manner that is reversible only through isotope enrichment, and therefore, highly resistant to proliferation. To the extent that can be economically and technically justified, the down-blended uranium product will be made suitable for use as commercial reactor fuel. Such down-blended uranium product can also be disposed of as waste if chemical or isotopic impurities preclude its use as reactor fuel

Studies on the instrumentation of a beam-tube medium flux reactor

International Nuclear Information System (INIS)

Axmann, A.; Pollet, J.L.; Queudot, J.

1979-01-01

In the years 1977/78, the ad hoc commitee for medium-flux reactor development of the Federal Ministry for Research and Technology developed constructional concepts for a medium-flux reactor to be utilized by beam tube experiments. The HMI has elaborated contributions for discussions of the subject of instrumentation, in particular for experiments in solid state physics. These contributions are contained in the report. (orig./RW) [de

Thermal flux flattering and increase of reactor output

Energy Technology Data Exchange (ETDEWEB)

Horowitz, J; Bussac, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1959-07-01

It is worthwhile, when building power reactors, to have excess reactivity in order to increase rating by fitting closely together the heat sources and the cooling possibilities. The power per unit volume of a graphite reactor can then be increased, given the power of the most heavily loaded channel. The solutions adopted for G.1, G.2, and E.D.F.1 are described here, and also the improvements based on the actual neutron flux flattening, the introduction of several zones for the coolant, the variation of uranium rod and coolant channel diameters according to their location, and finally the change in lattice pitch. The perturbation of neutron flux due to variation of mean absorption in the lattice is also discussed. (author)

Self-adjointness of the fast flux in a pressurized water reactor

International Nuclear Information System (INIS)

Mosteller, R.D.

1985-01-01

Most computer codes for the analysis of systems transients rely on a simplified representation of the active core, typically employing either a one-dimensional or a point kinetics model. The collapsing of neutronics data from multidimensional steady-state calculations normally employs flux/flux-adjoint weighting. The multidimensional calculations, however, usually are performed only for the forward problem, not the adjoint. The collapsing methodologies employed in generating the neutronics input for transient codes typically construct adjoint fluxes from the assumption that the fast flux is self-adjoint. Until now, no further verification of this assumption has been undertaken for thermal reactors. As part of the verification effort for EPRI's reactor analysis support package, the validity of this assumption now has been investigated for a modern pressurized water reactor (PWR). The PDQ-7 code was employed to perform two-group fine-mesh forward and adjoint calculations for a two-dimensional representation of Zion Unit 2 at beginning of life, based on the standard PWR ARMP model. It has been verified that the fast flux is very nearly self-adjoint in a PWR. However, a significant error can arise during the subsequent construction of the thermal adjoint flux unless allowance is made for the difference between the forward and adjoint thermal buckling terms. When such a difference is included, the thermal adjoint flux can be estimated very accurately

Characteristics of isotope-selective chemical reactor with gas-separating device

International Nuclear Information System (INIS)

Gorshunov, N.M.; Kalitin, S.A.; Laguntsov, N.I.; Neshchimenko, Yu.P.; Sulaberidze, G.A.

1988-01-01

A study was made on characteristics of separating stage, composed of isotope-selective chemical (or photochemical) reactor and membrane separating cascade (MSC), designated for separation of isotope-enriched products from lean reagents. MSC represents the counterflow cascade for separation of two-component mixtures. Calculations show that for the process of carton isotope separation the electric power expences for MSC operation are equal to 20 kWxh/g of CO 2 final product at 13 C isotope content in it equal to 75%. Application of the membrane gas-separating cascade at rather small electric power expenses enables to perform cascading of isotope separation in the course of nonequilibrium chemical reactions

Device for detecting neutron flux in nuclear reactor. [BWR

Energy Technology Data Exchange (ETDEWEB)

Bessho, Y; Nishizawa, Y

1976-07-30

The object of the invention is to ensure accuracy in the operation of the nuclear reactor by reducing the difference that results between the readings of a Traversing Incore Probe (TIP) and a Local Power Range Monitor (LPRM) when the neutron flux distribution undergoes a change. In an apparatus for detecting neutrons in a nuclear reactor, an LPRM sensor comprising a layer containing a substance capable of nuclear fission, a section filled with argon gas and a collector is constructed so as to surround a TIP within a TIP guide tube at the height of the reactor axis. In this way, the LPRM detects the average value of neutron distribution in the region surrounding the TIP, so that no great difference between the readings of both the sensors is produced even if the neutron flux distribution is changed.

TORT application in reactor pressure vessel neutron flux calculations

International Nuclear Information System (INIS)

Belousov, S.I.; Ilieva, K.D.; Antonov, S.Y.

1994-01-01

The neutron flux values onto reactor pressure vessel for WWER-1000 and WWER-440 reactors, at the places important for metal embrittlement surveillance have been calculated by 3 dimensional code TORT and synthesis method. The comparison of the results received by both methods confirms their good consistency. (authors). 13 refs., 4 tabs

Measuring neutron flux density in near-vessel space of a commercial WWER-1000 reactor

International Nuclear Information System (INIS)

Borodkin, G.I.; Eremin, A.N.; Lomakin, S.S.; Morozov, A.G.

1987-01-01

Distribution of neutron flux density in two experimental channels on the reactor vessel external surface and in ionization chamber channel of a commercial WWER-1000 reactor, is measured by the activation detector technique. Azimuthal distributions of fast and thermal neutron fluxes and height distributions of fast neutron flux density within energy range >1.2 and 2.3 MeV are obtained. Conclusion is made, that reactor core state and its structural peculiarities in the measurement range essentially affect space and energy distribution of neutron field near the vessel. It should be taken into account when determining permissible neutron fluence for the reactor vessel

Flux-limited diffusion coefficients in reactor physics applications

International Nuclear Information System (INIS)

Pounders, J.; Rahnema, F.; Szilard, R.

2007-01-01

Flux-limited diffusion theory has been successfully applied to problems in radiative transfer and radiation hydrodynamics, but its relevance to reactor physics has not yet been explored. The current investigation compares the performance of a flux-limited diffusion coefficient against the traditionally defined transport cross section. A one-dimensional BWR benchmark problem is examined at both the assembly and full-core level with varying degrees of heterogeneity. (authors)

Ten years operating experience at the Fast Flux Test Facility: A decade of excellence

International Nuclear Information System (INIS)

Swaim, D.J.; Waldo, J.B.; Farabee, O.A.

1991-07-01

The Fast Flux Test Facility is a 400 MW(t) fast reactor cooled by three sodium loops. The Fast Flux Test Facility is managed by the Westinghouse Hanford Company for the US Department of Energy. The Fast Flux Test Facility was designed and constructed to provide irradiation testing of fuels and materials for the US Department of Energy Liquid Metal Reactor research program. Facility activities have increased to include fusion power materials testing, passive safety testing, isotope production, and international collaboration. 5 figs

On flux effects in a low alloy steel from a Swedish reactor pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Boåsen, Magnus, E-mail: boasen@kth.se [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Efsing, Pål [Department of Solid Mechanics, Royal Institute of Technology (KTH), SE-100 44 Stockholm (Sweden); Ehrnstén, Ulla [VTT Technical Research Centre of Finland Ltd, PO Box 1000, FI-02044 VTT (Finland)

2017-02-15

This study aims to investigate the presence of Unstable Matrix Defects in irradiated pressure vessel steel from weldments of the Swedish PWR Ringhals 4 (R4). Hardness tests have been performed on low flux (surveillance material) and high flux (Halden reactor) irradiated material samples in combination with heat treatments at temperatures of 330, 360 and 390 °C in order to reveal eventual recovery of any hardening features induced by irradiation. The experiments carried out in this study could not reveal any hardness recovery related to Unstable Matrix Defects at relevant temperatures. However, a difference in hardness recovery was found between the low and the high flux samples at heat treatments at higher temperatures than expected for the annihilation of Unstable Matrix Defects–the observed recovery is here attributed to differences of the solute clusters formed by the high and low flux irradiations. - Highlights: • Hardness testing is combined with post irradiation annealing at 330, 360 and 390 °C. • Unstable matrix defects is studied in a reactor pressure vessel steel. • Comparison between surveillance material and accelerated irradiation. • No evidence of unstable matrix defects, i.e. not present in studied material. • Difference in hardness recovery between irradiation conditions found at 390 °C.

A new detector for the measurement of neutron flux in nuclear reactors

International Nuclear Information System (INIS)

Koch, L.; Labeyrie, J.; Tarassenko, S.

1958-01-01

The detector described is designed for the instantaneous measurement of thermal neutron fluxes, in the presence of high γ ray activity; this detector can withstand temperatures as high as 500 deg. C. It is based on the following principle: radioactive atoms resulting from heavy-nucleus fission are carried by a gas flow to a detector recording their β and γ disintegration. Thermal neutron fluxes as low as few neutrons per cm 2 per second can be measured. This detector may be used to control a nuclear reactor, to plot the thermal flux distribution with an excellent definition (1 mm 2 ) for fluxes higher than 10 8 n/cm 2 /s. The time response of the system to a sharp variation of flux is limited, in case of large fluxes, to the transit time of the gas flow between the fission product emitter and the detector; of the order of one tenth of a sec per meter of piping. The detector may also be applied for spectroscopy of fission products eider than 0,1 s. (author) [fr

Exospheric density and escape fluxes of atomic isotopes on Venus and Mars

International Nuclear Information System (INIS)

Wallis, M.K.

1978-01-01

Energetic neutrals in dissociative recombinations near or above the exobase provided an important component of exospheric density and escape fluxes. Plasma thermal velocities provide the main contribution to the velocity spread and an exact integral for the escape flux applicable in marginal cases is found for a simple atmosphere and collisional cut-off. Atomic fragments from recombination of diatomic oxygen and nitrogen ions in the Venus and Mars atmospheres are examined and density integrals derived. The oxygen escape flux on Mars is half that previously estimated and there is very little isotope preference supplementing diffusive separation. However, escape of the heavier 15 N isotope is low by a factor two. Reinterpretation of its 75% enrichment as detected by Viking leads to a range 0.4-1.4 mbar for the primeval nitrogen content on Mars. (author)

OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE IN TUNGSTEN SUBJECTED TO NEUTRON FLUX WITH PKA SPECTRUM CORRESPONDING TO THE HFIR

Energy Technology Data Exchange (ETDEWEB)

Nandipati, Giridhar; Setyawan, Wahyu; Heinisch, Howard L.; Roche, Kenneth J.; Kurtz, Richard J.; Wirth, Brian D.

2015-12-31

The objective of this work is to study the damage accumulation in pure tungsten (W) subjected to neutron bombardment with a primary knock-on atom (PKA) spectrum corresponding to the High Flux Isotope Reactor (HFIR), using the object kinetic Monte Carlo (OKMC) method.

Isotopic evidence for nitrous oxide production pathways in a partial nitritation-anammox reactor.

Science.gov (United States)

Harris, Eliza; Joss, Adriano; Emmenegger, Lukas; Kipf, Marco; Wolf, Benjamin; Mohn, Joachim; Wunderlin, Pascal

2015-10-15

Nitrous oxide (N2O) production pathways in a single stage, continuously fed partial nitritation-anammox reactor were investigated using online isotopic analysis of offgas N2O with quantum cascade laser absorption spectroscopy (QCLAS). N2O emissions increased when reactor operating conditions were not optimal, for example, high dissolved oxygen concentration. SP measurements indicated that the increase in N2O was due to enhanced nitrifier denitrification, generally related to nitrite build-up in the reactor. The results of this study confirm that process control via online N2O monitoring is an ideal method to detect imbalances in reactor operation and regulate aeration, to ensure optimal reactor conditions and minimise N2O emissions. Under normal operating conditions, the N2O isotopic site preference (SP) was much higher than expected - up to 40‰ - which could not be explained within the current understanding of N2O production pathways. Various targeted experiments were conducted to investigate the characteristics of N2O formation in the reactor. The high SP measurements during both normal operating and experimental conditions could potentially be explained by a number of hypotheses: i) unexpectedly strong heterotrophic N2O reduction, ii) unknown inorganic or anammox-associated N2O production pathway, iii) previous underestimation of SP fractionation during N2O production from NH2OH, or strong variations in SP from this pathway depending on reactor conditions. The second hypothesis - an unknown or incompletely characterised production pathway - was most consistent with results, however the other possibilities cannot be discounted. Further experiments are needed to distinguish between these hypotheses and fully resolve N2O production pathways in PN-anammox systems. Copyright © 2015 Elsevier Ltd. All rights reserved.

Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1961-12-15

This report includes the following separate parts: Thermal neutron flux in the experimental channels od RA reactor; Epithermal neutron flux in the experimental channels od RA reactor; Fast neutron flux in the experimental channels od RA reactor; Thermal neutron flux in the thermal column and biological experimental channel; Neutronic measurements in the RA reactor cell; Temperature reactivity coefficient of the RA reactor; design of the device for measuring the activity of wire. [Serbo-Croat] Ovaj izvestaj sadrzi sledece referate: Fluks termalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks epitermalnih neutrona u eksperimentalnim kanalima reaktora RA; Fluks brzih neutrona u eksperimentalnim kanalima reaktora RA; Fluks termalnih neurona u termalnoj koloni i bioloskom eksperimentalnom kanalu; Neutronska merenja u elementarnoj celiji reaktora RA; Temperaturni koeficijent reaktivnosti reaktora RA; Projekat uredjaja za merenje radioaktivnosti zice.

Dosimetry work and calculations in connection with the irradiation of large devices in the high flux materials testing reactor BR2

International Nuclear Information System (INIS)

De Raedt, C.; Leenders, L.; Tourwe, H.; Farrar, H. IV.

1982-01-01

For about fifteen years the high flux reactor BR2 has been involved in the testing of fast reactor fuel pins. In order to simulate the fast reactor neutron environment most devices are irradiated under cadmium screen, cutting off the thermal flux component. Extensive neutronic calculations are performed to help the optimization of the fuel bundle design. The actual experiments are preceded by irradiations of their mock-ups in BR02, the zero power model of BR2. The mock-up irradiations, supported by supplementary calculations, are performed for the determination of the main neutronic characteristics of the irradiation proper in BR2 and for the determination of the corresponding operation data. At the end of the BR2 irradiation, the experimental results, such as burn-ups, neutron fluences, helium production in the fuel pin claddings, etc. are correlated by neutronic calculations in order to examine the consistency of the post-irradiation results and to validate the routine calculation procedure and cross-section data employed. A comparison is made in this paper between neutronic calculation results and some post-irradiation data for MOL 7D, a cadmium screened sodium cooled loop containing a nineteen fuel pin bundle

A Graphite Isotope Ratio Method: A Primer on Estimating Plutonium Production in Graphite Moderated Reactors

International Nuclear Information System (INIS)

Gesh, Christopher J.

2004-01-01

The Graphite Isotope Ratio Method (GIRM) is a technique used to estimate the total plutonium production in a graphite-moderated reactor. The cumulative plutonium production in that reactor can be accurately determined by measuring neutron irradiation induced isotopic ratio changes in certain impurity elements within the graphite moderator. The method does not require detailed knowledge of a reactor's operating history, although that knowledge can decrease the uncertainty of the production estimate. The basic premise of the Graphite Isotope Ratio Method is that the fluence in non-fuel core components is directly related to the cumulative plutonium production in the nuclear fuel

Seeing the sink beneath the source: an improved stable isotope tracer method for measuring highly variable gross fluxes of methyl halides

Science.gov (United States)

Rhew, R. C.

2011-12-01

Measuring methyl bromide (CH3Br) and methyl chloride (CH3Cl) fluxes in terrestrial ecosystems is complicated by the presence of simultaneous production (typically associated with plants and/or fungi) and consumption (typically associated with soils). Thus, specific sites within an ecosystem can act as either a net source or net sink, depending on season, soil conditions, or vegetative cover. To interpret the highly variable net fluxes found in many of these ecosystems, a stable isotope tracer technique has been developed to measure gross fluxes of CH3Br and CH3Cl. This method entails adding small amounts of 13CH3Br and 13CH3Cl to an incubation chamber, monitoring the headspace concentration changes of both 13C and 13C isotopologues, and applying a box model to simultaneously solve for gross production and consumption. Over the last decade, this technique has been successfully applied to laboratory soil incubations and field studies from a variety of ecosystems, including boreal forest, annual grasslands, shortgrass steppe, oak-savanna woodland, and Arctic tundra. These studies demonstrate that gross uptake rates are strongly affected by soil moisture within ecosystems but are on average much lower than previously estimated, and that gross production rates are highly dependent on plant species enclosed, with minor production within the soils as well. Measuring gross uptake rates is more challenging in ecosystems with large net emissions of methyl halides, such as coastal salt marshes, rice fields and certain grassland sites. Using the tallgrass prairie of Kansas as a case study, four slightly different models to calculate gross fluxes are compared. These models are largely in agreement except at sites with large emissions (i.e., sites with Amorpha shrubs), where one of the models most robustly quantifies gross consumption. This improved stable isotope tracer method is used to track the separate responses of gross production and gross consumption of methyl halides

Proposal for a new method of reactor neutron flux distribution determination

Energy Technology Data Exchange (ETDEWEB)

Popic, V R [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1964-01-15

A method, based on the measurements of the activity produced in a medium flowing with variable velocity through a reactor, for the determination of the neutron flux distribution inside a reactor is considered theoretically (author)

Control Rods in high-Flux Swimming-Pool Reactors; Les Barres de Controle dans les Piles Piscines a Haut Flux; Reguliruyushchie sterzhni dlya reaktorov bassejnovogo tipa s vysokoj plotnost'yu nejtronnogo potoka; Las Barras de Control en los Reactores Tipo Piscina de Flujo Elevado

Energy Technology Data Exchange (ETDEWEB)

Ageroni, P.; Blum, P.; Denielou, G.; Denis, P.; Meunier, C. [Centre d' Etudes Nucleaires de Grenoble (France)

1964-06-15

Control-rod problems in open swimming-pool high-flux and high specific power research reactors are examined in the light of the calibrations and experiments made during the construction of the SILOE reactor. Control-rod operating experience for this reactor at 13 MW is also described. 2. The following are considered in turn: (a) Reactivity balances and reactivity values for the different types of rod tested (cadmium, B4C , rare earths and combinations of these different elements). (b) Flux peaks set up in the core by the presence of the control rods, their incidence on the specific power, the fast fluxes that can be obtained and means of increasing them. (c ) The technological problems involved in constructing the rods. (d) In-pile cooling, vibration, deformation and scram-time problems. 3. In conclusion, current studies on control rods in open swimming-pool reactors operating in the 10 - 30 1W range are briefly summarized. (author) [French] 1. Les problemes poses par les barres de controle dans les reacteurs de recherche de type piscine ouverte a haute puissance specifique et haut flux sont examines a la lumiere des calculs et des experiences effectues pendant la construction du reacteur SILOE. Les resultats de l'experience de fonctionnement a 13 MW de ce reacteur sont egalement presentes en ce qui concerne les barres de controle. 2. On examine successivement: a) les bilans de reactivite et les valeurs en reactivite des differents types de barres qui ont ete essayes (Cadmium, B 4C , terres rares et combinaisons de ces differents elements). b) Les pics de flux crees dans le coeur par la presence de barres de controle, leur incidence sur la puissance specifique, et les flux rapides que l'on peut obtenir ainsi que les moyens correspondants d'accroitre ces flux. c) Les problemes technologiques poses par la construction des barres. d) Les problemes de refrigeration, de vibration, de deformation, de temps de chute en pile. 3. En conclusion on decrit sommairement les

The neutron small-angle camera D11 at the high-flux reactor, Grenoble

International Nuclear Information System (INIS)

Ibel, K.

1976-01-01

The neutron small-angle scattering system at the high-flux reactor in Grenoble consists of three major parts: the supply of cold neutrons via bent neutron guides; the small-angle camera D11; and the data handling facilities. The camera D11 has an overall length of 80 m. The effective length of the camera is variable. The full length of the collimator before the fixed sample position can be reduced by movable neutron guides; the second flight path of 40 m full length contains detector sites in various positions. Thus a large range of momentum transfers can be used with the same relative resolution. Scattering angles between 5 x 10 -4 and 0.5 rad and neutron wavelengths from 0.2 to 2.0 nm are available. A large-area position-sensitive detector is used which allows simultaneous recording of intensities scattered at different angles; it is a multiwire proportional chamber. 3808 elements of 1 cm 2 are arranged in a two-dimensional matrix. (Auth.)

Investigation of the effects of radiolytic-gas bubbles on the long-term operation of solution reactors for medical-isotope production

Science.gov (United States)

Souto Mantecon, Francisco Javier

One of the most common and important medical radioisotopes is 99Mo, which is currently produced using the target irradiation technology in heterogeneous nuclear reactors. The medical isotope 99Mo can also be produced from uranium fission using aqueous homogeneous solution reactors. In solution reactors, 99Mo is generated directly in the fuel solution, resulting in potential advantages when compared with the target irradiation process in heterogeneous reactors, such as lower reactor power, less waste heat, and reduction by a factor of about 100 in the generation of spent fuel. The commercial production of medical isotopes in solution reactors requires steady-state operation at about 200 kW. At this power regime, the formation of radiolytic-gas bubbles creates a void volume in the fuel solution that introduces a negative coefficient of reactivity, resulting in power reduction and instabilities that may impede reactor operation for medical-isotope production. A model has been developed considering that reactivity effects are due to the increase in the fuel-solution temperature and the formation of radiolytic-gas bubbles. The model has been validated against experimental results from the Los Alamos National Laboratory uranyl fluoride Solution High-Energy Burst Assembly (SHEBA), and the SILENE uranyl nitrate solution reactor, commissioned at the Commissariat a l'Energie Atomique, in Valduc, France. The model shows the feasibility of solution reactors for the commercial production of medical isotopes and reveals some of the important parameters to consider in their design, including the fuel-solution type, 235U enrichment, uranium concentration, reactor vessel geometry, and neutron reflectors surrounding the reactor vessel. The work presented herein indicates that steady-state operation at 200 kW can be achieved with a solution reactor consisting of 120 L of uranyl nitrate solution enriched up to 20% with 235U and a uranium concentration of 145 kg/m3 in a graphite

New Monte Carlo-based method to evaluate fission fraction uncertainties for the reactor antineutrino experiment

Energy Technology Data Exchange (ETDEWEB)

Ma, X.B., E-mail: maxb@ncepu.edu.cn; Qiu, R.M.; Chen, Y.X.

2017-02-15

Uncertainties regarding fission fractions are essential in understanding antineutrino flux predictions in reactor antineutrino experiments. A new Monte Carlo-based method to evaluate the covariance coefficients between isotopes is proposed. The covariance coefficients are found to vary with reactor burnup and may change from positive to negative because of balance effects in fissioning. For example, between {sup 235}U and {sup 239}Pu, the covariance coefficient changes from 0.15 to −0.13. Using the equation relating fission fraction and atomic density, consistent uncertainties in the fission fraction and covariance matrix were obtained. The antineutrino flux uncertainty is 0.55%, which does not vary with reactor burnup. The new value is about 8.3% smaller. - Highlights: • The covariance coefficients between isotopes vs reactor burnup may change its sign because of two opposite effects. • The relation between fission fraction uncertainty and atomic density are first studied. • A new MC-based method of evaluating the covariance coefficients between isotopes was proposed.

Connection factor calculation for isotopic neutron flux measurements with foil detectors

International Nuclear Information System (INIS)

Avila L, J.

1987-01-01

Thermal and resonance neutron self-shielding factors, neutron flux distortion and edge effects as well as a connection factor for neutron flux profile around a foil detector have been calculated. A general expression for resonance self shielding factor is presented in order to take into account the most important resonances for a given isotope. A computer program SPRESYTER.BAS was written and results for In-115 and Au-197 foils are given

Transport calculation of neutron flux distribution in reflector of PW reactor

International Nuclear Information System (INIS)

Remec, I.

1982-01-01

Two-dimensional transport calculation of the neutron flux and spectrum in the equatorial plain of PW reactor, using computer program DOT 3, is presented. Results show significant differences between neutron fields in which test samples and reactor vessel are exposed. (author)

Isotopes accumulation in the thermal column of TRIGA reactor

International Nuclear Information System (INIS)

Iorgulis, C.; Diaconu, D.; Gugiu, D.; Csaba, R.

2013-01-01

The correlation of impurity observed in the virgin graphite and radionuclide content and activities measured in the irradiated graphite needs to know the irradiated history. This is a challenging process if impurity content and irradiation conditions are not accurately known. This is the case of the irradiated graphite in the thermal column of Institute for Nuclear Research Pitesti (INR)14 MW TRIGA reactor. To overcome incomplete impurity content and the unknown position in the column of the measured irradiated graphite available for characterisation and comparison, a set of preliminary simulations were performed. Following Eu 152 /Eu 154 ration they allowed the estimation of an impurity content and irradiation conditions leading to measured activities. Based on these data the radio-isotope accumulation in different positions in the thermal column was predicted. Modelling performed by INR used advanced prediction packages (e.g. WIMS, MCNP ORIGEN-S from Scale 5) to assess the isotopic content of MTR graphite types with irradiation history specific for a TRIGA research reactor. Some certain calculations points from the column were selected in order to model the burnup and isotopes productions using ORIGEN from SCALE code system. (authors)

Preliminary study of a flux converter for experimental reactor

International Nuclear Information System (INIS)

Malouch, M.F.

1998-01-01

The purpose of this project is to define the characteristics of a flux converter dedicated to increase the fast neutron flux in irradiation devices placed in the core of Osiris experimental reactor. This preliminary work has dealt with the neutronic and thermal-hydraulic aspects of this problem. The synthesis of the results produced by the codes APOLLO2, DAIXY, MERCURE5.3 and FLICA-3M shows that a cylindrical converter equipped with 5 fissile rings can enhance the fast flux by a 35% factor in an experimental device set in its center. (A.C.)

Equipment for thermal neutron flux measurements in reactor R2

Energy Technology Data Exchange (ETDEWEB)

Johansson, E; Nilsson, T; Claeson, S

1960-04-15

For most of the thermal neutron flux measurements in reactor R2 cobalt wires will be used. The loading and removal of these wires from the reactor core will be performed by means of a long aluminium tube and electromagnets. After irradiation the wires will be scanned in a semi-automatic device.

Solution of the isotopic depletion equation using decomposition method and analytical solution

Energy Technology Data Exchange (ETDEWEB)

Prata, Fabiano S.; Silva, Fernando C.; Martinez, Aquilino S., E-mail: fprata@con.ufrj.br, E-mail: fernando@con.ufrj.br, E-mail: aquilino@lmp.ufrj.br [Coordenacao dos Programas de Pos-Graduacao de Engenharia (PEN/COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Nuclear

2011-07-01

In this paper an analytical calculation of the isotopic depletion equations is proposed, featuring a chain of major isotopes found in a typical PWR reactor. Part of this chain allows feedback reactions of (n,2n) type. The method is based on decoupling the equations describing feedback from the rest of the chain by using the decomposition method, with analytical solutions for the other isotopes present in the chain. The method was implemented in a PWR reactor simulation code, that makes use of the nodal expansion method (NEM) to solve the neutron diffusion equation, describing the spatial distribution of neutron flux inside the reactor core. Because isotopic depletion calculation module is the most computationally intensive process within simulation systems of nuclear reactor core, it is justified to look for a method that is both efficient and fast, with the objective of evaluating a larger number of core configurations in a short amount of time. (author)

Solution of the isotopic depletion equation using decomposition method and analytical solution

International Nuclear Information System (INIS)

Prata, Fabiano S.; Silva, Fernando C.; Martinez, Aquilino S.

2011-01-01

In this paper an analytical calculation of the isotopic depletion equations is proposed, featuring a chain of major isotopes found in a typical PWR reactor. Part of this chain allows feedback reactions of (n,2n) type. The method is based on decoupling the equations describing feedback from the rest of the chain by using the decomposition method, with analytical solutions for the other isotopes present in the chain. The method was implemented in a PWR reactor simulation code, that makes use of the nodal expansion method (NEM) to solve the neutron diffusion equation, describing the spatial distribution of neutron flux inside the reactor core. Because isotopic depletion calculation module is the most computationally intensive process within simulation systems of nuclear reactor core, it is justified to look for a method that is both efficient and fast, with the objective of evaluating a larger number of core configurations in a short amount of time. (author)

Dosimetry of mixed gamma - neutron fluxes in the active zone of working reactor and gamma-flux after quenching

International Nuclear Information System (INIS)

Mussaeva, M.A.; Zinov'ev, V.; Ibragimova, E.M.; Muminov, M.I.

2006-01-01

vacancy, varied within 0.57 - 2.8. Besides, pure SiO 2 samples in the Cd - can filled with water were irradiated in the thermal column of operating reactor for 6 hours. Under these conditions the fast neutron flux was estimated as weak as 6·10 10 n/cm 2 s, the fluence was 1.3·10 15 cm -2 . The optical density of band 215 nm was 2.5, while the neutron fluence was ∼30 times less. Thus, the concentration of E ' -centers does not correlate with a neutron fluence. To extract the contribution from gamma-rays into the induced optical absorption in the glass matrix, samples of pure SiO 2 were irradiated by gamma-rays in 4 hours after quenching the reactor at the ionization current of 50 nA during 30 minutes, 12 and 24 hours; next time in 9 hours after the quenching at 40 nA and for 120 hours at 10 nA. In this case the gamma-spectrum did not include 10 MeV line from oxygen due to the short life-time, which prevails in the spectrum of working reactor. Maximal dose of γ-radiation of the quenched reactor was shown to induce the band at 215 nm up to the density of 0.5. When the sample was in contact with water the efficiency of E'-center production was 2 times higher that in dry condition. Thus, the high efficiency of structure defect production in SiO 2 glass owes to the influence of 10 MeV γ-radiation of the working reactor. The work was carried out under the grant F2.1.2 from Center of Science and Technology of Uzbekistan and supported by NATO CBP.EAP.CLG.981765. (author)

The high flux reactor Petten, A multi-purpose research and test facility for the future of nuclear energy

International Nuclear Information System (INIS)

Bergmans, H.; Duijves, K.; Conrad, R.; Markgraf, J.F.W.; May, R.; Moss, R.L.; Sordon, G.; Tartaglia, G.P.

1996-01-01

The High Flux Reactor (HFR) at Petten, is owned by the European Commission (EC) and managed by the Institute for Advanced Materials (IAM) of the Joint Research Centre (JRC) of the EC. Its operation has been entrusted since 1962 to the Netherlands Energy Research Foundation (ECN). The HFR is one of the most powerful multi-purpose research and test reactors in the world. Together with the ECN hot cells at Petten, it has provided since three decades an integral and full complement of irradiation and examination services as required by current and future research and development for nuclear energy, industry and research organizations. Since 1963, the HFR has recognized record of consistent, reliable and high availability of more than 250 days of operation per year. The HFR has 20 in-core and 12 poolside irradiation positions, plus 12 beam tubes. With a variety of dedicated irradiation devices, and with its long-standing experience in executing small and large irradiation projects, the HFR is particularly suited for fuel, materials and components testing for all reactor lines, including thermonuclear fusion reactors. In addition, processing with neutrons and gamma rays, neutron-based research and inspection services are employed by industry and research, such as activation analysis, boron neutron capture therapy, neutron radiography and neutron diffraction. Moreover, in recent years, HFRs' mission has been broadened within the area of radioisotopes production, where, within a few years, the HFR has attained the European leadership in production volume

Main technical options of the Jules Horowitz reactor project to achieve high flux performances and high safety level

International Nuclear Information System (INIS)

Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

2003-01-01

Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 kW/I for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: - A safety approach based upon the defence-in-depth principle. - A strategy of generic approaches to assess experimental risks in the facility. - Internal events analysis taking into account risks linked to reactor and experiments (e.g., radioactive source-term). - Systematic consideration of external hazards (e.g., earthquake, airplane crash) and internal hazards. - Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (authors)

Main technical options of the Jules Horowitz Reactor project to achieve high flux performances and high safety level

International Nuclear Information System (INIS)

Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P.

2003-01-01

Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it and will offer a quite larger experimental field. It has the ambition to provide the necessary nuclear data and to maintain a fission research capability in Europe after 2010. The Jules Horowitz Reactor will represent a significant step in terms of performances and experimental capabilities. This paper will present the main design option resulting from the preliminary studies. The choice of the specific power around 600 KW/l for the reference core configuration is a key decision to ensure the required flux level. Consequently many choices have to be made regarding the materials used in the core and the fuel element design. These involve many specific qualifications including codes validation. The main safety options are based on: 1) A safety approach based upon the defence-in-depth principle. 2) A strategy of generic approaches to assess experimental risks in the facility. 3) Internal events analysis taking into account risks linked to reactor and experiments (eg., radioactive source-term). 4) Systematic consideration of external hazards (eg., earthquake, airplane crash) and internal hazards. 5) Design of containment to manage and mitigate a severe reactor accident (consideration of 'BORAX' accident, according to french safety practice for MTRs, beyond design basis reactivity insertion accident, involving core melting and core destruction phenomena). (author)

Reference equilibrium core with central flux irradiation facility for Pakistan research reactor-1

International Nuclear Information System (INIS)

Israr, M.; Shami, Qamar-ud-din; Pervez, S.

1997-11-01

In order to assess various core parameters a reference equilibrium core with Low Enriched Uranium (LEU) fuel for Pakistan Research Reactor (PARR-1) was assembled. Due to increased volume of reference core, the average neutron flux reduced as compared to the first higher power operation. To get a higher neutron flux an irradiation facility was created in centre of the reference equilibrium core where the advantage of the neutron flux peaking was taken. Various low power experiments were performed in order to evaluate control rods worth and neutron flux mapping inside the core. The neutron flux inside the central irradiation facility almost doubled. With this arrangement reactor operation time was cut down from 72 hours to 48 hours for the production of the required specific radioactivity. (author)

RA Research reactor, Part I: Technical and operational properties of the RA reactor; Analiza sigurnosti rada Reaktora RA I-III, Deo I: Tehnicke i pogonske karakteristike reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Zecevic, V; Nikolic, M; Popovic, B; Milosevic, M; Milic, M; Strugar, P; Pesic, M; Nikolic, V; Rajic, M; Radivojevic, J; Jankovic, M [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

RA reactor is a research reactor with rather high power density. Apart from research it is used for isotope production and industrial applications due to high reactivity excess (about 11%). It is a thermal reactor, heavy water moderated, cooled by D{sub 2}O, and H{sub 2}O, with a graphite reflector. Nominal power is 6.5 MW. Fuel is 2% enriched metal uranium, reactor core height is 1220 mm, and diameter is 1405 mm. Reactor lattice is square with lattice pitch 130 mm. There is 6 horizontal experimental channels and a graphite column. There is a total of 84 fuel channels and 45 experimental channels in the core. Maximum thermal neutron flux is 5.5 10{sup 13} n/cm{sup 2} s at nominal power level.

Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements

International Nuclear Information System (INIS)

Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

2015-01-01

For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. - Highlights: • Neutron flux redistribution due to control rod movement in JSI TRIGA has been studied. • Detector response sensitivity to the control rod position has been minimized. • Optimal radial and axial detector positions have been determined

Seismic strengthening of the ILL High Flux Reactor building

International Nuclear Information System (INIS)

Germane, Lionel; Plewinski, Francois; Thiry, Jean-Michel

2006-01-01

The Institut Max von Laue - Paul Langevin is an international research organisation and world leader in neutron science and technology. Since 1971 it has been operating the ILL HFR (High-Flux Reactor), the most intense continuous neutron source in the world. The ILL is governed by an international cooperation agreement between France, Germany and the United Kingdom; the fourth ten-year extension to the agreement was signed at the end of 2002, thus ensuring that the Institute will continue to operate until at least the end of 2013. In 2002 the facility underwent a general safety review, including an assessment of the impact of a safe shutdown earthquake. A broader programme for upgrading the installations and improving safety levels is now under way. As this has been treated in another paper, we will focus here on the seismic study carried out on the reactor building. The paper has the following contents: 1. Context; 1.1. Presentation of the ILL; 1.2. Description of the installations; 1.3. Safety objectives in the event of an earthquake; 1.4. Safety functions to be guaranteed in the event of an earthquake; 1.5. Safety functions required of the building; 2. Description of the building; 3. Organisation of the project; 3.1. Background; 3.2. Organisation; 4. General Methodology of the studies; 5. Progress of the studies; 5.1. Definition of the strengthening measures; 5.2. Validation of the strengthening option; 6. Seismic strengthening of the building; 6.1. Description of the strengthening measures; 6.2. Implementation of the strengthening measures; 6.2.1. Pilot operation; 6.2.2. Main operation; 7. Conclusion. To summarize, the presence of specialists in the ILL team, and the fact that the initial studies were performed by the project team itself, improved our general understanding of the issues and facilitated dialogue and exchange between all those involved (operators, technicians, outside experts, technical contractors and the French safety authorities). Everyone was

Flux Analysis of Free Amino Sugars and Amino Acids in Soils by Isotope Tracing with a Novel Liquid Chromatography/High Resolution Mass Spectrometry Platform.

Science.gov (United States)

Hu, Yuntao; Zheng, Qing; Wanek, Wolfgang

2017-09-05

Soil fluxomics analysis can provide pivotal information for understanding soil biochemical pathways and their regulation, but direct measurement methods are rare. Here, we describe an approach to measure soil extracellular metabolite (amino sugar and amino acid) concentrations and fluxes based on a 15 N isotope pool dilution technique via liquid chromatography and high-resolution mass spectrometry. We produced commercially unavailable 15 N and 13 C labeled amino sugars and amino acids by hydrolyzing peptidoglycan isolated from isotopically labeled bacterial biomass and used them as tracers ( 15 N) and internal standards ( 13 C). High-resolution (Orbitrap Exactive) MS with a resolution of 50 000 allowed us to separate different stable isotope labeled analogues across a large range of metabolites. The utilization of 13 C internal standards greatly improved the accuracy and reliability of absolute quantification. We successfully applied this method to two types of soils and quantified the extracellular gross fluxes of 2 amino sugars, 18 amino acids, and 4 amino acid enantiomers. Compared to the influx and efflux rates of most amino acids, similar ones were found for glucosamine, indicating that this amino sugar is released through peptidoglycan and chitin decomposition and serves as an important nitrogen source for soil microorganisms. d-Alanine and d-glutamic acid derived from peptidoglycan decomposition exhibited similar turnover rates as their l-enantiomers. This novel approach offers new strategies to advance our understanding of the production and transformation pathways of soil organic N metabolites, including the unknown contributions of peptidoglycan and chitin decomposition to soil organic N cycling.

Flux Analysis of Free Amino Sugars and Amino Acids in Soils by Isotope Tracing with a Novel Liquid Chromatography/High Resolution Mass Spectrometry Platform

Science.gov (United States)

2017-01-01

Soil fluxomics analysis can provide pivotal information for understanding soil biochemical pathways and their regulation, but direct measurement methods are rare. Here, we describe an approach to measure soil extracellular metabolite (amino sugar and amino acid) concentrations and fluxes based on a 15N isotope pool dilution technique via liquid chromatography and high-resolution mass spectrometry. We produced commercially unavailable 15N and 13C labeled amino sugars and amino acids by hydrolyzing peptidoglycan isolated from isotopically labeled bacterial biomass and used them as tracers (15N) and internal standards (13C). High-resolution (Orbitrap Exactive) MS with a resolution of 50 000 allowed us to separate different stable isotope labeled analogues across a large range of metabolites. The utilization of 13C internal standards greatly improved the accuracy and reliability of absolute quantification. We successfully applied this method to two types of soils and quantified the extracellular gross fluxes of 2 amino sugars, 18 amino acids, and 4 amino acid enantiomers. Compared to the influx and efflux rates of most amino acids, similar ones were found for glucosamine, indicating that this amino sugar is released through peptidoglycan and chitin decomposition and serves as an important nitrogen source for soil microorganisms. d-Alanine and d-glutamic acid derived from peptidoglycan decomposition exhibited similar turnover rates as their l-enantiomers. This novel approach offers new strategies to advance our understanding of the production and transformation pathways of soil organic N metabolites, including the unknown contributions of peptidoglycan and chitin decomposition to soil organic N cycling. PMID:28776982

Flux measurement in ZBR at the TRIGA Mark II reactor

International Nuclear Information System (INIS)

Dauke, M.

2005-01-01

The determination of the neutron flux in the TRIGA-2-Vienna reactor was the objective of this research. The theory of the method (4π-β detectors) is presented as well as the determination of the maximum flux, gold-cadmium differential measurement, cobalt-wire measurement, finally a comparison of all results was made and interpreted. (nevyjel)

Fuel burnup analysis for the Moroccan TRIGA research reactor

International Nuclear Information System (INIS)

El Bakkari, B.; El Bardouni, T.; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Boukhal, H.; Zoubair, M.

2013-01-01

Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

An optimization study of peak thermal neutron flux in moderators of advanced repetitive pulse reactors

International Nuclear Information System (INIS)

Asaoka, Takumi; Watanabe, N.

1976-01-01

In achieving a high peak thermal neutron flux in hydrogenous moderators installed in repetitive pulse reactors, the core-moderator arrangement can play as much an important role as the moderator design itself. However, the effect of the former has not been adequately emphasized to date, while a rather extensive study has been made on the latter. The present study concerns with a core-moderator system parameter optimization for a repetitive accelerator pulsed fast reactor. The results have shown that small differences in the arrangement resulting from the optimizations of various parameters are significant and the effects can be summed up to give an increase in the peak thermal flux by a factor of about two. (auth.)

Analysis of JKT01 Neutron Flux Detector Measurements In RSG-GAS Reactor Using LabVIEW

Science.gov (United States)

Rokhmadi; Nur Rachman, Agus; Sujarwono; Taryo, Taswanda; Sunaryo, Geni Rina

2018-02-01

The RSG-GAS Reactor, one of the Indonesia research reactors and located in Serpong, is owned by the National Nuclear Energy Agency (BATAN). The RSG-GAS reactor has operated since 1987 and some instrumentation and control systems are considered to be degraded and ageing. It is therefore, necessary to evaluate the safety of all instrumentation and controls and one of the component systems to be evaluated is the performance of JKT01 neutron flux detector. Neutron Flux Detector JKT01 basically detects neutron fluxes in the reactor core and converts it into electrical signals. The electrical signal is then forwarded to the amplifier (Amplifier) to become the input of the reactor protection system. One output of it is transferred to the Main Control Room (RKU) showing on the analog meter as an indicator used by the reactor operator. To simulate all of this matter, a program to simulate the output of the JKT01 Neutron Flux Detector using LabVIEW was developed. The simulated data is estimated using a lot of equations also formulated in LabVIEW. The calculation results are also displayed on the interface using LabVIEW available in the PC. By using this simulation program, it is successful to perform anomaly detection experiments on the JKT01 detector of RSG-GAS Reactor. The simulation results showed that the anomaly JKT01 neutron flux using electrical-current-base are respectively, 1.5×,1.7× and 2.0×.

Breeding description for fast reactors and symbiotic reactor systems

International Nuclear Information System (INIS)

Hanan, N.A.

1979-01-01

A mathematical model was developed to provide a breeding description for fast reactors and symbiotic reactor systems by means of figures of merit type quantities. The model was used to investigate the effect of several parameters and different fuel usage strategies on the figures of merit which provide the breeding description. The integrated fuel cycle model for a single-reactor is reviewed. The excess discharge is automatically used to fuel identical reactors. The resulting model describes the accumulation of fuel in a system of identical reactors. Finite burnup and out-of-pile delays and losses are treated in the model. The model is then extended from fast breeder park to symbiotic reactor systems. The asymptotic behavior of the fuel accumulation is analyzed. The asymptotic growth rate appears as the largest eigenvalue in the solution of the characteristic equations of the time dependent differential balance equations for the system. The eigenvector corresponding to the growth rate is the core equilibrium composition. The analogy of the long-term fuel cycle equations, in the framework of this model, and the neutron balance equations is explored. An eigenvalue problem adjoint to the one generated by the characteristic equations of the system is defined. The eigenvector corresponding to the largest eigenvalue, i.e. to the growth rate, represents the ''isotopic breeding worths.'' Analogously to the neutron adjoint flux it is shown that the isotopic breeding worths represent the importance of an isotope for breeding, i.e. for the growth rate of a system

MURLI, 1-D Flux, Reaction Rate in Cylindrical Geometry Thermal Reactor Lattice by Transport

International Nuclear Information System (INIS)

Huria, H.C.

1985-01-01

1 - Description of problem or function: MURLI is an integral transport theory code to calculate fluxes and reaction rates in one- dimensional cylindrical geometry lattice cells of a thermal reactor. For a specified buckling, it computes k-effective using few-group diffusion theory and a few-group collapsed set of Cross sections. The code can optionally be used to solve a first order differential equation for the number density of fissile, fertile and fission product nuclei as a function of time, and to recalculate fluxes, reaction rates and k-effective at different stages of burnup. A 27-group cross section data library is included. There are four pseudo-fission products each associated with the decay chains of plutonium and uranium isotopes in addition to Rh-105, Xe-135, Np-239, U-236, Am-241, Am-242 and Am-243. There is also data for one lumped pseudo-fission product. 2 - Method of solution: Multiple collision probabilities and escape probabilities are calculated for each cylindrical shell region assuming protons are born uniformly and isotropically over the entire region volume. The equations of integral transport theory can then be solved for neutron flux. The first order differential burnup equation is solved by a fourth order Runge-Kutta method. 3 - Restrictions on the complexity of the problem: There are maxima of 8 fissionable elements, 8 resonant elements, and 20 spatial regions

Environmental effects on the response of self-powered flux detectors in CANDU reactors

International Nuclear Information System (INIS)

Lynch, G.F.; Shields, R.B.; Joslin, C.W.

1976-01-01

Self-powered flux detectors are playing an increasingly important role in the control and safety systems of CANDU-type reactors. In this paper we report on recent experiments to determine how local reactor conditions affect the output signals from self-powered detectors with vanadium, platinum and cobalt emitters. The results are interpreted in terms of variations in the local neutron, γ-ray and electron fluxes. (author)

Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

International Nuclear Information System (INIS)

Oakley, D.J.

1978-01-01

The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

Multivariate statistical pattern recognition system for reactor noise analysis

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

1976-01-01

A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system

Multivariate statistical pattern recognition system for reactor noise analysis

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.; Sides, W.H. Jr.; Kryter, R.C.

1975-01-01

A multivariate statistical pattern recognition system for reactor noise analysis was developed. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, and updating capabilities. System design emphasizes control of the false-alarm rate. The ability of the system to learn normal patterns of reactor behavior and to recognize deviations from these patterns was evaluated by experiments at the ORNL High-Flux Isotope Reactor (HFIR). Power perturbations of less than 0.1 percent of the mean value in selected frequency ranges were detected by the system. 19 references

Behaviour of aged and new flux detectors in Darlington reactors

Energy Technology Data Exchange (ETDEWEB)

Banica, C.; Foster, M., E-mail: Constantin.Banica@OPG.com [Ontario Power Generation, Darlington Nuclear, Bowmanville, Ontario (Canada)

2013-07-01

In-core neutron flux detectors are used for protective and safety functions in the Darlington NGS 'A' CANDU reactors. This paper presents new observations regarding the aging of flux detectors, including response to fuelling, response to unit shutdown and indicators of detector noise. Comparisons of detector signals before and after replacement confirm previous assumptions about aging effects. (author)

Comparison between different flux traps assembled in the core of the nuclear reactor IPEN/MB-01 by measuring of the thermal and epithermal neutron fluxes using activation foils

International Nuclear Information System (INIS)

Mura, Luiz Ernesto Credidio; Bitelli, Ulysses d'Utra; Mura, Luis Felipe Liambos; Carluccio, Thiago; Andrade, Graciete Simoes de

2011-01-01

The production of radioisotopes is one of the most important applications of nuclear research reactors. This study investigated a method called Flux Trap, which is used to increase the yield of production of radioisotopes in nuclear reactors. The method consists in the rearrangement of the fuel rods to allow the increase of the thermal neutron flux in the irradiation region inside the reactor core, without changing the standard reactor power level. Various configurations were assembled with the objective of finding the configuration with the highest thermal neutron flux in the region of irradiation. The method of activation analysis was used to measure the thermal neutron flux and determine the most efficient reactor core configuration . It was found that there was an increase in the thermal neutron flux of 337% in the most efficient configuration, which demonstrates the effectiveness of the method. (author)

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-04-01

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

Analyses of divertor high heat-flux components on thermal and electromagnetic loads

Energy Technology Data Exchange (ETDEWEB)

Araki, M.; Kitamura, K.; Suzuki, S. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Urata, K. [Mitsubishi Geavy Industries Ltd., 2-5-1, Marunouchi,Chiyoda-ku, Tokyo 100 (Japan)

1998-09-01

In the International Thermonuclear Experimental Reactor (ITER), the divertor high heat-flux components are subjected not to only severe heat and particle loads, but also to large electromagnetic loads during reactor operation. A great deal of R and D has been carried out throughout the world with regard to the design of robust high heat-flux components. Based on R and D results, small and intermediate size mock-ups constructed from various armor tile materials have been successfully developed with respect to a thermomechanical point of view. However, little analysis has been carried out with regard to the elastic stresses induced with in the high heat-flux components via the electromagnetic loads during a plasma disruption. Furthermore, past research has only considered thermomechanical and electromagnetic loadings separately and uncoupled. Therefore, a systematic analysis of the combined effects of thermomechanical and electromagnetic loadings has been performed, with the analytical results assessed by ASME section 3 evaluation code. (orig.) 20 refs.

Analyses of divertor high heat-flux components on thermal and electromagnetic loads

International Nuclear Information System (INIS)

Araki, M.; Kitamura, K.; Suzuki, S.

1998-01-01

In the International Thermonuclear Experimental Reactor (ITER), the divertor high heat-flux components are subjected not to only severe heat and particle loads, but also to large electromagnetic loads during reactor operation. A great deal of R and D has been carried out throughout the world with regard to the design of robust high heat-flux components. Based on R and D results, small and intermediate size mock-ups constructed from various armor tile materials have been successfully developed with respect to a thermomechanical point of view. However, little analysis has been carried out with regard to the elastic stresses induced with in the high heat-flux components via the electromagnetic loads during a plasma disruption. Furthermore, past research has only considered thermomechanical and electromagnetic loadings separately and uncoupled. Therefore, a systematic analysis of the combined effects of thermomechanical and electromagnetic loadings has been performed, with the analytical results assessed by ASME section 3 evaluation code. (orig.)

Critical heat flux correlation analysis for PWR reactors with low mass flow

International Nuclear Information System (INIS)

Carajilescov, Pedro

1996-01-01

The major limit in the thermalhydraulic design of water cooled reactors consists in the occurrence of critical heat flux, which is verified by correlation of large range of validity. In the present work, the major design correlations were analyzed, through comparisons with experimental data, for utilization in PWR with low mass flux in the core. The results show that the EPRI correlation, with modifications, gives conservative results, from the safety point of view, with lower data spreading, being the most indicated for the reactor thermal design. (author)

Radiation effects on reactor pressure vessel supports

International Nuclear Information System (INIS)

Johnson, R.E.

1996-05-01

The purpose of this report is to present the findings from the work done in accordance with the Task Action Plan developed to resolve the Nuclear Regulatory Commission (NRC) Generic Safety Issue No. 15, (GSI-15). GSI-15 was established to evaluate the potential for low-temperature, low-flux-level neutron irradiation to embrittle reactor pressure vessel (RPV) supports to the point of compromising plant safety. An evaluation of surveillance samples from the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) had suggested that some materials used for RPV supports in pressurized-water reactors could exhibit higher than expected embrittlement rates. However, further tests designed to evaluate the applicability of the HFIR data to reactor RPV supports under operating conditions led to the conclusion that RPV supports could be evaluated using traditional method. It was found that the unique HFIR radiation environment allowed the gamma radiation to contribute significantly to the embrittlement. The shielding provided by the thick steel RPV shell ensures that degradation of RPV supports from gamma irradiation is improbable or minimal. The findings reported herein were used, in part, as the basis for technical resolution of the issue

System for unattended surveillance of nuclear reactor behavior

International Nuclear Information System (INIS)

Gonzalez, R.C.; Howington, L.C.

1977-01-01

A multivariate statistical pattern recognition system for reactor noise analysis is presented. The basis of the system is a transformation for decoupling correlated variables and algorithms for inferring probability density functions. The system is adaptable to a variety of statistical properties of the data, and it has learning, tracking, updating, and dimensionality reduction capabilities. System design emphasizes control of the false-alarm rate. Its abilities to learn normal patterns and to recognize deviations from these patterns were evaluated by experiments at the ORNL High-Flux Isotope Reactor. Power perturbations of less than 0.1% of the mean value in selected frequency ranges were readily detected by the pattern recognition system

A parametric study on characteristics for nuclear design of high-performance research reactor

International Nuclear Information System (INIS)

Joe, D. G.; Lee, C. S.; Lee, B. C.; Seo, C. G.; Chae, H. T.; Park, C.

2003-01-01

A conceptual design of advanced research reactor with high neutron performance has been performed at KAERI based on design and operation experience obtained from HANARO. In this study, nuclear characteristics of design parameters such as various types of fuel assemblies, structural materials of core and fuel assembly, and the number of absorber rods were analyzed. Among rod, plate and tube type fuel assemblies considered, tube type assembly seems to be preferable as a high performance research reactor fuel because of high thermal margin and neutron flux in reflector. Aluminium block as a structural material of core was shown to be superior to flow tube due to higher reactivity and thermal flux in reflector. The stiffener to fix plates in th fuel assembly had the no impact on fast flux in central trap. The reduction of thermal flux in reflector caused by the stiffener was about 7%. If the control absorber rods of 4 mm thickness were chosen, it would be possible to operate the reactor with fresh fuel assemblies from the initial core

The CEA research reactors

International Nuclear Information System (INIS)

Schwartz, J.P.

1993-01-01

Two main research reactors, specifically designed, PEGASE reactor and Laue-Langevin high flux reactor, are presented. The PEGASE reactor was designed at the end of the 50s for the study of the gas cooled reactor fuel element behaviour under irradiation; the HFR reactor, was designed in the late 60s to serve as a high yield and high level neutron source. Historical backgrounds, core and fuel characteristics and design, flux characteristics, etc., are presented. 5 figs

A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

International Nuclear Information System (INIS)

1998-01-01

The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events

A Level 1+ Probabilistic Safety Assessment of the High Flux Australian Reactor. Vol 1

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The Department of Industry, Science and Tourism selected PLG, an EQE International Company, to systematically and independently evaluate the safety of the High Flux Australian Reactor (HIFAR), located at Lucas Heights, New South Wales. PLG performed a comprehensive probabilistic safety assessment (PSA) to quantify the risks posed by operation of HIFAR . The PSA identified possible accident scenarios, estimated their likelihood of occurrence, and assigned each scenario to a consequence category; i.e., end state. The accident scenarios developed included the possible release of radioactive material from irradiated nuclear fuel and of tritium releases from reactor coolant. The study team developed a recommended set of safety criteria against which the results of the PSA may be judged. HIFAR was found to exceed one of the two primary safety objectives and two of the five secondary safety objectives. Reactor coolant leaks, earthquakes, and coolant pump trips were the accident initiators that contributed most to scenarios that could result in fuel overheating. Scenarios initiated by earthquakes were the reason the frequency criterion for the one primary safety objective was exceeded. Overall, the plant safety status has been shown to be generally good with no evidence of major safety-related problems from its operation. One design deficiency associated with the emergency core cooling system was identified that should be corrected as soon as possible. Additionally, several analytical issues have been identified that should be investigated further. The results from these additional investigations should be used to determine whether additional plant and procedural changes are required, or if further evaluations of postulated severe accidents are warranted. Supporting information can be found in Appendix A for the seismic analysis and in the Appendix B for selected other external events refs., 139 tabs., 85 figs. Prepared for Department of Industry, Science and Tourism

Dalhousie SLOWPOKE-2 reactor: A nuclear analytical chemistry facility

International Nuclear Information System (INIS)

Chatt, A.; Holzbecher, J.

1990-01-01

SLOWPOKE is an acronym for Safe Low POwer Kritical Experiment. The SOWPOKE-2 is a compact, inherently safe, swimming-pool-type reactor designed by the Atomic Energy of Canada Limited for neutron activation analysis (NAA) and isotope production. The Dalhousie University SLOWPOKE-2 reactor (DUSR) has been operating since 1976; a large beryllium reflector was added in 1986 to extend its lifetime by another 8 to 10 yr. The DUSR is generally operated at half-power with a maximum thermal flux of 1.1 x 10 12 n/cm 2 ·s in the inner pneumatic sites and that of 5.4 x 10 11 n/cm 2 ·s in the outer sites. Despite this comparatively low flux, SLOWPOKE-2 reactors have many beneficial features that are continuously being exploited at the DUSR facility for developing nuclear analytical methods for fundamental as well as applied studies. Although NAA is a well-established analytical technique, much of the activation analysis being performed in most facilities has been limited to methods using fairly long-lived nuclides. The approach at the DUSR facility has been to utilize the highly homogeneous, stable, and reproducible neutron flux to develop NAA methods based on short-lived nuclides. SLOWPOKE reactors have a fairly high epithermal neutron flux, which is being advantageously used for determining several trace elements in complex matrices. Radiochemical NAA (RNAA) methods using coprecipitation, distillation, and ion-exchange separations have been used for the determination of very low levels of several elements in biological materials

Transient neutrons flux behaviour in a spherical reactor core

International Nuclear Information System (INIS)

Souza, A.W.A. de.

1978-11-01

This work studies the transient neutron flux in a fast reactor of spherical geometry. The burning of U 235 nuclei is equated and two kinds of reflector were studied. The numeric solutions are then compared with the results for those reflectors. (author) [pt

Role of plasma enhanced atomic layer deposition reactor wall conditions on radical and ion substrate fluxes

Energy Technology Data Exchange (ETDEWEB)

Sowa, Mark J., E-mail: msowa@ultratech.com [Ultratech/Cambridge NanoTech, 130 Turner Street, Building 2, Waltham, Massachusetts 02453 (United States)

2014-01-15

Chamber wall conditions, such as wall temperature and film deposits, have long been known to influence plasma source performance on thin film processing equipment. Plasma physical characteristics depend on conductive/insulating properties of chamber walls. Radical fluxes depend on plasma characteristics as well as wall recombination rates, which can be wall material and temperature dependent. Variations in substrate delivery of plasma generated species (radicals, ions, etc.) impact the resulting etch or deposition process resulting in process drift. Plasma enhanced atomic layer deposition is known to depend strongly on substrate radical flux, but film properties can be influenced by other plasma generated phenomena, such as ion bombardment. In this paper, the chamber wall conditions on a plasma enhanced atomic layer deposition process are investigated. The downstream oxygen radical and ion fluxes from an inductively coupled plasma source are indirectly monitored in temperature controlled (25–190 °C) stainless steel and quartz reactors over a range of oxygen flow rates. Etch rates of a photoresist coated quartz crystal microbalance are used to study the oxygen radical flux dependence on reactor characteristics. Plasma density estimates from Langmuir probe ion saturation current measurements are used to study the ion flux dependence on reactor characteristics. Reactor temperature was not found to impact radical and ion fluxes substantially. Radical and ion fluxes were higher for quartz walls compared to stainless steel walls over all oxygen flow rates considered. The radical flux to ion flux ratio is likely to be a critical parameter for the deposition of consistent film properties. Reactor wall material, gas flow rate/pressure, and distance from the plasma source all impact the radical to ion flux ratio. These results indicate maintaining chamber wall conditions will be important for delivering consistent results from plasma enhanced atomic layer deposition

Steel slag carbonation in a flow-through reactor system: the role of fluid-flux.

Science.gov (United States)

Berryman, Eleanor J; Williams-Jones, Anthony E; Migdisov, Artashes A

2015-01-01

Steel production is currently the largest industrial source of atmospheric CO2. As annual steel production continues to grow, the need for effective methods of reducing its carbon footprint increases correspondingly. The carbonation of the calcium-bearing phases in steel slag generated during basic oxygen furnace (BOF) steel production, in particular its major constituent, larnite {Ca2SiO4}, which is a structural analogue of olivine {(MgFe)2SiO4}, the main mineral subjected to natural carbonation in peridotites, offers the potential to offset some of these emissions. However, the controls on the nature and efficiency of steel slag carbonation are yet to be completely understood. Experiments were conducted exposing steel slag grains to a CO2-H2O mixture in both batch and flow-through reactors to investigate the impact of temperature, fluid flux, and reaction gradient on the dissolution and carbonation of steel slag. The results of these experiments show that dissolution and carbonation of BOF steel slag are more efficient in a flow-through reactor than in the batch reactors used in most previous studies. Moreover, they show that fluid flux needs to be optimized in addition to grain size, pressure, and temperature, in order to maximize the efficiency of carbonation. Based on these results, a two-stage reactor consisting of a high and a low fluid-flux chamber is proposed for CO2 sequestration by steel slag carbonation, allowing dissolution of the slag and precipitation of calcium carbonate to occur within a single flow-through system. Copyright © 2014. Published by Elsevier B.V.

Flux distribution measurements in the Bruce A unit 1 reactor

International Nuclear Information System (INIS)

Okazaki, A.; Kettner, D.A.; Mohindra, V.K.

1977-07-01

Flux distribution measurements were made by copper wire activation during low power commissioning of the unit 1 reactor of the Bruce A generating station. The distribution was measured along one diameter near the axial and horizontal midplanes of the reactor core. The activity distribution along the copper wire was measured by wire scanners with NaI detectors. The experiments were made for five configurations of reactivity control mechanisms. (author)

1984 Operation of the high flux reactor

International Nuclear Information System (INIS)

1985-01-01

The programme resources in 1984 were largely devoted to the replacement of the old reactor vessel and its peripheral equipment. The original vessel had been in operation for more than 20 years and doubts had arisen about the condition of the aluminium tank after so long an exposure to neutrons. The operation, which had never been attempted before on a reactor of that size and complexity was planned and prepared over a number of years to take advantage of the occasion to provide a much improved vessel, incorporating the latest design features. The plant was shut down at the end of November 1983 and the 14 months operation began with a short cooling-off period for decay of short lived radioactivity followed by removal of the old tank and its dissection into pieces convenient for consolidation and storage as radioactive waste. After decontamination of the shielding pool, the new vessel and neutron beam tubes were installed and the reactor was recommissioned. Routine 45 MW operation was resumed on 14 February 1985 and has been uneventful since then

Australia's new high performance research reactor

International Nuclear Information System (INIS)

Miller, R.; Abbate, P.M.

2003-01-01

A contract for the design and construction of the Replacement Research Reactor was signed in July 2000 between ANSTO and INVAP from Argentina. Since then the detailed design has been completed, a construction authorization has been obtained, and construction has commenced. The reactor design embodies modern safety thinking together with innovative solutions to ensure a highly safe and reliable plant. Also significant effort has been placed on providing the facility with diverse and ample facilities to maximize its use for irradiating material for radioisotope production as well as providing high neutron fluxes for neutron beam research. The project management organization and planing is commensurate with the complexity of the project and the number of players involved. (author)

Review of inservice inspection and nondestructive examination practices at DOE Category A test and research reactors

International Nuclear Information System (INIS)

Anderson, M.T.; Aldrich, D.A.

1990-09-01

In-service inspection (ISI) programs are used at commercial nuclear power plants for monitoring the pressure boundary integrity of various systems and components to ensure their continued safe operation. The Department of Energy (DOE) operates several test and research reactors. This report represents an evaluation of the ISI and nondestructive examination (NDE) practices at five DOE Category A (> 20 MW thermal) reactors as compared, where applicable, to the current ISI activities of commercial nuclear power facilities. The purpose of an inservice inspection (ISI) program is to establish regular surveillance of safety-related components to ensure their safe and reliable operation. The integrity of materials comprising these components is generally monitored by means of periodic nondestructive examinations (NDE), which, if appropriately performed, provide methods for identifying degradation that could render components unable to perform their intended safety functions. The reactors evaluated during this review were the Experimental Breeder Reactor 2 and the Fast Flux Test Facility (liquid-metal cooled plants), the Advanced Test Reactor and the High Flux Isotopes Reactor (light-water cooled reactors), and the High Flux Beam Reactor (a heavy-water cooled facility). Although these facilities are extremely diverse in design and operation, they all have less stored energy, smaller inventories of radionuclides, and generally, more remote locations than commercial reactors. However, all DOE test and research facilities contain components similar to those of commercial reactors for which continued integrity is important to maintain plant safety. 10 refs., 6 tabs

Thermal neutron flux measurements using neutron-electron converters; Mesure de flux de neutrons thermiques avec des convertisseurs neutrons electrons

Energy Technology Data Exchange (ETDEWEB)

Le Meur, R; Lecomte, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1968-07-01

The operation of neutron-electron converters designed for measuring thermal neutron fluxes is examined. The principle is to produce short lived isotopes emitting beta particles, by activation, and to measure their activity not by extracting them from the reactor, but directly in the reactor using the emitted electrons to deflect the needle of a galvanometer placed outside the flux. After a theoretical study, the results of the measurements are presented; particular attention is paid to a new type of converter characterized by a layer structure. The converters are very useful for obtaining flux distributions with more than 10{sup 7} neutrons cm{sup -2}*sec{sup -1}. They work satisfactorily in pressurized carbon dioxide at 400 Celsius degrees. Some points still have to be cleared up however concerning interfering currents in the detectors and the behaviour of the dielectrics under irradiation. (authors) [French] On examine le fonctionnement de convertisseurs neutrons electrons destines a des mesures de flux de neutrons thermiques. Le principe est de former par activation des isotopes a periodes courtes et a emission beta et de mesurer leur activite non pas en les sortant du reacteur, mais directement en pile, utilisant les electrons emis pour faire devier l'aiguille d'un galvanometre place hors flux. Apres une etude theorique, on indique des resultats de mesures obtenus, en insistant particulierement sur un nouveau type de convertisseur, caracterise par sa structure stratifiee. Les convertisseurs sont tres interessants pour tracer, des cartes de flux a partir de 10{sup 7} neutrons cm{sup -2}*s{sup -1}. Ils sont utilisables pour des flux de 10{sup 14} neutrons cm{sup -2}*s{sup -1}. Ils fonctionnent correctement dans du gaz carbonique sous pression a 400 C. Des points restent cependant a eclaircir concernant les courants parasites dans les detecteurs et le comportement des dielectriques pendant leur irradiation. (auteur)

Research and materials irradiation reactors

International Nuclear Information System (INIS)

Ballagny, A.; Guigon, B.

2004-01-01

Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

Isotopic nuclear reactor with on-line separation

International Nuclear Information System (INIS)

Liviu, Popa-Simil

2007-01-01

In the new reactor-waste cycle design the nuclear reactor gets features of the living beings - resembling the plants/vegetation -. The separation of waste starts inside the fuel by using the fission reaction to separate the fission products from the fuel. The fuel, which is preferred to be highly isotopic enriched, is fabricated in beads smaller than the fission product range, immersed in a gentle flowing liquid drain. If this liquid is Lead Bismuth (LBE) the fission products will be lighter, while in Sodium-Potassium (NaK) will be heavier, except for gases. This drain liquid will collect both the fission products and the collision damage, drawing them slow to give time to short lives disintegration chains to take place inside the shielded nuclear reactor area outside the reactor core in a separation unit. While the drain liquid with the fission products is outside the reactor core few choices are available: - To solidify the drain liquid freezing all elements inside and transport the metal in cryogenic conditions to a remote separation unit, or to apply a separation partitioning process online stabilizing and packing the fission products only, or a combination of these two. The radioactivity of this drain liquid is smaller than that of the actual used fuel because it represents the accumulation of a very short period (about 1 month or less) and had enough time to cool down all the short lives. The separation unit on-line with the nuclear reactor is composed of a density separation unit, followed by a phase interface concentration unit which moves out of the LBE the fission products as lighter impurities, and an electrochemical separation unit for the fission products. Further, chemical separation, stabilization processes are applied and the fission products are delivered partitioned on groups of chemical compatible products. Finally the specific waste is about 1 Kg/Gw*day, to which the stabilization products have to be added which increases this mass by 10 times

Comparison of neutron fluxes obtained by 2-D and 3-D geometry with different shielding libraries in biological shield of the TRIGA MARK II reactor

International Nuclear Information System (INIS)

Bozic, M.; Zagar, T.; Ravnik, M.

2003-01-01

Neutron fluxes in different spatial locations in biological shield are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Libraries used with TORT code were BUGLE-96 library (coupled library with 47 neutron groups and 20 gamma groups) and VITAMIN-B6 library (coupled library with 199 neutron groups and 42 gamma groups). BUGLE-96 library is derived from VITAMIN-B6 library. 2-D and 3-D models for homogeneous type of problem (without inserted beam port 4) and problem with asymmetry (non-homogeneous problem; inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. The main purpose is to verify the possibility for using 2-D approximation model instead of large 3-D model in some calculations. Another purpose of this paper was to compare neutron spectral constants obtained from neutron fluxes (3-D model) determined with smaller BUGLE-96 library with new constants obtained from fluxes calculated with bigger VITAMIN-B6 library. These neutron spectral constants are used in isotopic calculation with SCALE code package (ORIGEN-S). In past only neutron spectral constants determined by neutron fluxes from BUGLE-96 library were used. Experimental results used for isotopic composition comparison are available from irradiation experiment with selected type of concrete and other materials in beam port 4 (irradiation channel 4) in TRIGA Mark II reactor. These experimental results were used as a benchmark in this paper. (author)

Measure of thermal neutron flux in the IPEN/MB-01 reactor using 197 Au wire activation detectors

International Nuclear Information System (INIS)

Marques, Andre Luis Ferreira

1995-01-01

This dissertation has aimed at developing a neutron flux measurement technique by means of detectors activation analysis. The main task of this work was the implementation of this thermal neutron flux measurement technique, using gold wires as activation detectors in the IPEN/MB-01 reactor core. The neutron thermal flux spatial distribution was obtained by gold wire activation technique, with wire diameters of 0.125 mm and 0.250 mm in seven selected reactor experimental channels. The values of thermal flux were about 10 9 neutrons/cm 2 .s. This experiment has been the first one conducted with gold wires in the IPEN/MB-01 reactor, being this technique implemented for use by experiments in flux mapping of the core

Final Report Independent Verification Survey of the High Flux Beam Reactor, Building 802 Fan House Brookhaven National Laboratory Upton, New York

Energy Technology Data Exchange (ETDEWEB)

Harpeneau, Evan M. [Oak Ridge Institute for Science and Education, Oak Ridge, TN (United States). Independent Environmental Assessment and Verification Program

2011-06-24

On May 9, 2011, ORISE conducted verification survey activities including scans, sampling, and the collection of smears of the remaining soils and off-gas pipe associated with the 802 Fan House within the HFBR (High Flux Beam Reactor) Complex at BNL. ORISE is of the opinion, based on independent scan and sample results obtained during verification activities at the HFBR 802 Fan House, that the FSS (final status survey) unit meets the applicable site cleanup objectives established for as left radiological conditions.

ITER TASK T26/28 (1995): Solubility, diffusion and absorption of hydrogen isotopes in potential fusion reactor ceramics

International Nuclear Information System (INIS)

Thompson, D.A.; Macauley-Newcombe, R.G.

1996-04-01

Ceramic insulators are integral parts of numerous components essential for the heating control and diagnostic measurement of fusion plasmas. For safe and reliable reactor operations it is necessary to be able to predict the resultant tritium inventories and permeation fluxes through these materials. Some materials being considered are Al 2 O 3 (both as single crystal sapphire and polycrystalline alumina) and BeO. This report contains results of ion-implantation, thermal absorption (diffusion loading) and ion-beam analysis experiments performed in 1994 and 1995 for ITER task T26/28. The combination of implantation and thermal absorption capabilities enable us to load samples with hydrogen isotopes under differing conditions. 13 figs., 1 tab., 11 refs

Estimation of radioactivity in structural materials of ETRR-1 reactor

Energy Technology Data Exchange (ETDEWEB)

Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

1997-12-31

Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

Pursuing nuclear energy with no nuclear contamination - from neutron flux reactor to deuteron flux reactor

International Nuclear Information System (INIS)

Li, X. Z.; Wei, Q. M.; Liu, B.; Zhu, X. G.; Ren, S. L.

2007-01-01

Pursuing nuclear energy with no nuclear contamination has been a long endeavor since the first fission reactor in 1942. Four major concepts have been the key issues: i.e. resonance, negative feed back, self-sustaining, nuclear radiation. When nuclear energy was just discovered in laboratory, the key issue was to enlarge it from the micro-scale to the macro-scale. Slowing-down the neutrons was the key issue to enhance the fission cross-section in order to build-up the neutron flux through the chain-reactions using resonance between neutron and fissile materials. Once the chain-reaction was realized, the negative feed-back was the key issue to keep the neutron flux at the allowable level. The negative reaction coefficient was introduced by the thermal expansion, and the resonant absorption in cadmium or boron was used to have a self-sustaining fission reactor with neutron flux. Then the strong neutron flux became the origin of all nuclear contamination, and a heavy shielding limits the application of the nuclear energy. The fusion approach to nuclear energy was much longer; nevertheless, it evolved with the similar issues. The resonance between deuteron and triton was resorted to enlarge the fusion cross section in order to keep a self-sustaining hot plasma. However, the 14 MeV neutron emission became the origin of all nuclear contamination again. Deuteron plus helium-3 fusion reaction was proposed to avoid neutron emission although there are two more difficulties: the helium-3 is supposed to be carried back from the moon; and much more higher temperature plasma has to be confined while 50 years needed to realized the deuteron-triton plasma already. Even if deuteron plus helium-3 fusion plasma might be realized in a much higher temperature plasma, we still have the neutron emission from the deuteron-deuteron fusion reaction in the deuteron plus helium-3 fusion plasma. Polarized deuteron-deuteron fusion reaction was proposed early in 1980's to select the neutron

Development of SiC Neutron Detector Assembly to Measure the Neutron Flux of the Reactor Core

Energy Technology Data Exchange (ETDEWEB)

Park, Se Hwan; Park, June Sic; Shin, Hee Sung; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Yong Kyun [Hanyang University, Seoul (Korea, Republic of)

2012-05-15

At present, the conventional detector to measure the neutron at harsh environment is a Self Powered Neutron Detector (SPND). Rhodium(Rh)-103 is in the SPND. When neutron is incident on the Rhodium, the neutron capture reaction occurs, and the Rh-103 is converted to Rh-104. The Rh-104 is decayed to Pd-104 by {beta}-decay, and electrons are generated as the decay products. Because of the half life of Rh-104, approximately 5 minutes are required for the SPND output to reach the equilibrium condition. Therefore the on-line monitoring of the nuclear reactor state is limited if the neutron flux in the reactor core is monitored with the SPND. Silicon carbide (SiC) has the possibility to be developed as neutron detector at harsh environment, because the SiC can be operative at high temperature and high neutron flux conditions. Previously, the basic operation properties of the SiC detector were studied. Also, the radiation response of the SiC detector was studied at high neutron and gamma dose rate. The measurement results for an ex-core neutron flux monitor or a neutron flux monitor of the spent fuel were published. The SiC detector was also developed as neutron detector to measure the fissile material with active interrogation method. However, the studies about the development of SiC detector are still limited. In the present work, the radiation damage effect of the SiC detector was studied. The detector structure was determined based on the study, and a neutron detector assembly was made with the SiC detectors. The neutron and gamma-ray response of the detector assembly is presented in this paper. The detector assembly was positioned in the HANARO research reactor core, the performance test was done. The preliminary results are also included in this paper

Influence of implanted helium on nickel resistance under simulation of plasma flux disruption in nuclear fusion reactor

International Nuclear Information System (INIS)

Kadin, B.A.; Pol'skij, V.I.; Yakushin, V.L.; Markin, A.V.; Tserevitinov, S.S.; Vasil'ev, V.I.

1992-01-01

Investigation results are presented of radiation erosion of constructive materials of the first wall of a thermonuclear reactor. The erosion is conditioned by successive repeated action of pulse processes, imitating plasma disruption, and helium ion fluxes at 40 keV and 2 x 10 21 -10 22 m -2 fluence. As imitating processes are used fluxes of deuterium high-temperature plasma. It is shown that preliminary action by high-temperature plasma leads to substantial suppression of radiation erosion, included by subsequent ion irradiation

Future prospects of reactor Ra at VINCA institute

International Nuclear Information System (INIS)

Davidovic, M.; Babic-Stojic, B.; Dobrijevic, R.

1997-01-01

Reactor RA at Nuclear Research Institute Vinca belongs to a group of the medium thermal neutron flux reactors, according to classification at end of nineties. At the beginning reactor RA has been used as a powerful source of neutrons and gamma-quanta for various experiments (interaction of neutrons and gamma-quanta with materials) and for production of artificial radioactive materials for commercial use. Very successful utilization of this neutron spectrum has been in its use for structural studies of crystal materials and liquid metals, for magnetic structure studies of various magnetic materials, as well as, dynamic properties of ferro magnetics, ferroelectrics, etc. This kind of spectrometers still exist at reactor RA and with an improved detection system could be used again if reactor starts functioning. Besides this, a part of activity was devoted to construction of neutron guide tubes for thermal neutrons and this could also be accomplished relatively easy in the future. A part of the activities of the reactor should in the future be devoted to the training of students in the field of solid state physics and nuclear physics. Particular attention will be paid to the use of established technologies in production of radioactive isotopes and a new class of isotopes for custom use will be developed as well as highly commercial and prospective products (silicon doping, radiography, etc.). (author)

Measurement and calculation of spatial and energetic neutron flux in the IEA-R1 reactor core

International Nuclear Information System (INIS)

Bittelli, U.D.

1988-01-01

This work presents spatial and energetic flux distribution measured in the IEA-R1 reactor core. The thermal neutron flux was measured by gold activation foils (bare and covered with cadmium) in the fuel element number 108 (reaction: 197 Au(n,γ) 198 Au) at 451W overall reactor power. The fast neutron flux was measured by indium activation foils (reaction: 115 In(n,n') 115m In) in the fuel elements number 94 at 4510W overall reactor power. The neutron energy spectrum was adjusted by SAND II code with the data produced by the irradiation of seven activation detectors in the fuel element number 94 at 4510 W overall reactor power. The following reactions were used: 58 Fe(n,γ) 59 Fe, 232 Th(n,γ) 233 Th, 197 Au(n,γ) 198 Au, 59 Co(n,γ) 60 Co, 54 Fe(n,p) 54 Mn, 24 Mg(n,p) 24 Na, 47 Ti(n,p) 47 Sc, 48 Ti(n,p) 48 Sc and 115 In(n,n') 115m In. The experimental results compared to those obtained by CITATION (spatial distribution flux) and HAMMER (energetic distribution flux) code, showed good agreement. The results presented in this work are a good contribution for a better knowledge of spatial and energetic neutron flux distribution in the IEA-R1 reactor core, besides that the experimental procedure is easily applicable to another situations. (autor) [pt

Experimental spectrum of reactor antineutrinos and spectra of main fissile isotopes

Energy Technology Data Exchange (ETDEWEB)

Sinev, V. V., E-mail: vsinev@pcbai10.inr.ruhep.ru [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation)

2013-05-15

Within the period between the years 1988 and 1990, the spectrum of positrons from the inverse-beta-decay reaction on a proton was measured at the Rovno atomic power plant in the course of experiments conducted there. The measured spectrum has the vastest statistics in relation to other neutrino experiments at nuclear reactors and the lowest threshold for positron detection. An experimental reactor-antineutrino spectrum was obtained on the basis of this positron spectrum and was recommended as a reference spectrum. The spectra of individual fissile isotopes were singled out from the measured antineutrino spectrum. These spectra can be used to analyze neutrino experiments performed at nuclear reactors for various compositions of the fuel in the reactor core.

Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

International Nuclear Information System (INIS)

Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez; Universidade Federal de Pernambuco

2017-01-01

The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly "9"9Mo. Compare to multipurpose research reactors, an AHR dedicated for "9"9Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

Neutronic and thermal-hydraulic studies of aqueous homogeneous reactor for medical isotopes production

Energy Technology Data Exchange (ETDEWEB)

Perez, Daniel Milian; Lorenzo, Daniel E. Milian; Lira, Carlos A. Brayner de Oliveira; Garcia, Lorena P. Rodríguez, E-mail: milianperez89@gmail.com, E-mail: dmilian@instec.cu, E-mail: lorenapilar1109@gmail.com, E-mail: cabol@ufpe.br [Higher Institute of Technologies and Applied Sciences (InSTEC), Havana (Cuba); Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Departamento de Energia Nuclear

2017-11-01

The use of Aqueous Homogenous Reactors (AHR) is one of the most promissory alternatives to produce medical isotopes, mainly {sup 99}Mo. Compare to multipurpose research reactors, an AHR dedicated for {sup 99}Mo production has advantages because of their low cost, small critical mass, inherent passive safety, and simplified fuel handling, processing, and purification characteristics. This article presents the current state of research in our working group on this topic. Are presented and discussed the group validation efforts with benchmarking exercises that include neutronic and thermal-hydraulic results of two solution reactors, the SUPO and ARGUS reactors. Neutronic and thermal-hydraulic results of 75 kWth AHR based on the ARGUS reactor LEU configuration are presented. The neutronic studies included the determination of parameters such as reflector thickness, critical height, medical isotopes production and others. Thermal-hydraulics studies were focused on demonstrating that sufficient cooling capacity exists to prevent fuel overheating. In addition, the effects of some calculation parameters on the computational modeling of temperature, velocity and gas volume fraction during steady-state operation of an AHR are discussed. The neutronic and thermal-hydraulics studies have been performed with the MCNPX version 2.6e computational code and the version 14 of ANSYS CFX respectively. Our group studies and the results obtained contribute to demonstrate the feasibility of using AHR for the production of medical isotopes, however additional studies are still necessary to confirm these results and contribute to development and demonstration of their technical, safety, and economic viability. (author)

Doping-Induced Isotopic Mg11B2 Bulk Superconductor for Fusion Application

Directory of Open Access Journals (Sweden)

Qi Cai

2017-03-01

Full Text Available Superconducting wires are widely used for fabricating magnetic coils in fusion reactors. Superconducting magnet system represents a key determinant of the thermal efficiency and the construction/operating costs of such a reactor. In consideration of the stability of 11B against fast neutron irradiation and its lower induced radioactivation properties, MgB2 superconductor with 11B serving as the boron source is an alternative candidate for use in fusion reactors with a severe high neutron flux environment. In the present work, the glycine-doped Mg11B2 bulk superconductor was synthesized from isotopic 11B powder to enhance the high field properties. The critical current density was enhanced (103 A·cm−2 at 20 K and 5 T over the entire field in contrast with the sample prepared from natural boron.

Surveillance of a nuclear reactor by use of a pattern recognition methodology

International Nuclear Information System (INIS)

Dubuisson, B.; Lavison, P.

1980-01-01

A multivariate nonparametric pattern recognition system is described for the surveillance of a high-flux isotope reactor. Two nonparametric methods are worked out: one using the Bayes rule with the Rosenblatt-Parzen estimator for the probability law, and one using the k-nearest neighbor rule. Performances are evaluated by comparing the probability of misclassification between the two chosen classes: the first corresponds to a nonaction of the reactor operator on its power and the second to an action of the pilot. Processing is performed on the power signal of the reactor which is an observation corrupted by noise. The system has been tested on several experiences and implemented to work in real time on the reactor. The aim is to conceive a computer-aided decision system for the reactor's pilot. 17 refs

A high temperature reactor for ship propulsion

International Nuclear Information System (INIS)

Lobet, P.; Seigel, R.; Thompson, A.C.; Beadnell, R.M.; Beeley, P.A.

2002-01-01

The initial thermal hydraulic and physics design of a high temperature gas cooled reactor for ship propulsion is described. The choice of thermodynamic cycle and thermal power is made to suit the marine application. Several configurations of a Helium cooled, Graphite moderated reactor are then analysed using the WIMS and MONK codes from AEA Technology. Two geometries of fuel elements formed using micro spheres in prismatic blocks, and various arrangements of control rods and poison rods are examined. Reactivity calculations through life are made and a pattern of rod insertion to flatten the flux is proposed and analysed. Thermal hydraulic calculations are made to find maximum fuel temperature under high power with optimized flow distribution. Maximum temperature after loss of flow and temperatures in the reactor vessel are also computed. The temperatures are significantly below the known limits for the type of fuel proposed. It is concluded that the reactor can provide the required power and lifetime between refueling within likely space and weight constraints. (author)

Basic research using the 250 kW research reactor of the Jozef Stefan Institute

International Nuclear Information System (INIS)

Dimic, V.

1984-01-01

The 250 kW TRIGA Mark II reactor is a light water reactor with solid fuel elements in which the zirconium hydride moderator is homogeneously distributed between enriched uranium. The reactor therefore has a large prompt negative temperature coefficient of reactivity; the fuel also has a very high retention of radioactive fission products. The experimental facilities include a rotary specimen rack, a central in-core radiation thimble, a pneumatic transfer system and pulsing capability. Other experimental facilities include two radial and two tangential beam tubes, a graphite thermal column and a graphite thermalizing column. At the steady state power of 250 kW the peak flux is 1x10 13 n/cm 2 in the central test position. In addition, pulsing to about 2000 MW is usually provided giving peak fluxes of about 2x10 16 n/cm 2 sec. All TRIGA reactors produce a core-average thermal neutron flux of about 10 7 n.v. per watt. Only with very large accelerators can such high fluxes be achieved. The types of research could be summarized as follows: thermal neutron scattering, neutron radiography, neutron and nuclear physics, activation analysis, radiochemistry, biology and medicine, and teaching and training. Typical applied research with a 250 kW reactor has been conducted in medicine, in biology, archaeology, metallurgy and materials science, engineering and criminology. It is well known that research reactors have been used routinely to produce isotopes for industry and medicine. We can conclude that the 250 kW TRIGA reactor is a useful and wide ranging source of radiation for basic and applied research. The operation cost for this instrument is relatively low. (author)

Tensile and impact testing of an HFBR [High Flux Beam Reactor] control rod follower

International Nuclear Information System (INIS)

Czajkowski, C.J.; Schuster, M.H.; Roberts, T.C.; Milian, L.W.

1989-08-01

The Materials Technology Group of the Department of Nuclear Energy (DNE) at Brookhaven National Laboratory (BNL) undertook a program to machine and test specimens from a control rod follower from the High Flux Beam Reactor (HFBR). Tensile and Charpy impact specimens were machined and tested from non-irradiated aluminum alloys in addition to irradiated 6061-T6 from the HFBR. The tensile test results on irradiated material showed a two-fold increase in tensile strength to a maximum of 100.6 ksi. The impact resistance of the irradiated material showed a six-fold decrease in values (3 in-lb average) compared to similar non-irradiated material. Fracture toughness (K I ) specimens were tested on an unirradiated compositionally and dimensionally similar (to HFBR follower) 6061 T-6 material with K max values of 24.8 ± 1.0 Ksi√in (average) being obtained. The report concludes that the specimens produced during the program yielded reproducible and believable results and that proper quality assurance was provided throughout the program. 9 figs., 6 tabs

Design of a Multi-Spectrum CANDU-based Reactor, MSCR, with 37-element fuel bundles using SERPENT code

International Nuclear Information System (INIS)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.; Chan, P.

2015-01-01

The burning of highly-enriched uranium and plutonium from dismantled nuclear warhead material in the new design nuclear power plants represents an important step towards nonproliferation. The blending of these highly enriched uranium and plutonium with with uranium dioxide from the spent fuel of CANDU reactors, or mixing it with depleted uranium would need a very long time to dispose of this material. Consequently, considering that more efficient transmutation of actinides occurs in fast neutron reactors, a novel Multi-Spectrum CANDU Reactor, has been designed on the basis of the CANDU6 reactor with two concentric regions. The simulations of the MSCR were carried out using the SERPENT code. The inner or fast neutron spectrum core is fuelled by different levels of enriched uranium oxides. The helium is used as a coolant in the fast neutron core. The outer or the thermal neutron spectrum core is fuelled with natural uranium with heavy water as both moderator and coolant. Both cores use 37- element fuel bundles. The size of the two cores and the percentage level of enrichment of the fresh fuel in the fast core were optimized according to the criticality safety of the whole reactor. The excess reactivity, the regeneration factor, radial and axial flux shapes of the MSCR reactor were calculated at different of the concentration of fissile isotope 235 U of uranium fuel at the fast neutron spectrum core. The effect of variation of the concentration of the fissile isotope on the fluxes in both cores at each energy bin has been studied. (author)

Design of a Multi-Spectrum CANDU-based Reactor, MSCR, with 37-element fuel bundles using SERPENT code

Energy Technology Data Exchange (ETDEWEB)

Hussein, M.S.; Bonin, H.W.; Lewis, B.J.; Chan, P., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca, E-mail: Paul.Chan@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, ON (Canada)

2015-07-01

The burning of highly-enriched uranium and plutonium from dismantled nuclear warhead material in the new design nuclear power plants represents an important step towards nonproliferation. The blending of these highly enriched uranium and plutonium with with uranium dioxide from the spent fuel of CANDU reactors, or mixing it with depleted uranium would need a very long time to dispose of this material. Consequently, considering that more efficient transmutation of actinides occurs in fast neutron reactors, a novel Multi-Spectrum CANDU Reactor, has been designed on the basis of the CANDU6 reactor with two concentric regions. The simulations of the MSCR were carried out using the SERPENT code. The inner or fast neutron spectrum core is fuelled by different levels of enriched uranium oxides. The helium is used as a coolant in the fast neutron core. The outer or the thermal neutron spectrum core is fuelled with natural uranium with heavy water as both moderator and coolant. Both cores use 37- element fuel bundles. The size of the two cores and the percentage level of enrichment of the fresh fuel in the fast core were optimized according to the criticality safety of the whole reactor. The excess reactivity, the regeneration factor, radial and axial flux shapes of the MSCR reactor were calculated at different of the concentration of fissile isotope {sup 235}U of uranium fuel at the fast neutron spectrum core. The effect of variation of the concentration of the fissile isotope on the fluxes in both cores at each energy bin has been studied. (author)

Accurate determination of Curium and Californium isotopic ratios by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS) in 248Cm samples for transmutation studies

Energy Technology Data Exchange (ETDEWEB)

Gourgiotis, A.; Isnard, H.; Aubert, M.; Dupont, E.; AlMahamid, I.; Cassette, P.; Panebianco, S.; Letourneau, A.; Chartier, F.; Tian, G.; Rao, L.; Lukens, W.

2011-02-01

The French Atomic Energy Commission has carried out several experiments including the mini-INCA (INcineration of Actinides) project for the study of minor-actinide transmutation processes in high intensity thermal neutron fluxes, in view of proposing solutions to reduce the radiotoxicity of long-lived nuclear wastes. In this context, a Cm sample enriched in {sup 248}Cm ({approx}97 %) was irradiated in thermal neutron flux at the High Flux Reactor (HFR) of the Laue-Langevin Institute (ILL). This work describes a quadrupole ICP-MS (ICP-QMS) analytical procedure for precise and accurate isotopic composition determination of Cm before sample irradiation and of Cm and Cf after sample irradiation. The factors that affect the accuracy and reproducibility of isotopic ratio measurements by ICP-QMS, such as peak centre correction, detector dead time, mass bias, abundance sensitivity and hydrides formation, instrumental background, and memory blank were carefully evaluated and corrected. Uncertainties of the isotopic ratios, taking into account internal precision of isotope ratio measurements, peak tailing, and hydrides formations ranged from 0.3% to 1.3%. This uncertainties range is quite acceptable for the nuclear data to be used in transmutation studies.

Publication of the second amendment to the German-French Convention on the construction and operation of a very high flux reactor and to its complementary agreement

International Nuclear Information System (INIS)

1982-01-01

Full text in German, English, and French of the amendment to the London Convention of December 9, 1981 between the Federal Republic of Germany, France, Great Britain and Northern Ireland concerning the operating cost and the use of the high flux reactor in Grenoble at the Max-von-Laue Institute. (HP) [de

Errors of absolute methods of reactor neutron activation analysis caused by non-1/E epithermal neutron spectra

International Nuclear Information System (INIS)

Erdtmann, G.

1993-08-01

A sufficiently accurate characterization of the neutron flux and spectrum, i.e. the determination of the thermal flux, the flux ratio and the epithermal flux spectrum shape factor, α, is a prerequisite for all types of absolute and monostandard methods of reactor neutron activation analysis. A convenient method for these measurements is the bare triple monitor method. However, the results of this method, are very imprecise, because there are high error propagation factors form the counting errors of the monitor activities. Procedures are described to calculate the errors of the flux parameters, the α-dependent cross-section ratios, and of the analytical results from the errors of the activities of the monitor isotopes. They are included in FORTRAN programs which also allow a graphical representation of the results. A great number of examples were calculated for ten different irradiation facilities in four reactors and for 28 elements. Plots of the results are presented and discussed. (orig./HP) [de

Hydrogen isotopes transport parameters in fusion reactor materials

International Nuclear Information System (INIS)

Serra, E.; Ogorodnikova, O.V.

1998-01-01

This work presents a review of hydrogen isotopes-materials interactions in various materials of interest for fusion reactors. The relevant parameters cover mainly diffusivity, solubility, trap concentration and energy difference between trap and solution sites. The list of materials includes the martensitic steels (MANET, Batman and F82H-mod.), beryllium, aluminium, beryllium oxide, aluminium oxide, copper, tungsten and molybdenum. Some experimental work on the parameters that describe the surface effects is also mentioned. (orig.)

The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

International Nuclear Information System (INIS)

Aziz, M.; El Messiry, A.M.

2000-01-01

The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

BR2 Reactor: Introduction

International Nuclear Information System (INIS)

Moons, F.

2007-01-01

The irradiations in the BR2 reactor are in collaboration with or at the request of third parties such as the European Commission, the IAEA, research centres and utilities, reactor vendors or fuel manufacturers. The reactor also contributes significantly to the production of radioisotopes for medical and industrial applications, to neutron silicon doping for the semiconductor industry and to scientific irradiations for universities. Along the ongoing programmes on fuel and materials development, several new irradiation devices are in use or in design. Amongst others a loop providing enhanced cooling for novel materials testing reactor fuel, a device for high temperature gas cooled fuel as well as a rig for the irradiation of metallurgical samples in a Pb-Bi environment. A full scale 3-D heterogeneous model of BR2 is available. The model describes the real hyperbolic arrangement of the reactor and includes the detailed 3-D space dependent distribution of the isotopic fuel depletion in the fuel elements. The model is validated on the reactivity measurements of several tens of BR2 operation cycles. The accurate calculations of the axial and radial distributions of the poisoning of the beryllium matrix by 3 He, 6 Li and 3T are verified on the measured reactivity losses used to predict the reactivity behavior for the coming decades. The model calculates the main functionals in reactor physics like: conventional thermal and equivalent fission neutron fluxes, number of displacements per atom, fission rate, thermal power characteristics as heat flux and linear power density, neutron/gamma heating, determination of the fission energy deposited in fuel plates/rods, neutron multiplication factor and fuel burn-up. For each reactor irradiation project, a detailed geometry model of the experimental device and of its neighborhood is developed. Neutron fluxes are predicted within approximately 10 percent in comparison with the dosimetry measurements. Fission rate, heat flux and

Digital, remote control system for a 2-MW research reactor

International Nuclear Information System (INIS)

Battle, R.E.; Corbett, G.K.

1988-01-01

A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

Status of Kijang Resarch Reactor Project

Energy Technology Data Exchange (ETDEWEB)

Kim, Jee. Y; Kwon, T. H.; Kim, Jun. Y.; Oh, G. B. [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The High-flux Advanced Neutron Application Reactor (HANARO) is a multi-purpose reactor in Korea Atomic Energy Research Institute (KAERI) and is being utilized for neutron scattering experiments, material and fuel tests for nuclear power plants, radio-isotope (RI) productions, silicon doping, neutron activation analysis, and neutron radiography. In medical applications, the majority of RIs produced using HANARO are I-131 and Ir-192. Other RIs such as Mo-99 are coming from imports. The self-sufficiency of RI demand becomes an important issue for the public health service in Korea. In this regard the Kijang Research Reactor (KJRR) project was officially launched on the first of April 2012 in need to provide the self-sufficiency of RI demand including Mo-99, increase the neutron transportation doping (NTD) capacity and develop technologies related to the research reactor. When CP is granted, the first excavation is planned to start at the end of this year. In next year, pouring the first concrete and energizing 154kV will follow. In 2018, it is planned to complete utility building construction and reactor building construction.

Investigation of hydrogen bubbles behavior in tungsten by high-flux hydrogen implantation

Science.gov (United States)

Zhao, Jiangtao; Meng, Xuan; Guan, Xingcai; Wang, Qiang; Fang, Kaihong; Xu, Xiaohui; Lu, Yongkai; Gao, Jun; Liu, Zhenlin; Wang, Tieshan

2018-05-01

Hydrogen isotopes retention and bubbles formation are critical issues for tungsten as plasma-facing material in future fusion reactors. In this work, the formation and growing up behavior of hydrogen bubbles in tungsten were investigated experimentally. The planar TEM samples were implanted by 6.0keV hydrogens to a fluence of 3.38 ×1018 H ṡ cm-2 at room temperature, and well-defined hydrogen bubbles were observed by TEM. It was demonstrated that hydrogen bubbles formed when exposed to a fluence of 1.5 ×1018 H ṡ cm-2 , and the hydrogen bubbles grew up with the implantation fluence. In addition, the bubbles' size appeared larger with higher beam flux until saturated at a certain flux, even though the total fluence was kept the same. Finally, in order to understand the thermal annealing effect on the bubbles behavior, hydrogen-implanted samples were annealed at 400, 600, 800, and 1000 °C for 3 h. It was obvious that hydrogen bubbles' morphology changed at temperatures higher than 800 °C.

Fuel requirements for isotope production and reasearch reactors: Possible alternative ways of meeting non-proliferation objectives

International Nuclear Information System (INIS)

There is a continuing need for access to medium-to-high flux research reactors of intermediate power level (5-50 MW) for the production of industrial and medical radioisotopes, for the provision of neutron beams and for materials research. The construction of further reactors of this type is likely. To obtain the required flux levels in adequate volumes and at the lowest capital cost, past practice has been to design a small-core reactor around a fuel element concept using fully enriched uranium, that is, uranium enriched to 80% U-235 or greater. In recent years, however, it has been recognised that the use of fully enriched uranium in research reactors could give rise to significant risks of nuclear weapons proliferation. Accordingly, there would be advantage if research reactors could be operated on low enriched fuel, that is, enrichment levels of 20% or less. It is the purpose of this paper to explore the implications for proliferation of the enrichment level of research reactor fuel and to draw attention to possible options for reducing proliferation concerns which warrant further study. It does not, however, consider research reactors using very low enriched or natural uranium fuel. The paper is offered to stimulate discussion of the issues and the views expressed do not necessarily represent any formal Australian position

Maintenance management at the High Flux Isotope Reactor

International Nuclear Information System (INIS)

Farrar, M.B.

1982-02-01

Maintenance procedures are described for mechanical and electrical equipment; nuclear and process instrumentation; operational maintenance; equipment and systems inspections; and HFIR quality assurance

Thermal neutron flux measurements in the rotary specimen rack of the IPR-R1 TRIGA reactor

Energy Technology Data Exchange (ETDEWEB)

Souza, Rose Mary G. do Prado; Rodrigues, Rogério R.; Souza, Luiz Claudio A., E-mail: souzarm@cdtn.br, E-mail: rrr@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2017-07-01

The thermal neutron flux in the rotary specimen rack of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center (CDTN), Belo Horizonte, Brazil, has been measured by the neutron activation method, using bare and cadmium covered gold foils. Those foils were irradiated in the rotary specimen rack with the reactor at 100 kW. The reactor core configuration has 63 fuel elements, composed of 59 original aluminum-clad elements and 4 stainless steel-clad fuel elements. The gamma activities of the foils were measured using Ge spectrometer. The perturbations of the thermal neutron flux caused by the introduction of an absorbing foil into the medium were considered in order to obtain accurate determination of the flux. The thermal neutron flux obtained was 7.4 x 10{sup 11} n.cm{sup -2}.s{sup -1}. (author)

Method for accounting for macroscopic heterogeneities in reactor material balance generation in fuel cycle simulations

Energy Technology Data Exchange (ETDEWEB)

Bagdatlioglu, Cem, E-mail: cemb@utexas.edu; Schneider, Erich

2016-06-15

Highlights: • Describes addition of spatially dependent power sharing to a previous methodology. • The methodology is used for calculating the input and output isotopics and burnup. • Generalizes to simulate reactors with strong spatial and flux heterogeneities. • Presents cases where the old approach would not have been sufficient. - Abstract: This paper describes the addition of spatially dependent power sharing to a methodology used for calculating the input and output isotopics and burnup of nuclear reactors within a nuclear fuel cycle simulator. Neutron balance and depletion calculations are carried out using pre-calculated fluence-based libraries. These libraries track the transmutation and neutron economy evolution of unit masses of nuclides available in input fuel. The work presented in the paper generalizes the method to simulate reactors that contain more than one type of fuel as well as strong spatial and flux heterogeneities, for instance breeders with a driver–blanket configuration. To achieve this, spatial flux calculations are used to determine the fluence-dependent relative average fluxes inside macroscopic spatial regions. These fluxes are then used to determine the average power of macroscopic spatial regions as well as to more accurately calculate region-specific transmutation rates. The paper presents several cases where the fluence based approach alone would not have been sufficient to determine results.