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Sample records for brookhaven graphite research reactor

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
Seismic research on graphite reactor core

International Nuclear Information System (INIS)

Lai Shigang; Sun Libin; Zhang Zhengming

2013-01-01

Background: Reactors with graphite core structure include production reactor, water-cooled graphite reactor, gas-cooled reactor, high-temperature gas-cooled reactor and so on. Multi-body graphite core structure has nonlinear response under seismic excitation, which is different from the response of general civil structure, metal connection structure or bolted structure. Purpose: In order to provide references for the designing and construction of HTR-PM. This paper reviews the history of reactor seismic research evaluation from certain countries, and summarizes the research methods and research results. Methods: By comparing the methods adopted in different gas-cooled reactor cores, inspiration for our own HTR seismic research was achieved. Results and Conclusions: In this paper, the research ideas of graphite core seismic during the process of designing, constructing and operating HTR-10 are expounded. Also the project progress of HTR-PM and the research on side reflection with the theory of similarity is introduced. (authors)
Deployment of Smart 3D Subsurface Contaminant Characterization at the Brookhaven Graphite Research Reactor

International Nuclear Information System (INIS)

Sullivan, T.; Heiser, J.; Kalb, P.; Milian, L.; Newson, C.; Lilimpakas, M.; Daniels, T.

2002-01-01

The Brookhaven Graphite Research Reactor (BGRR) Historical Site Assessment (BNL 1999) identified contamination inside the Below Grade Ducts (BGD) resulting from the deposition of fission and activation products from the pile on the inner carbon steel liner during reactor operations. Due to partial flooding of the BGD since shutdown, some of this contamination may have leaked out of the ducts into the surrounding soils. The baseline remediation plan for cleanup of contaminated soils beneath the BGD involves complete removal of the ducts, followed by surveying the underlying and surrounding soils, then removing soil that has been contaminated above cleanup goals. Alternatively, if soil contamination around and beneath the BGD is either non-existent/minimal (below cleanup goals) or is very localized and can be ''surgically removed'' at a reasonable cost, the BGD can be decontaminated and left in place. The focus of this Department of Energy Accelerated Site Technology Deployment (DOE ASTD) project was to determine the extent (location, type, and level) of soil contamination surrounding the BGD and to present this data to the stakeholders as part of the Engineering Evaluation/Cost Analysis (EE/CA) process. A suite of innovative characterization tools was used to complete the characterization of the soil surrounding the BGD in a cost-effective and timely fashion and in a manner acceptable to the stakeholders. The tools consisted of a tracer gas leak detection system that was used to define the gaseous leak paths out of the BGD and guide soil characterization studies, a small-footprint Geoprobe to reach areas surrounding the BGD that were difficult to access, two novel, field-deployed, radiological analysis systems (ISOCS and BetaScint) and a three-dimensional (3D) visualization system to facilitate data analysis/interpretation. All of the technologies performed as well or better than expected and the characterization could not have been completed in the same time or at
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
Brookhaven leak reactor to close

CERN Multimedia

MacIlwain, C

1999-01-01

The DOE has announced that the High Flux Beam Reactor at Brookhaven is to close for good. Though the news was not unexpected researchers were angry the decision had been taken before the review to assess the impact of reopening the reactor had been concluded (1 page).
Design of the Graphite Reflectors in Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Lee, Jin Haeng; Cho, Yeong Garp; Kim, Tae Kyu; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

Graphite is often used as one of reflector materials for research reactors because of its low neutron absorption cross-section, good moderating properties, and relatively low and stable price. In addition, graphite has excellent properties at high temperatures, so it is widely used as a core material in high temperature reactors. However, its material characteristics such as strength, elastic modulus, thermal expansion coefficient, dimensional change, and thermal conductivity sensitively depend on neutron fluence, temperature, and its manufacturing process. In addition, the Wigner energy and the treatment of the graphite waste such as C-14 should also be considered. For the design of the graphite reflectors, it is therefore essential to understand the material characteristics of chosen graphite materials at given conditions. Especially, the dimensional changes and the thermal conductivity are very important factors to design the nuclear components using graphite as a nonstructural material. Hence, in this study, the material characteristics of graphite are investigated via some experiments in literature. Improving design methods for graphite reflectors in research reactors are then suggested to minimize the problems, and the advantages and disadvantages of each method are also discussed
A neutronic feasibility study for LEU conversion of the Brookhaven Medical Research Reactor (BMRR).

Energy Technology Data Exchange (ETDEWEB)

Hanan, N. A.

1998-01-14

A neutronic feasibility study for converting the Brookhaven Medical Research Reactor from HEU to LEU fuel was performed at Argonne National Laboratory in cooperation with Brookhaven National Laboratory. Two possible LEU cores were identified that would provide nearly the same neutron flux and spectrum as the present HEU core at irradiation facilities that are used for Boron Neutron Capture Therapy and for animal research. One core has 17 and the other has 18 LEU MTR-type fuel assemblies with uranium densities of 2.5g U/cm{sup 3} or less in the fuel meat. This LEU fuel is fully-qualified for routine use. Thermal hydraulics and safety analyses need to be performed to complete the feasibility study.
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
Epithermal beam development at the BMRR [Brookhaven Medical Research Reactor]: Dosimetric evaluation

International Nuclear Information System (INIS)

Saraf, S.K.; Fairchild, R.G.; Kalef-Ezra, J.; Laster, B.H.; Fiarman, S.; Ramsey, E.; Ioannina Univ.; Brookhaven National Lab., Upton, NY; State Univ. of New York, Stony Brook, NY

1989-01-01

The utilization of an epithermal neutron beam for neutron capture therapy (NCT) is desirable because of the increased tissue penetration relative to a thermal neutron beam. Over the past few years, modifications have been and continue to be made at the Brookhaven Medical Research Reactor (BMRR) by changing its filter components to produce an optimal epithermal beam. An optimal epithermal beam should contain a low fast neutron contamination and no thermal neutrons in the incident beam. Recently a new moderator for the epithermal beam has been installed at the epithermal port of the BMRR and has accomplished this task. This new moderator is a combination of alumina (Al 2 O 3 ) bricks and aluminum (Al) plates. A 0.51 mm thick cadmium (Cd) sheet has reduced the thermal neutron intensity drastically. Furthermore, an 11.5 cm thick bismuth (Bi) plate installed at the port surface has reduced the gamma dose component to negligible levels. Foil activation techniques have been employed by using bare gold and cadmium-covered gold foil to determine thermal as well as epithermal neutron fluence. Fast neutron fluence has been determined by indium foil counting. Fast neutron and gamma dose in soft tissue, free in air, is being determined by the paired ionization chamber technique, using tissue equivalent (TE) and graphite chambers. Thermoluminescent dosimeters (TLD-700) have also been used to determine the gamma dose independently. This paper describes the methods involved in the measurements of the above mentioned parameters. Formulations have been developed and the various corrections involved have been detailed. 12 refs
Determination of Cl-36 in Irradiated Reactor Graphite

International Nuclear Information System (INIS)

Beer, H.-F.; Schumann, D.; Stowasser, T.; Hartmann, E.; Kramer, A.

2016-01-01

Two of the three research reactors at the Paul Scherrer Institute (PSI), the reactors DIORIT and PROTEUS, contained reactor graphite. Whereas the former research reactor DIORIT has been dismantled completely the PROTEUS is subject to a future decommissioning. In case of the DIORIT the reactor graphite was conditioned applying a procedure developed at PSI. In this case the 36 Cl content had to be determined after the conditioning. The result is reported in this paper. The radionuclide inventory including 36 Cl of the graphite used in PROTEUS was measured and the results are reported in here. It has been proven that the graphite from PROTEUS has a radionuclide inventory near the detection limits. All determined radionuclide activities are far below the Swiss exemptions limits. The graphite from PROTEUS therefore poses no radioactive waste. In contrast, the 36 Cl content of graphite from DIORIT is well above the exemption limits. (author)
Carbon-14 in neutron-irradiated graphite for graphite-moderated reactors. Joint research

Energy Technology Data Exchange (ETDEWEB)

Fujii, Kimio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Matsuo, Hideto [Radioactive Waste Management and Nuclear Facility Decommissioning Technology Center, Tokyo (Japan)

2002-12-01

The graphite moderated gas cooled reactor operated by the Japan Atomic Power Company was stopped its commercial operation on March 1998, and the decommissioning process has been started. Graphite material is often used as the moderator and the reflector materials in the core of the gas cooled reactor. During the operation, a long life nuclide of {sup 14}C is generated in the graphite by several transmutation reactions. Separation of {sup 14}C isotope and the development of the separation method have been recognized to be critical issues for the decommissioning of the reactor core. To understand the current methodologies for the carbon isotope separation, literature on the subject was surveyed. Also, those on the physical and chemical behavior of {sup 14}C were surveyed. This is because the larger part of the nuclides in the graphite is produced from {sup 14}N by (n,p) reaction, and the location of them in the material tends to be different from those of the other carbon atoms. This report summarizes the result of survey on the open literature about the behavior of {sup 14}C and the separation methods, including the list of the literature on these subjects. (author)
Small Modular Reactors (468th Brookhaven Lecture)

International Nuclear Information System (INIS)

Bari, Robert

2011-01-01

With good reason, much more media attention has focused on nuclear power plants than solar farms, wind farms, or hydroelectric plants during the past month and a half. But as nations around the world demand more energy to power everything from cell phone batteries to drinking water pumps to foundries, nuclear plants are the only non-greenhouse-gas producing option that can be built to operate almost anywhere, and can continue to generate power during droughts, after the sun sets, and when winds die down. To supply this demand for power, designers around the world are competing to develop more affordable nuclear reactors of the future: small modular reactors. Brookhaven Lab is working with DOE to ensure that these reactors are designed to be safe for workers, members of surrounding communities, and the environment and to ensure that the radioactive materials and technology will only be used for peaceful purposes, not weapons. In his talk, Bari will discuss the advantages and challenges of small modular reactors and what drives both international and domestic interest in them. He will also explain how Brookhaven Lab and DOE are working to address the challenges and provide a framework for small modular reactors to be commercialized.
Brookhaven highlights - Brookhaven National Laboratory 1995

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-09-01

This report highlights research conducted at Brookhaven National Laboratory in the following areas: alternating gradient synchrotron; physics; biology; national synchrotron light source; department of applied science; medical; chemistry; department of advanced technology; reactor; safety and environmental protection; instrumentation; and computing and communications.
Some equipment for graphite research in swimming pool reactors

International Nuclear Information System (INIS)

Seguin, M.; Arragon, Ph.; Dupont, G.; Gentil, J.; Tanis, G.

1964-01-01

The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [fr
Channel uranium-graphite reactor mounting

International Nuclear Information System (INIS)

Polushkin, K.K.; Kuznetsov, A.G.; Zheleznyakov, B.N.

1981-01-01

According to theoretical principles of general engineering technology the engineering experience of construction-mounting works at the NPP with channel uranium-graphite reactors is systematized. Main parameters and structural features of the 1000 MW channel uranium-graphite reactors are considered. The succession of mounting operations, premounting equipment and pipelines preparation and mounting works technique are described. The most efficient methods of fitting, welding and machining of reactor elements are recommended. Main problems of technical control service are discussed. A typical netted diagram of main equipment of channel uranium-graphite reactors mounting is given
Graphite materials for nuclear reactors

International Nuclear Information System (INIS)

Oku, Tatsuo

1991-01-01

Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)
Ageing Management of Beryllium and Graphite Blocks in Research Reactor MARIA

Energy Technology Data Exchange (ETDEWEB)

Golab, A. [National Centre for Nuclear Research, Warsaw (Poland)

2013-07-01

In the paper the phenomenon of beryllium moderator poisoning by thermal neutron absorption and the method and results of this phenomenon control is presented. Also the phenomenon of graphite blocks damage due to fast neutrons accumulation and the methods and results of this process supervising is described. These methods refer especially to: visual inspection of their state and radiography of graphite blocks. Special attention is paid to permanent estimate of fast neutron fluency accumulated in blocks and methods of their shuffling in the reactor core. The shuffling makes possible to increase the lifetime of beryllium and graphite blocks and decrease the cost of reactor operation.
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
Design of small-animal thermal neutron irradiation facility at the Brookhaven Medical Research Reactor

International Nuclear Information System (INIS)

Liu, H.B.

1996-01-01

The broad beam facility (BBF) at the Brookhaven Medical Research Reactor (BMRR) can provide a thermal neutron beam with flux intensity and quality comparable to the beam currently used for research on neutron capture therapy using cell-culture and small-animal irradiations. Monte Carlo computations were made, first, to compare with the dosimetric measurements at the existing BBF and, second, to calculate the neutron and gamma fluxes and doses expected at the proposed BBF. Multiple cell cultures or small animals could be irradiated simultaneously at the so-modified BBF under conditions similar to or better than those individual animals irradiated at the existing thermal neutron irradiation Facility (TNIF) of the BMRR. The flux intensity of the collimated thermal neutron beam at the proposed BBF would be 1.7 x 10 10 n/cm 2 ·s at 3-MW reactor power, the same as at the TNIF. However, the proposed collimated beam would have much lower gamma (0.89 x 10 -11 cGy·cm 2 /n th ) and fast neutron (0.58 x 10 -11 cGy·cm 2 /n th ) contaminations, 64 and 19% of those at the TNIF, respectively. The feasibility of remodeling the facility is discussed
Upgrades of the epithermal neutron beam at the Brookhaven Medical Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Liu, Hungyuan B.; Brugger, R.M.; Rorer, D.C.

1994-12-31

The first epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR) was installed in 1988 and produced a neutron beam that was satisfactory for the development of NCT with epithermal neutrons. This beam was used routinely until 1992 when the beam was upgraded by rearranging fuel elements in the reactor core to achieve a 50% increase in usable flux. Next, after computer modeling studies, it was proposed that the Al and Al{sub 2}O{sub 3} moderator material in the shutter that produced the epithermal neutrons could be rearranged to enhance the beam further. However, this modification was not started because a better option appeared, namely to use fission plates to move the source of fission neutrons closer to the moderator and the patient irradiation position to achieve more efficient moderation and production of epithermal neutrons. A fission plate converter (FPC) source has been designed recently and, to test the concept, implementation of this upgrade has started. The predicted beam parameters will be 12 x 10{sup 9} n{sub epi}/cm{sup 2}sec accompanying with doses from fast neutrons and gamma rays per epithermal neutron of 2.8 x 10{sup -11} and < 1 x 10{sup -11} cGycm{sup 2}/n, respectively, and a current-to-flux ratio of epithermal neutrons of 0.78. This conversion could be completed by late 1996.

Nuclear graphite for high temperature reactors

International Nuclear Information System (INIS)

Marsden, B.J.

2001-01-01

The cores and reflectors in modern High Temperature Gas Cooled Reactors (HTRs) are constructed from graphite components. There are two main designs; the Pebble Bed design and the Prism design. In both of these designs the graphite not only acts as a moderator, but is also a major structural component that may provide channels for the fuel and coolant gas, channels for control and safety shut off devices and provide thermal and neutron shielding. In addition, graphite components may act as a heat sink or conduction path during reactor trips and transients. During reactor operation, many of the graphite component physical properties are significantly changed by irradiation. These changes lead to the generation of significant internal shrinkage stresses and thermal shut down stresses that could lead to component failure. In addition, if the graphite is irradiated to a very high irradiation dose, irradiation swelling can lead to a rapid reduction in modulus and strength, making the component friable.The irradiation behaviour of graphite is strongly dependent on its virgin microstructure, which is determined by the manufacturing route. Nevertheless, there are available, irradiation data on many obsolete graphites of known microstructures. There is also a well-developed physical understanding of the process of irradiation damage in graphite. This paper proposes a specification for graphite suitable for modern HTRs. (author)
Brookhaven Reactor Experiment Control Facility, a distributed function computer network

International Nuclear Information System (INIS)

Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

1975-11-01

A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented
Actinides in irradiated graphite of RBMK-1500 reactor

Energy Technology Data Exchange (ETDEWEB)

Plukienė, R., E-mail: rita@ar.fi.lt; Plukis, A.; Barkauskas, V.; Gudelis, A.; Gvozdaitė, R.; Duškesas, G.; Remeikis, V.

2014-10-01

Highlights: • Activation of actinides in the graphite of the RBMK-1500 reactor was analyzed. • Numerical modeling using SCALE 6.1 and MCNPX was used for actinide calculation. • Measurements of the irradiated graphite sample were used for model validation. • Results are important for further decommissioning process of the RBMK type reactors. - Abstract: The activation of graphite in the nuclear power plants is the problem of high importance related with later graphite reprocessing or disposal. The activation of actinide impurities in graphite due to their toxicity determines a particular long term risk to waste management. In this work the activation of actinides in the graphite constructions of the RBMK-1500 reactor is determined by nuclear spectrometry measurements of the irradiated graphite sample from the Ignalina NPP Unit I and by means of numerical modeling using two independent codes SCALE 6.1 (using TRITON-VI sequence) and MCNPX (v2.7 with CINDER). Both models take into account the 3D RBMK-1500 reactor core fragment with explicit graphite construction including a stack and a sleeve but with a different simplification level concerning surrounding graphite and construction of control roads. The verification of the model has been performed by comparing calculated and measured isotope ratios of actinides. Also good prediction capabilities of the actinide activation in the irradiated graphite have been found for both calculation approaches. The initial U impurity concentration in the graphite model has been adjusted taking into account the experimental results. The specific activities of actinides in the irradiated RBMK-1500 graphite constructions have been obtained and differences between numerical simulation results, different structural parts (sleeve and stack) as well as comparison with previous results (Ancius et al., 2005) have been discussed. The obtained results are important for further decommissioning process of the Ignalina NPP and other RBMK
Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

Energy Technology Data Exchange (ETDEWEB)

Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

2003-02-01

This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)
Irradiation test plan of oxidation-resistant graphite in WWR-K Research Reactor

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Sakaba, Nariaki; Osaki, Hirotaka; Kato, Hideki; Fujitsuka, Kunihiro; Muto, Takenori; Gizatulin, Shamil; Shaimerdenov, Asset; Dyussambayev, Daulet; Chakrov, Petr

2014-01-01

Graphite materials are used for the in-core components of High Temperature Gas-cooled Reactor (HTGR) which is a graphite-moderated and helium gas-cooled reactor. In the case of air ingress accident in HTGR, SiO_2 protective layer is formed on the surface of SiC layer in TRISO CFP and oxidation of SiC does not proceed and fission products are retained inside the fuel particle. A new safety concept for the HTGR, called Naturally Safe HTGR, has been recently proposed. To enhance the safety of Naturally Safe HTGR ultimately, it is expected that oxidation-resistant graphite is used for graphite components to prevent the TRISO CFPs and fuel compacts from failure. SiC coating is one of candidate methods for oxidation-resistant graphite. JAEA and four graphite companies launched R&Ds to develop the oxidation-resistant graphite and the International Science and Technology Center (ISTC) partner project with JAEA and INP was launched to investigate the irradiation effects on the oxidation-resistant graphite. To determine grades of the oxidation-resistant graphite which will be adopted as irradiation test, a preliminary oxidation test was carried out. This paper described the results of the preliminary oxidation test, the plan of out-of-pile test, irradiation test and post-irradiation test (PIE) of the oxidation-resistant graphite. The results of the preliminary oxidation test showed that the integrity of the oxidation resistant graphite was confirmed and that all of grades used in the preliminary test can be adopted as the irradiation test. Target irradiation temperature was determined to be 1473 (K) and neutron fluence was determined to be from 0.54 × 10"2"5through 1.4 × 10"2"5 (/m"2, E>0.18MeV). Weight change, oxidation rate, activation energy, surface condition, etc. will be evaluated in out-of-pile test and weight change, irradiation effect on oxidation rate and activation energy, surface condition, etc. will be evaluated in PIE. (author)
Brookhaven highlights. Report on research, October 1, 1992--September 30, 1993

Energy Technology Data Exchange (ETDEWEB)

Rowe, M.S.; Belford, M.; Cohen, A.; Greenberg, D.; Seubert, L. [eds.

1993-12-31

This report highlights the research activities of Brookhaven National Laboratory during the period dating from October 1, 1992 through September 30, 1993. There are contributions to the report from different programs and departments within the laboratory. These include technology transfer, RHIC, Alternating Gradient Synchrotron, physics, biology, national synchrotron light source, applied science, medical science, advanced technology, chemistry, reactor physics, safety and environmental protection, instrumentation, and computing and communications.
Brookhaven highlights

International Nuclear Information System (INIS)

Rowe, M.S.; Belford, M.; Cohen, A.; Greenberg, D.; Seubert, L.

1993-01-01

This report highlights the research activities of Brookhaven National Laboratory during the period dating from October 1, 1992 through September 30, 1993. There are contributions to the report from different programs and departments within the laboratory. These include technology transfer, RHIC, Alternating Gradient Synchrotron, physics, biology, national synchrotron light source, applied science, medical science, advanced technology, chemistry, reactor physics, safety and environmental protection, instrumentation, and computing and communications
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
Graphite development for gas-cooled reactors in the USA

International Nuclear Information System (INIS)

Burchell, T.D.

1991-01-01

This document discusses Modular High-Temperature Gas-Cooled Reactor (MHTGR) graphite activities in the USA which currently include the following research and development tasks: coke examination; effects of irradiation; variability of physical properties (mechanical, thermal-physical, and fracture); fatigue behavior, oxidation behavior; NDE techniques; structural design criteria; and carbon-carbon composite control rod clad materials. These tasks support nuclear grade graphite manufacturing technology including nondestructive examination of billets and components. Moreover, data shall be furnished to support design and licensing of graphite components for the MHTGR
OPTIMIZATION OF THE EPITHERMAL NEUTRON BEAM FOR BORON NEUTRON CAPTURE THERAPY AT THE BROOKHAVEN MEDICAL RESEARCH REACTOR.

Energy Technology Data Exchange (ETDEWEB)

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

2002-08-18

Clinical trials of Boron Neutron Capture Therapy for patients with malignant brain tumor had been carried out for half a decade, using an epithermal neutron beam at the Brookhaven's Medical Reactor. The decision to permanently close this reactor in 2000 cut short the efforts to implement a new conceptual design to optimize this beam in preparation for use with possible new protocols. Details of the conceptual design to produce a higher intensity, more forward-directed neutron beam with less contamination from gamma rays, fast and thermal neutrons are presented here for their potential applicability to other reactor facilities. Monte Carlo calculations were used to predict the flux and absorbed dose produced by the proposed design. The results were benchmarked by the dose rate and flux measurements taken at the facility then in use.
Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing
Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

Energy Technology Data Exchange (ETDEWEB)

Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L.; Kuper, J.B.H.

1991-01-01

This report highlights Brookhaven National Laboratory's activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed. (GHH)
Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

Energy Technology Data Exchange (ETDEWEB)

Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L.; Kuper, J.B.H.

1991-12-31

This report highlights Brookhaven National Laboratory`s activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed. (GHH)
Brookhaven highlights for fiscal year 1991, October 1, 1990--September 30, 1991

International Nuclear Information System (INIS)

Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L.; Kuper, J.B.H.

1991-01-01

This report highlights Brookhaven National Laboratory's activities for fiscal year 1991. Topics from the four research divisions: Computing and Communications, Instrumentation, Reactors, and Safety and Environmental Protection are presented. The research programs at Brookhaven are diverse, as is reflected by the nine different scientific departments: Accelerator Development, Alternating Gradient Synchrotron, Applied Science, Biology, Chemistry, Medical, National Synchrotron Light Source, Nuclear Energy, and Physics. Administrative and managerial information about Brookhaven are also disclosed
Brookhaven highlights, October 1979-September 1980

International Nuclear Information System (INIS)

1980-01-01

Highlights are given for the research areas of the Brookhaven National Laboratory. These areas include high energy physics, physics and chemistry, life sciences, applied energy science (energy and environment, and nuclear energy), and support activities (including mathematics, instrumentation, reactors, and safety)
Graphite moderated reactor for thermoelectric generation

International Nuclear Information System (INIS)

Akazawa, Issei; Yamada, Akira; Mizogami, Yorikata

1998-01-01

Fuel rods filled with cladded fuel particles distributed and filled are buried each at a predetermined distance in graphite blocks situated in a reactor core. Perforation channels for helium gas as coolants are formed to the periphery thereof passing through vertically. An alkali metal thermoelectric power generation module is disposed to the upper lid of a reactor container while being supported by a securing receptacle. Helium gas in the coolant channels in the graphite blocks in the reactor core absorbs nuclear reaction heat, to be heated to a high temperature, rises upwardly by the reduction of the specific gravity, and then flows into an upper space above the laminated graphite block layer. Then the gas collides against a ceiling and turns, and flows down in a circular gap around the circumference of the alkali metal thermoelectric generation module. In this case, it transfers heat to the alkali metal thermoelectric generation module. (I.N.)
Brookhaven highlights, October 1979-September 1980

Energy Technology Data Exchange (ETDEWEB)

1980-01-01

Highlights are given for the research areas of the Brookhaven National Laboratory. These areas include high energy physics, physics and chemistry, life sciences, applied energy science (energy and environment, and nuclear energy), and support activities (including mathematics, instrumentation, reactors, and safety). (GHT)
Installation and testing of an optimized epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR)

Energy Technology Data Exchange (ETDEWEB)

Fairchild, R.G.; Kalef-Ezra, J.; Saraf, S.K.; Fiarman, S.; Ramsey, E.; Wielopolski, L.; Laster, B.; Wheeler, F. (Brookhaven National Lab., Upton, NY (USA); Ioannina Univ. (Greece); Brookhaven National Lab., Upton, NY (USA); State Univ. of New York, Stony Brook, NY (USA). Health Science Center; Brookhaven National Lab., Upton, NY (USA); EG and G Idaho, Inc., Idaho Falls, ID (USA))

1989-01-01

Various calculations indicate that an optimized epithermal neutron beam can be produced by moderating fission neutrons either with a combination of Al and D{sub 2}O, or with Al{sub 2}O{sub 3}. We have designed, installed and tested an Al{sub 2}O{sub 3} moderated epithermal neutron beam at the Brookhaven Medical Research Reactor (BMRR). The epithermal neutron fluence rate of 1.8 {times} 10{sup 9} n/cm{sup 2}-sec produces a peak thermal neutron fluence rate of 1.9 to 2.8 {times} 10{sup 9} n/cm{sup 2}-sec in a tissue equivalent (TE) phantom head, depending on the configuration. Thus a single therapy treatment of 5 {times} 10{sup 12} n/cm{sup 2} can be delivered in 30--45 minutes. All irradiation times are given for a BMRR power of 3 MW, which is the highest power which can be delivered continuously. 18 refs., 8 figs., 4 tabs.
A safety assessment of the use of graphite in nuclear reactors licensed by the US NRC

International Nuclear Information System (INIS)

Schweitzer, D.G.; Gurinsky, D.H.; Kaplan, E.; Sastre, C.

1987-09-01

This report reviews existing literature and knowledge on graphite burning and on stored energy accumulation and releases in order to assess what role, if any, a stored energy release can have in initiating or contributing to hypothetical graphite burning scenarios in research reactors. It also addresses the question of graphite ignition and self-sustained combustion in the event of a loss-of-coolant accident (LOCA). The conditions necessary to initiate and maintain graphite burning are summarized and discussed. From analyses of existing information it is concluded that only stored energy accumulations and releases below the burning temperature (650 0 C) are pertinent. After reviewing the existing knowledge on stored energy it is possible to show that stored energy releases do not occur spontaneously, and that the maximum stored energy that can be released from any reactor containing graphite is a very small fraction of the energy produced during the first few minutes of a burning incident. The conclusions from these analyses are that the potential to initiate or maintain a graphite burning incident is essentially independent of the stored energy in the graphite, and depends on other factors that are unique for these reactors, research reactors, and for Fort St. Vrain. In order to have self-sustained rapid graphite oxidation in any of these reactors, certain necessary conditions of geometry, temperature, oxygen supply, reaction product removal, and a favorable heat balance must be maintained. There is no new evidence associated with either the Windscale Accident or the Chernobyl Accident that indicates a credible potential for a graphite burning accident in any of the reactors considered in this review
Graphite for high-temperature reactors

International Nuclear Information System (INIS)

Hammer, W.; Leushacke, D.F.; Nickel, H.; Theymann, W.

1976-01-01

The different graphites necessary for HTRs are being developed, produced and tested within the Federal German ''Development Programme Nuclear Graphite''. Up to now, batches of the following graphite grades have been manufactured and fully characterized by the SIGRI Company to demonstrate reproducibility: pitch coke graphite AS2-500 for the hexagonal fuel elements and exchangeable reflector blocks; special pitch coke graphite ASI2-500 for reflector blocks of the pebble-bed reactor and as back-up material for the hexagonal fuel elements; graphite for core support columns. The material data obtained fulfill most of the requirements under present specifications. Production of large-size blocks for the permanent side reflector and the core support blocks is under way. The test programme covers all areas important for characterizing and judging HTR-graphites. In-pile testing comprises evaluation of the material for irradiation-induced changes of dimensions, mechanical and thermal properties - including behaviour under temperature cycling and creep behaviour - as well as irradiating fuel element segments and blocks. Testing out-of-pile includes: evaluation of corrosion rates and influence of corrosion on strength; strength measurements; including failure criteria. The test programme has been carried out extensively on the AS2-graphite, and the results obtained show that this graphite is suitable as HTGR fuel element graphite. (author)

Nonlinear seismic analysis of a graphite reactor core

International Nuclear Information System (INIS)

Laframboise, W.L.; Desmond, T.P.

1988-01-01

Design and construction of the Department of Energy's N-Reactor located in Richland, Washington was begun in the late 1950s and completed in the early 1960s. Since then, the reactor core's structural integrity has been under review and is considered by some to be a possible safety concern. The reactor core is moderated by graphite. The safety concern stems from the degradation of the graphite due to the effects of long-term irradiation. To assess the safety of the reactor core when subjected to seismic loads, a dynamic time-history structural analysis was performed. The graphite core consists of 89 layers of numerous graphite blocks which are assembled in a 'lincoln-log' lattice. This assembly permits venting of steam in the event of a pressure tube rupture. However, such a design gives rise to a highly nonlinear structure when subjected to earthquake loads. The structural model accounted for the nonlinear interlayer sliding and for the closure and opening of gaps between the graphite blocks. The model was subjected to simulated earthquake loading, and the time-varying response of selected elements critical to safety were monitored. The analytically predicted responses (displacements and strains) were compared to allowable responses to assess margins of safety. (orig.)
Distribution of the thermal neutron field around the graphite reflector of the Dalat Nuclear Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Binh, Nguyen Duc; Tuan, Nguyen Minh; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

1994-10-01

Thermal neutron flux distributions around the graphite reflector of the Dalat Nuclear Research Reactor are determined by the method for neutron activating Cu foils. The major results are as follows: a/The axial distributions at the inner and outer margins of the graphite reflector have unsymmetrical shapes, similar to axial distributions in the core. There is a dissimilarity between the distribution curves at the inner margin and those at the outer margin of the reflector. b/ The radial distribution on the upper surface of the graphite reflector is measured and is described by the two-group neutron diffusion theory. The maximal value of the curve lies at the position of R{sub m}ax = 22.5 cm. c/ The distribution in the twenty water irradiation holes around the rotary specimen rack is obtained. (author). 3 refs., 5 figs., 1 tab.
Graphite core design in UK reactors

International Nuclear Information System (INIS)

Davies, M.W.

1996-01-01

The cores in the first power producing Magnox reactors in the UK were designed with only a limited amount of information available regarding the anisotropic dimensional change behaviour of Pile Grade graphite. As more information was gained it was necessary to make modifications to the design, some minor, some major. As the cores being built became larger, and with the switch to the Advanced Gas-cooled Reactor (AGR) with its much higher power density, additional problems had to be overcome such as increased dimensional change and radiolytic oxidation by the carbon dioxide coolant. For the AGRs a more isotropic graphite was required, with a lower initial open pore volume and higher strength. Gilsocarbon graphite was developed and was selected for all the AGRs built in the UK. Methane bearing coolants are used to limit radiolytic oxidation. (author). 5 figs
Characterization, treatment and conditioning of radioactive graphite from decommissioning of nuclear reactors

International Nuclear Information System (INIS)

2006-09-01

Graphite has been used as a moderator and reflector of neutrons in more than 100 nuclear power plants and in many research and plutonium-production reactors. It is used primarily as a neutron reflector or neutron moderator, although graphite is also used for other features of reactor cores, such as fuel sleeves. Many of the graphite-moderated reactors are now quite old, with some already shutdown. Therefore radioactive graphite dismantling and the management of radioactive graphite waste are becoming an increasingly important issue for a number of IAEA Member States. Worldwide, there are more than 230 000 tonnes of radioactive graphite which will eventually need to be managed as radioactive waste. Proper management of radioactive graphite waste requires complex planning and the implementation of several interrelated operations. There are two basic options for graphite waste management: (1) packaging of non-conditioned graphite waste with subsequent direct disposal of the waste packages, and (2) conditioning of graphite waste (principally either by incineration or calcination) with separate disposal of any waste products produced, such as incinerator ash. In both cases, the specific properties of graphite - such as Wigner energy, graphite dust explosibility, and radioactive gases released from waste graphite - have a potential impact on the safety of radioactive graphite waste management and need to be carefully considered. Radioactive graphite waste management is not specifically addressed in IAEA publications. Only general and limited information is available in publications dealing with decommissioning of nuclear reactors. This report provides a comprehensive discussion of radioactive graphite waste characterization, handling, conditioning and disposal throughout the operating and decommissioning life cycle. The first draft report was prepared at a meeting on 23-27 February 1998. A technical meeting (TM) was held in October 1999 in coincidence with the Seminar on
Draft of standard for graphite core components in high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

2010-01-01

For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)
Some equipment for graphite research in swimming pool reactors; Quelques dispositifs d'etude du graphite dans les piles piscines

Energy Technology Data Exchange (ETDEWEB)

Seguin, M; Arragon, Ph; Dupont, G; Gentil, J; Tanis, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1964-07-01

The irradiation devices described are used for research concerning reactors of the natural uranium type, moderated by graphite and cooled by carbon dioxide. The devices are generally designed for use in swimming pool reactors. The following points have been particularly studied: - maximum use of the irradiation volume, - use of the simplest technological solutions, - standardization of certain constituent parts. This standardization calls for precision machining and careful assembling; these requirements are also true when a relatively low irradiation temperature is required and the nuclear heating is pronounced. Finally, the design of these devices is suitable for the irradiation of other fissile or non-fissile materials. (authors) [French] Les dispositifs d'irradiation decrits servent aux etudes relatives a la filiere des reacteurs a uranium naturel, moderes au graphite et refroidis par le gaz carbonique. Ils sont generalement concus pour etre utilises dans des piles piscines. L'accent a ete mis sur: - l'utilisation au maximum du volume d'irradiation, - le recours aux solutions technologiques les plus simples, - la standardisation de certaines parties constitutives. Cette standardisation impose un usinage precis et un montage soigne, lesquels sont egalement necessaires lorsqu'on doit obtenir une temperature d'irradiation relativement basse alors que l'echauffement nucleaire est important. Enfin, la conception de ces dispositifs est valable pour irradier d'autres materiaux non fissiles ou fissiles. (auteurs)
Experience of on-site disposal of production uranium-graphite nuclear reactor.

Science.gov (United States)

Pavliuk, Alexander O; Kotlyarevskiy, Sergey G; Bespala, Evgeny V; Zakharova, Elena V; Ermolaev, Vyacheslav M; Volkova, Anna G

2018-04-01

The paper reported the experience gained in the course of decommissioning EI-2 Production Uranium-Graphite Nuclear Reactor. EI-2 was a production Uranium-Graphite Nuclear Reactor located on the Production and Demonstration Center for Uranium-Graphite Reactors JSC (PDC UGR JSC) site of Seversk City, Tomsk Region, Russia. EI-2 commenced its operation in 1958, and was shut down on December 28, 1990, having operated for the period of 33 years all together. The extra pure grade graphite for the moderator, water for the coolant, and uranium metal for the fuel were used in the reactor. During the operation nitrogen gas was passed through the graphite stack of the reactor. In the process of decommissioning the PDC UGR JSC site the cavities in the reactor space were filled with clay-based materials. A specific composite barrier material based on clays and minerals of Siberian Region was developed for the purpose. Numerical modeling demonstrated the developed clay composite would make efficient geological barriers preventing release of radionuclides into the environment. Copyright © 2018 Elsevier Ltd. All rights reserved.
Epithermal neutron beam design for neutron capture therapy at the Power Burst Facility and the Brookhaven Medical Research Reactor

International Nuclear Information System (INIS)

Wheeler, F.J.; Parsons, D.K.; Rushton, B.L.; Nigg, D.W.

1990-01-01

Nuclear design studies have been performed for two reactor-based epithermal neutron beams for cancer treatment by neutron capture therapy (NCT). An intermediate-intensity epithermal beam has been designed and implemented at the Brookhaven Medical Research Reactor (BMRR). Measurements show that the BMRR design predictions for the principal characteristics of this beam are accurate. A canine program for research into the biological effects of NCT is now under way at BMRR. The design for a high-intensity epithermal beam with minimal contamination from undesirable radiation components has been finalized for the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. This design will be implemented when it is determined that human NCT trials are advisable. The PBF beam will exhibit approximately an order of magnitude improvement in absolute epithermal flux intensity over that available in the BMRR, and its angular distribution and spectral characteristics will be more advantageous for NCT. The combined effects of beam intensity, angular distribution, spectrum, and contaminant level allow the desired tumor radiation dose to be delivered in much shorter times than are possible with the currently available BMRR beam, with a significant reduction (factor of 3 to 5) in collateral dose due to beam contaminants
A German research project about applicable graphite cutting techniques

International Nuclear Information System (INIS)

Holland, D.; Quade, U.; Bach, F.W.; Wilk, P.

2001-01-01

In Germany, too, quite large quantities of irradiated nuclear graphite, used in research and prototype reactors, are waiting for an environmental way of disposal. While incineration of nuclear graphite does not seem to be a publicly acceptable way, cutting and packaging into ductile cast iron containers could be a suitable way of disposal in Germany. Nevertheless, the cutting of graphite is also a very difficult technique by which a large amount of secondary waste or dust might occur. An applicable graphite cutting technique is needed. Therefore, a group of 13 German partners, consisting of one university, six research reactor operators, one technical inspection authority, three engineering companies, one industrial cutting specialist and one commercial dismantling company, decided in 1999 to start a research project to develop an applicable technique for cutting irradiated nuclear graphite. Aim of the project is to find the most suitable cutting techniques for the existing shapes of graphite blocks with a minimum of waste production rate. At the same time it will be learned how to sample the dust and collect it in a filter system. The following techniques will be tested and evaluated: thermal cutting, water jet cutting, mechanical cutting with a saw, plasma arc cutting, drilling. The subsequent evaluation will concentrate on dust production, possible irradiation of staff, time and practicability under different constraints. This research project is funded by the German Minister of Education and Research under the number 02 S 7849 for a period of two years. A brief overview about the work to be carried out in the project will be given. (author)
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.
Graphite surveillance in N Reactor

International Nuclear Information System (INIS)

Woodruff, E.M.

1991-09-01

Graphite dimensional changes in N Reactor during its 24 yr operating history are reviewed. Test irradiation results, block measurements, stack profiles, top of reflector motion monitors, and visual observations of distortion are described. 18 refs., 14 figs., 1 tab
How Big Science Came to Long Island: the Birth of Brookhaven Lab (429th Brookhaven Lecture)

International Nuclear Information System (INIS)

Crease, Robert P.

2007-01-01

Robert P. Crease, historian for the U.S. Department of Energy's Brookhaven National Laboratory and Chair of the Philosophy Department at Stony Brook University, will give two talks on the Laboratory's history on October 31 and December 12. Crease's October 31 talk, titled 'How Big Science Came to Long Island: The Birth of Brookhaven Lab,' will cover the founding of the Laboratory soon after World War II as a peacetime facility to construct and maintain basic research facilities, such as nuclear reactors and particle accelerators, that were too large for single institutions to build and operate. He will discuss the key figures involved in starting the Laboratory, including Nobel laureates I.I. Rabi and Norman Ramsey, as well as Donald Dexter Van Slyke, one of the most renowned medical researchers in American history. Crease also will focus on the many problems that had to be overcome in creating the Laboratory and designing its first big machines, as well as the evolving relations of the Laboratory with the surrounding Long Island community and news media. Throughout his talk, Crease will tell fascinating stories about Brookhaven's scientists and their research.
Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

International Nuclear Information System (INIS)

Anfimov, S.S.

2001-01-01

In this paper an radioactive waste processing of graphite from graphite moderated nuclear reactors at its decommissioning is discussed. Methods of processing of irradiated graphite are presented. It can be concluded that advanced methods for graphite radioactive waste handling are available nowadays. Implementation of these methods will allow to enhance environmental safety of nuclear power that will benefit its progress in the future
Conditioning for definitive storage of radioactive graphite bricks from reactor decommissioning

International Nuclear Information System (INIS)

Costes, J.R.; Koch, C.; Tassigny, C. de; Vidal, H.; Raymond, A.

1990-01-01

The decommissioning of gas-graphite reactors in the EC (e.g. French UNGGs, British Magnox reactors and AGRs, and reactors in Spain and in Italy) will produce large amounts of graphite bricks. This graphite cannot be accepted without particular conditioning by the existing shallow land disposal sites. The aim of the study is to examine the behaviour of graphite waste and to develop a conditioning technique which makes this waste acceptable for shallow land disposal sites. 18 kg of graphite core samples with an outside diameter of 74 mm were removed from the G2 gas-cooled reactor at Marcoule. Their radioactivity is highly dependent on the position of the graphite bricks inside the reactor. Measured results indicate an activity range of 100-400 MBq/kg with 90% Tritium, 5% 14 C, 3% 60 Co, 1.5% 63 Ni. Repeated porosity analyses showed that open porosity ranging from 0 to 100 μm exceeded 23 vol% in the graphite. Water penetration kinetics were investigated in unimpregnated graphite and resulted in impregnation by water of 50-90% of the open porosity. Preliminary lixiviation tests on the crude samples showed quick lixidegree of Cs (several per cent) and of 60 Co, and 133 Ba at a lesser degree. The proposed conditioning technique does not involve a simple coating but true impregnation by a tar-epoxy mixture. The bricks recovered intact from the core by robot services will be placed one by one inside a cylindrical metallic container. But this container may corrode and the bricks may become fragmented in the future, the normally porous graphite will be unaffected by leaching since it is proved that all pores larger than 0.1 μm will be filled with the tar-epoxy mixture. This is a true long-term waste packaging concept. The very simple technology required for industrial implementation is discussed
Structural performance of a graphite blanket in fusion reactors

International Nuclear Information System (INIS)

Wolfer, W.G.; Watson, R.D.

1978-01-01

Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design
Graphite stack corrosion of BUGEY-1 reactor (synthesis)

International Nuclear Information System (INIS)

Petit, A.; Brie, M.

1996-01-01

The definitive shutdown date for the BUGEY-1 reactor was May 27th, 1994, after 12.18 full power equivalent years and this document briefly describes some of the feedback of experience from operation of this reactor. The radiolytic corrosion of graphite stack is the major problem for BUGEY-1 reactor, despite the inhibition of the reaction by small quantities of CH 4 added to the coolant gas. The mechanical behaviour of the pile is predicted using the ''INCA'' code (stress calculation), which uses the results of graphite weight loss variation determined using the ''USURE'' code. The weight loss of graphite is determined by annually taking core samples from the channel walls. The results of the last test programme undertaken after the definitive shutdown of BUGEY-1 have enabled an experimental graph to be established showing the evolution of the compression resistance (perpendicular and parallel direction to the extrusion axis) as a function of the weight loss. The numerous analyses, made on the samples carried out in the most sensitive regions, have allowed to verify that no brutal degradation of the mechanical properties of graphite happens for the high value of weight loss up to 40% (maximum weight loss reached locally). (author). 10 refs, 3 figs, 4 tabs
Raw materials for reflector graphite (for reactors)

International Nuclear Information System (INIS)

Wilhelmi, G.; Mindermann, D.

1992-01-01

The manufacturing concept for the core components of German high temperature reactor (HTR) types of graphite was previously entirely directed to the use of German tar coke (St coke). As the plants for producing this material no longer complied technically with the current environmental protection requirements, one had to assume that they would soon be shut down. To prevent bottlenecks in the erection of future HTR plants, alternative cokes produced by modern processes by Japanese manufacturers were checked for their suitability for the manufacture of reactor graphite. This report describes the investigations carried out on these materials from the safe delayed coking process. The project work, apart from analysis of the main data of the candidate coke considered, included the processing of the raw materials into directly and secondarily extruded graphite rods on the laboratory scale, including characterisation. As the results show, the material data achieved with the previous raw material can be reproduced with Japanese St coke. The tar coke LPC-A from the Nippon Steel Chemical Co., Ltd was decided on as the new standard coke for manufacturing reflector graphite. (orig.) With 15 tabs., 2 figs [de
New Brookhaven chief seeks cross-cutting research

CERN Multimedia

Jones, D

2003-01-01

Brookhaven National Laboratory will pursue opportunities for promoting commercial development of energy systems and other technologies while focusing on the lab's primary mission of basic science research, according to the incoming BNL director, Praveen Chaudhari (1 page).
Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

International Nuclear Information System (INIS)

Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

2005-01-01

G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)
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.

Developments in natural uranium - graphite reactors

International Nuclear Information System (INIS)

Bourgeois, J.

1964-01-01

The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4
Management of radioactive waste in nuclear power: handling of irradiated graphite from water-cooled graphite reactors

International Nuclear Information System (INIS)

Anfimov, S.S.

2000-01-01

As a result of decommissioning of water-cooled graphite-moderated reactors, a large amount of rad-waste in the form of graphite stack fragments is generated (on average 1500-2000 tons per reactor). That is why it is essentially important, although complex from the technical point of view, to develop advanced technologies based on up-to-date remotely-controlled systems for unmanned dismantling of the graphite stack containing highly-active long-lived radionuclides and for conditioning of irradiated graphite (IG) for the purposes of transportation and subsequent long term and ecologically safe storage either on NPP sites or in special-purpose geological repositories. The main characteristics critical for radiation and nuclear hazards of the graphite stack are as follows: the graphite stack is contaminated with nuclear fuel that has gotten there as a result of the accidents; the graphite mass is 992 tons, total activity -6?104 Ci (at the time of unit shutdown); the fuel mass in the reactor stack amounts to 100-140 kg, as estimated by IPPE and RDIPE, respectively; γ-radiation dose rate in the stack cells varies from 4 to 4300 R/h, with the prevailing values being in the range from 50 to 100 R/h. In this paper the traditional methods of rad-waste handling as bituminization technology, cementing technology are discussed. In terms of IG handling technology two lines were identified: long-term storage of conditioned IG and IG disposal by means of incineration. The specific cost of graphite immobilization in a radiation-resistant polymeric matrix amounts to -2600 USD per 1 t of graphite, whereas the specific cost of immobilization in slag-stone containers with an inorganic binder (cement) is -1400 USD per 1 t of graphite. On the other hand, volume of conditioned IG rad-waste subject for disposal, if obtained by means of the first technology, is 2-2.5 times less than the volume of rad-waste generated by means of the second technology. It can be concluded from the above that
Management of graphite material: a key issue for High Temperature Gas Reactor system (HTGR)

International Nuclear Information System (INIS)

Bourdeloie, C.; Marimbeau, P.; Robin, J.C.; Cellier, F.

2005-01-01

Graphite material is used in nuclear High Temperature Gas-cooled Reactors (HTGR, Fig.1) as moderator, thermal absorber and also as structural components of the core (Fig.2). This type of reactor was selected by the Generation IV forum as a potential high temperature provider for supplying hydrogen production plants and is under development in France in the frame of the AREVA ANTARES program. In order to select graphite grades to be used in these future reactors, the requirements for mechanical, thermal, physical-chemical properties must match the internal environment of the nuclear core, especially with regard to irradiation effect. Another important aspect that must be addressed early in design is the waste issue. Indeed, it is necessary to reduce the amount of nuclear waste produced by operation of the reactor during its lifetime. Preliminary assessment of the nuclear waste output for an ANTARES type 280 MWe HTGR over 60 year-lifetime gives an estimated 6000 m 3 of activated graphite waste. Thus, reducing the graphite waste production is an important issue for any HTGR system. First, this paper presents a preliminary inventory of graphite waste fluxes coming from a HTGR, in mass and volume, with magnitudes of radiological activities based on activation calculations of graphite during its stay in the core of the reactor. Normalized data corresponding to an output of 1 GWe.year electricity allows comparison of the waste production with other nuclear reactor systems. Second, possible routes to manage irradiated graphite waste are addressed in both the context of French nuclear waste management rules and by comparison to other national regulations. Routes for graphite waste disposal studied in different countries (concerning existing irradiated graphite waste) will be discussed with regard to new issues of large graphite waste from HTGR. Alternative or complementary solutions aiming at lowering volume of graphite waste to be managed will be presented. For example
Graphite-water steam-generating reactor in the USSR

Energy Technology Data Exchange (ETDEWEB)

Dollezhal, N A [AN SSSR, Moscow

1981-10-01

One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.
Removal of 14C from Irradiated Graphite for Graphite Recycle and Waste Volume Reduction

International Nuclear Information System (INIS)

Dunzik-Gougar, Mary Lou; Windes, Will; Marsden, Barry

2014-01-01

The aim of the research presented here was to identify the chemical form of 14 C in irradiated graphite. A greater understanding of the chemical form of this longest-lived isotope in irradiated graphite will inform not only management of legacy waste, but also development of next generation gas-cooled reactors. Approximately 250,000 metric tons of irradiated graphite waste exists worldwide, with the largest single quantity originating in the Magnox and AGR reactors of UK. The waste quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation I gas-cooled, graphite moderated reactors. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 14 C, with a half-life of 5730 years.
Water cooled reactor technology: Safety research abstracts no. 1

International Nuclear Information System (INIS)

1990-01-01

The Commission of the European Communities, the International Atomic Energy Agency and the Nuclear Energy Agency of the OECD publish these Nuclear Safety Research Abstracts within the framework of their efforts to enhance the safety of nuclear power plants and to promote the exchange of research information. The abstracts are of nuclear safety related research projects for: pressurized light water cooled and moderated reactors (PWRs); boiling light water cooled and moderated reactors (BWRs); light water cooled and graphite moderated reactors (LWGRs); pressurized heavy water cooled and moderated reactors (PHWRs); gas cooled graphite moderated reactors (GCRs). Abstracts of nuclear safety research projects for fast breeder reactors are published independently by the Nuclear Energy Agency of the OECD and are not included in this joint publication. The intention of the collaborating international organizations is to publish such a document biannually. Work has been undertaken to develop a common computerized system with on-line access to the stored information
Research reactors; Les piles de recherche

Energy Technology Data Exchange (ETDEWEB)

Kowarski, L. [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires]|[Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

1955-07-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing
Research reactors; Les piles de recherche

Energy Technology Data Exchange (ETDEWEB)

Kowarski, L [Commissariat a l' Energie Atomique, Paris (France). Centre d' Etudes Nucleaires; [Organisation europeenne pour la Recherche Nucleaire, Geneve (Switzerland)

1955-07-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing
Brookhaven highlights

International Nuclear Information System (INIS)

Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L.

1992-01-01

This publication provides a broad overview of the research programs and efforts being conducted, built, designed, and planned at Brookhaven National Laboratory. This work covers a broad range of scientific disciplines. Major facilities include the Alternating Gradient Synchrotron (AGS), with its newly completed booster, the National Synchrotron Light Source (NSLS), the High Flux Beam Reactor (HFBR), and the RHIC, which is under construction. Departments within the laboratory include the AGS department, accelerator development, physics, chemistry, biology, NSLS, medical, nuclear energy, and interdepartmental research efforts. Research ranges from the pure sciences, in nuclear physics and high energy physics as one example, to environmental work in applied science to study climatic effects, from efforts in biology which are a component of the human genome project to the study, production, and characterization of new materials. The paper provides an overview of the laboratory operations during 1992, including staffing, research, honors, funding, and general laboratory plans for the future
Brookhaven highlights

Energy Technology Data Exchange (ETDEWEB)

Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L. (eds.)

1992-01-01

This publication provides a broad overview of the research programs and efforts being conducted, built, designed, and planned at Brookhaven National Laboratory. This work covers a broad range of scientific disciplines. Major facilities include the Alternating Gradient Synchrotron (AGS), with its newly completed booster, the National Synchrotron Light Source (NSLS), the High Flux Beam Reactor (HFBR), and the RHIC, which is under construction. Departments within the laboratory include the AGS department, accelerator development, physics, chemistry, biology, NSLS, medical, nuclear energy, and interdepartmental research efforts. Research ranges from the pure sciences, in nuclear physics and high energy physics as one example, to environmental work in applied science to study climatic effects, from efforts in biology which are a component of the human genome project to the study, production, and characterization of new materials. The paper provides an overview of the laboratory operations during 1992, including staffing, research, honors, funding, and general laboratory plans for the future.
Tests for removal of Co-60 and Eu-154 from irradiated graphite in the TRIGA Reactor

International Nuclear Information System (INIS)

Arsene, Carmen

2009-01-01

The irradiated graphite in Romania is mainly generated in the thermal columns of TRIGA and WWER-S research reactors (about 9 tones). It was found that the radionuclide content of the graphite irradiated in the TRIGA research reactor is mainly due to C-14 (103 Bq/g), Eu-152 (600-700 Bq/g) and Co-60 (130-150 Bq/g) and low amounts of Eu-154 and Cs-137, depending on location in the thermal column and on irradiation history. In order to minimize the waste inventory and volume in view of their final disposal, in the present paper we show the results of experiments performed for developing and optimizing methods for the chemical decontamination of the irradiated graphite. These procedures are based on strong alkaline solutions for Eu-152 and strong acid solutions for Co-60. The influence of the process parameters on the decontamination factor is investigated. (authors)
Nuclear graphite wear properties and estimation of graphite dust production in HTR-10

Energy Technology Data Exchange (ETDEWEB)

Luo, Xiaowei, E-mail: xwluo@tsinghua.edu.cn; Wang, Xiaoxin; Shi, Li; Yu, Xiaoyu; Yu, Suyuan

2017-04-15

Highlights: • Graphite dust. • The wear properties of graphite. • Pebble bed. • High Temperature Gas-cooled Reactor. • Fuel element. - Abstract: The issue of the graphite dust has been a research focus for the safety of High Temperature Gas-cooled Reactors (HTGRs), especially for the pebble bed reactors. Most of the graphite dust is produced from the wear of fuel elements during cycling of fuel elements. However, due to the complexity of the motion of the fuel elements in the pebble bed, there is no systematic method developed to predict the amount the graphite dust in a pebble bed reactor. In this paper, the study of the flow of the fuel elements in the pebble bed was carried out. Both theoretical calculation and numerical analysis by Discrete Element Method (DEM) software PFC3D were conducted to obtain the normal forces and sliding distances of the fuel elements in pebble bed. The wearing theory was then integrated with PFC3D to estimate the amount of the graphite dust in a pebble bed reactor, 10 MW High Temperature gas-cooled test Reactor (HTR-10).
Temperature distribution in graphite during annealing in air cooled reactors

International Nuclear Information System (INIS)

Oliveira Avila, C.R. de.

1989-01-01

A model for the evaluation temperature distributions in graphite during annealing operation in graphite. Moderated an-cooled reactors, is presented. One single channel and one dimension for air and graphite were considered. A numerical method based on finite control volumes was used for partioning the mathematical equations. The problem solution involves the use of unsteady equations of mass, momentum and energy conservation for air, and energy conservation for graphite. The source term was considered as stored energy release during annealing for describing energy conservation in the graphite. The coupling of energy conservation equations in air and graphite is performed by the heat transfer term betwen air and graphite. The results agree with experimental data. A sensitivity analysis shown that the termal conductivity of graphite and the maximum inlet channel temperature have great effect on the maximum temperature reached in graphite during the annealing. (author)
Fracture toughness testing of a reactor grade graphite

Energy Technology Data Exchange (ETDEWEB)

Roeding, M.; Klein, G.; Schiffers, H.; Nickel, H.

1976-03-15

Fracture mechanics is a well established tool for the assessment of brittle fracture in metallic structural materials. In this paper an attempt is made to apply fracture mechanics to a reactor-grade graphite. The effect of several test parameters on the stress intensity factor was measured; this was found to lie in the range 25 and 50 N/mm/sup -3/2/. The results are discussed in terms of the well known mechanical characteristics of graphite.
Study of new structures adapted to gas-graphite and gas-heavy water reactors

International Nuclear Information System (INIS)

Martin, R.; Roche, R.

1964-01-01

The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [fr
Assessment of different mechanisms of C-14 production in irradiated graphite of RBMK-1500 reactors

International Nuclear Information System (INIS)

Narkunas, Ernestas; Smaizys, Arturas; Poskas, Povilas; Kilda, Raimondas

2010-01-01

Two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at the Ignalina Nuclear Power Plant (INPP) are under decommissioning now. The total mass of irradiated graphite in the cores of both units is more than 3600 tons. The main source of uncertainty in the numerical assessment of graphite activity is the uncertainty of the initial impurities content in graphite. Nitrogen is one of the most important impurities, having a large neutron capture cross-section. This impurity may become the dominant source of C-14 production. RBMK reactors graphite stacks operate in the cooling mixture of helium-nitrogen gases and this may additionally increase the quantity of the nitrogen impurity. In this paper the results of the numerical modelling of graphite activation for the INPP Unit I reactor are presented. In order to evaluate the C-14 activity dependence on the nitrogen impurity content, several cases with different nitrogen content were modelled taking into account initial nitrogen impurity quantities in the graphite matrix and possible nitrogen quantities entrapped in the graphite pores from cooling gases. (orig.)
US graphite reactor D ampersand D experience

International Nuclear Information System (INIS)

Garrett, S.M.K.; Williams, N.C.

1997-02-01

This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE)
Project accent: graphite irradiated creep in a materials test reactor

International Nuclear Information System (INIS)

Brooking, M.

2014-01-01

Atkins manages a pioneering programme of irradiation experiments for EDF Energy. One of these projects is Project ACCENT, designed to obtain evidence of a beneficial physical property of the graphite, which may extend the life of the Advanced Gas-cooled Reactors (AGRs). The project team combines the in-house experience of EDF Energy with two supplier organisations (providing the material test reactors and testing facilities) and supporting consultancies (Atkins and an independent technical expert). This paper describes: - Brief summary of the Project; - Discussion of the challenges faced by the Project; and - Conclusion elaborating on the aims of the Project. These challenging experiments use bespoke technology and both un-irradiated (virgin) and irradiated AGR graphite. The results will help to better understand graphite irradiation-induced creep (or stress modified dimensional change) properties and therefore more accurately determine lifetime and safe operating envelopes of the AGRs. The first round of irradiation has been completed, with a second round about to commence. This is a key step to realising the full lifetime ambition for AGRs, demonstrating the relaxation of stresses within the graphite bricks. (authors)
Deuterium migration in nuclear graphite: consequences for the behavior of tritium in Gas Cooled Reactors and for the decontamination of irradiated graphite waste

International Nuclear Information System (INIS)

Le-Guillou, Mael

2014-01-01

In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2 H( 3 He,p) 4 He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300 C, and would be more efficient in dry inert gas than in humid gas. (author)
The use of graphite for the reduction of void reactivity in CANDU reactors

International Nuclear Information System (INIS)

Min, B.J.; Kim, B.G.; Sim, K-S.

1995-01-01

Coolant void reactivity can be reduced by using burnable poison in CANDU reactors. The use of graphite in the fuel bundle is introduced to reduce coolant void reactivity by adding an appropriate amount of burnable poison in the central rod. This study shows that sufficiently low void reactivity which in controllable by Reactor Regulating System (RRS) can be achieved by using graphite used fuel with slightly enriched uranium. Zero void reactivity can be also obtained by using graphite used fuel with a large central rod. A new fuel bundle with graphite rods can substantially reduce the void reactivity with less burnup penalty compared to previously proposed low void reactivity fuel with depleted uranium. (author)

Thermodynamic Simulation of Equilibrium Composition of Reaction Products at Dehydration of a Technological Channel in a Uranium-Graphite Reactor

Science.gov (United States)

Pavliuk, A. O.; Zagumennov, V. S.; Kotlyarevskiy, S. G.; Bespala, E. V.

2018-01-01

The problems of accumulation of nuclear fuel spills in the graphite stack in the course of operation of uranium-graphite nuclear reactors are considered. The results of thermodynamic analysis of the processes in the graphite stack at dehydration of a technological channel, fuel element shell unsealing and migration of fission products, and activation of stable nuclides in structural elements of the reactor and actinides inside the graphite moderator are given. The main chemical reactions and compounds that are produced in these modes in the reactor channel during its operation and that may be hazardous after its shutdown and decommissioning are presented. Thermodynamic simulation of the equilibrium composition is performed using the specialized code TERRA. The results of thermodynamic simulation of the equilibrium composition in different cases of technological channel dehydration in the course of the reactor operation show that, if the temperature inside the active core of the nuclear reactor increases to the melting temperature of the fuel element, oxides and carbides of nuclear fuel are produced. The mathematical model of the nonstationary heat transfer in a graphite stack of a uranium-graphite reactor in the case of the technological channel dehydration is presented. The results of calculated temperature evolution at the center of the fuel element, the replaceable graphite element, the air gap, and in the surface layer of the block graphite are given. The numerical results show that, in the case of dehydration of the technological channel in the uranium-graphite reactor with metallic uranium, the main reaction product is uranium dioxide UO2 in the condensed phase. Low probability of production of pyrophoric uranium compounds (UH3) in the graphite stack is proven, which allows one to disassemble the graphite stack without the risk of spontaneous graphite ignition in the course of decommissioning of the uranium-graphite nuclear reactor.
Brookhaven highlights, October 1978-September 1979

International Nuclear Information System (INIS)

1979-01-01

These highlights present an overview of the major research and development achievements at Brookhaven National Laboratory from October 1978 to September 1979. Specific areas covered include: accelerator and high energy physics programs; high energy physics research; the AGS and improvements to the AGS; neutral beam development; heavy ion fusion; superconducting power cables; ISABELLE storage rings; the BNL Tandem accelerator; heavy ion experiments at the Tandem; the High Flux Beam Reactor; medium energy physics; nuclear theory; atomic and applied physics; solid state physics; neutron scattering studies; x-ray scattering studies; solid state theory; defects and disorder in solids; surface physics; the National Synchrotron Light Source ; Chemistry Department; Biology Department; Medical Department; energy sciences; environmental sciences; energy technology programs; National Center for Analysis of Energy Systems; advanced reactor systems; nuclear safety; National Nuclear Data Center; nuclear materials safeguards; Applied Mathematics Department; and support activities
Necessity of research reactors

International Nuclear Information System (INIS)

Ito, Tetsuo

2016-01-01

Currently, only three educational research reactors at two universities exist in Japan: KUR, KUCA of Kyoto University and UTR-KINKI of Kinki University. UTR-KINKI is a light-water moderated, graphite reflected, heterogeneous enriched uranium thermal reactor, which began operation as a private university No. 1 reactor in 1961. It is a low power nuclear reactor for education and research with a maximum heat output of 1 W. Using this nuclear reactor, researches, practical training, experiments for training nuclear human resources, and nuclear knowledge dissemination activities are carried out. As of October 2016, research and practical training accompanied by operation are not carried out because it is stopped. The following five items can be cited as challenges faced by research reactors: (1) response to new regulatory standards and stagnation of research and education, (2) strengthening of nuclear material protection and nuclear fuel concentration reduction, (3) countermeasures against aging and next research reactor, (4) outflow and shortage of nuclear human resources, and (5) expansion of research reactor maintenance cost. This paper would like to make the following recommendations so that we can make contribution to the world in the field of nuclear power. (1) Communication between regulatory authorities and business operators regarding new regulatory standards compliance. (2) Response to various problems including spent fuel measures for long-term stable utilization of research reactors. (3) Personal exchanges among nuclear experts. (4) Expansion of nuclear related departments at universities to train nuclear human resources. (5) Training of world-class nuclear human resources, and succession and development of research and technologies. (A.O.)
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
Management of UKAEA graphite liabilities

International Nuclear Information System (INIS)

Wise, M.

2001-01-01

The UK Atomic Energy Authority (UKAEA) is responsible for managing its liabilities for redundant research reactors and other active facilities concerned with the development of the UK nuclear technology programme since 1947. These liabilities include irradiated graphite from a variety of different sources including low irradiation temperature reactor graphite (the Windscale Piles 1 and 2, British Energy Pile O and Graphite Low Energy Experimental Pile at Harwell and the Material Testing Reactors at Harwell and Dounreay), advanced gas-cooled reactor graphite (from the Windscale Advanced Gas-cooled Reactor) and graphite from fast reactor systems (neutron shield graphite from the Dounreay Prototype Fast Reactor and Dounreay Fast Reactor). The decommissioning and dismantling of these facilities will give rise to over 6,000 tonnes of graphite requiring disposal. The first graphite will be retrieved from the dismantling of Windscale Pile 1 and the Windscale Advanced Gas-cooled Reactor during the next five years. UKAEA has undertaken extensive studies to consider the best practicable options for disposing of these graphite liabilities in a manner that is safe whilst minimising the associated costs and technical risks. These options include (but are not limited to), disposal as Low Level Waste, incineration, or encapsulation and disposal as Intermediate Level Waste. There are a number of technical issues associated with each of these proposed disposal options; these include Wigner energy, radionuclide inventory determination, encapsulation of graphite dust, galvanic coupling interactions enhancing the corrosion of mild steel and public acceptability. UKAEA is currently developing packaging concepts and designing packaging plants for processing these graphite wastes in consultation with other holders of graphite wastes throughout Europe. 'Letters of Comfort' have been sought from both the Low Level Waste and the Intermediate Level Waste disposal organisations to support the
Studies on design principles and criteria of fuels and graphites for experimental multi-purpose very high temperature reactor

International Nuclear Information System (INIS)

Arai, Taketoshi; Sato, Sadao; Tani, Yutaro

1977-12-01

Design principles and criteria of fuels and graphites have been studied to determine the main design parameters of a reference core MARK-III of the Experimental Multi-purpose Very High Temperature Reactor. The present status of research and development for HTGR fuels and graphites is reviewed from a standpoint of their integrity and safety aspects, and is compared to the specific design requirements for the VHTR fuels and graphites. Consequently, reasonable materials specifications, safety criteria and design analysis methods are presented for coated fuel particle, fuel compact, graphite sleeve, core support graphite and neutron absorber material. These design principles and criteria will be refined by further experimental investigations. (auth.)
Assessment of management modes for graphite from reactor decommissioning

International Nuclear Information System (INIS)

White, I.F.; Smith, G.M.; Saunders, L.J.; Kaye, C.J.; Martin, T.J.; Clarke, G.H.; Wakerley, M.W.

1984-01-01

A technological and radiological assessment has been made of the management options for irradiated graphite wastes from the decommissioning of Magnox and advanced gas-cooled reactors. Detailed radionuclide inventories have been estimated, the main contribution being from activation of the graphite and its stable impurities. Three different packaging methods for graphite have been described; each could be used for either sea or land disposal, is logistically feasible and could be achieved at reasonable cost. Leaching tests have been carried out on small samples of irradiated graphite under a variety of conditions including those of the deep ocean bed; the different conditions had little effect on the observed leach rates of radiologically significant radionuclides. Radiological assessments were made of four generic options for disposal of packaged graphite: on the deep ocean bed, in deep geologic repositories at two different types of site, and by shallow land burial. Incineration of graphite was also considered, though this option presents logistical problems. With appropriate precautions during the lifetime of the Cobalt-60 content of the graphite, any of the options considered could give acceptably low doses to individuals, and all would merit further investigation in site-specific contexts
Treatment and Disposal of the Radioactive Graphite Waste of High-Temperature Gas-Cooled Reactor Spent Fuel

International Nuclear Information System (INIS)

Li Junfeng

2016-01-01

High-temperature gas-cooled reactors (HTGRs) represent one of the Gen IV reactors in the future market, with efficient generation of energy and the supply of process heat at high temperature utilised in many industrial processes. HTGR development has been carried out within China’s National High Technology Research and Development Program. The first industrial demonstration HTGR of 200 MWe is under construction in Shandong Province China. HTGRs use ceramic-coated fuel particles that are strong and highly resistant to irradiation. Graphite is used as moderator and helium is used as coolant. The fuel particles and the graphite block in which they are imbedded can withstand very high temperature (up to ~1600℃). Graphite waste presents as the fuel element components of HTGR with up to 95% of the whole element beside the graphite blocks in the core. For example, a 200 MWe reactor could discharge about 90,000 fuel elements with 17 tonnes irradiated graphite included each year. The core of the HTGR in China consists of a pebble bed with spherical fuel elements. The UO 2 fuel kernel particles (0.5mm diameter) (triple-coated isotropic fuel particles) are coated by several layers including inner buffer layer with less dense pyrocarbon, dense pyro-carbon, SiC layer and outer layer of dense pyro-carbon, which can prevent the leaking of fission products (Fig. 1). Spherical fuel elements (60mm diameter) consist of a 50mm diameter inner zone and 5mm thick shell of fuel free zone [3]. The inner zone contains about 8300 triple-coated isotropic fuel particles of 0.92mm in diameter dispersed in the graphite matrix
Brookhaven highlights, fiscal year 1985, October 1, 1984-September 30, 1985

Energy Technology Data Exchange (ETDEWEB)

1985-01-01

Activities at Brookhaven National Laboratory are briefly discussed. These include work at the National Synchrotron Light Source, the High Flux Beam Reactor, and the Alternating Gradient Synchrotron. Areas of research include heavy ion reactions, neutrino oscillations, low-level waste, nuclear data, medicine, biology, chemistry, parallel computing, optics. Also provided are general and administrative news, a financial report. (LEW)
Brookhaven highlights, fiscal year 1985, October 1, 1984-September 30, 1985

International Nuclear Information System (INIS)

1985-01-01

Activities at Brookhaven National Laboratory are briefly discussed. These include work at the National Synchrotron Light Source, the High Flux Beam Reactor, and the Alternating Gradient Synchrotron. Areas of research include heavy ion reactions, neutrino oscillations, low-level waste, nuclear data, medicine, biology, chemistry, parallel computing, optics. Also provided are general and administrative news, a financial report
Study of graphite reactivity worth on well-defined cores assembled on LR-0 reactor

International Nuclear Information System (INIS)

Košťál, Michal; Rypar, Vojtěch; Milčák, Ján; Juříček, Vlastimil; Losa, Evžen; Forget, Benoit; Harper, Sterling

2016-01-01

Highlights: • A light water critical facility for graphite reactivity worth measurements. • Comparison of calculated and measured k eff . • Effect of graphite description on k eff . - Abstract: Graphite is an often-used moderating material on the basis of its good moderating power and very low absorption cross section. This small absorption cross section permits the use of natural or low-enriched uranium in graphite moderated reactors. Graphite is now being considered as the moderator for Fluoride-salt-cooled High Temperature Reactors (FHR). The critical moderator level was measured for various graphite block configurations in an experimental dry assembly of the LR-0 reactor. Comparisons with experiments were performed between Monte Carlo simulation tools for which satisfactory agreement was obtained with the exception of some systematic discrepancies. The larger discrepancies were observed when using the ENDF/B-VII.0 library. To decrease the uncertainties, based on conservative assumptions, relative comparisons were done. The results provided by the different nuclear data libraries are within 3 sigma interval of experimental uncertainties. It has been determined that differences between the results of calculations are caused by variations in the (n,n), (n,n′), (n,g) reactions and also by various angular distributions, while the (n,g) cross section variations play only a minor role for these configurations.
High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

Energy Technology Data Exchange (ETDEWEB)

Mcwilliams, A. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-09-08

This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.
From USA operation experience of industrial uranium-graphite reactors

International Nuclear Information System (INIS)

Burdakov, N.S.

1996-01-01

The review on materials, presented by a group of the USA specialists at the seminar in Moscow on October 9-11, 1995 is considered. The above specialists shared their experience in operation of the Hanford industrial reactors, aimed at plutonium production for atomic bombs. The purpose of the above visit consisted in providing assistance to the Russian specialists by evaluation and modernization of operational conditions safety improvement of the RBMK type reactors. Special attention is paid to the behaviour of the graphite lining and channel tubes with an account of possible channel power interaction with the reactor structural units. The information on the experience of the Hanford reactor operation may be useful for specialists, operating the RBMK type reactors
Principle design and data of graphite components

International Nuclear Information System (INIS)

Ishihara, Masahiro; Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Oku, Tatsuo

2004-01-01

The High Temperature Engineering Test Reactor (HTTR) constructed by Japan Atomic Energy Research Institute (JAERI) is a graphite-moderated and helium-gas-cooled reactor with prismatic fuel elements of hexagonal blocks. The reactor internal structures of the HTTR are mainly made up of graphite components. As well known, the graphite is a brittle material and there were no available design criteria for brittle materials. Therefore, JAERI had to develop the design criteria taking account of the brittle fracture behavior. In this paper, concept and key specification of the developed graphite design criteria is described, and also an outline of the quality control specified in the design criteria is mentioned
A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

2007-11-01

The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.
MODELING THE ELECTROLYTIC DECHLORINATION OF TRICHLOROETHYLENE IN A GRANULAR GRAPHITE-PACKED REACTOR

Science.gov (United States)

A comprehensive reactor model was developed for the electrolytic dechlorination of trichloroethylene (TCE) at a granular-graphite cathode. The reactor model describes the dynamic processes of TCE dechlorination and adsorption, and the formation and dechlorination of all the major...
Calculation of the Thermal State of the Graphite Moderator of the RBMK Reactor

Directory of Open Access Journals (Sweden)

Vorobiev Alexander V.

2017-01-01

Full Text Available This work is devoted to study the temperature field of the graphite stack of the RBMK reactor. In work was analyzed the influence of contact pressure between the components of the masonry on the temperature of the graphite moderator.
Experimental study on air ingress during a primary pipe rupture accident with a graphite reactor core simulator

International Nuclear Information System (INIS)

Takeda, Tetsuaki; Hishida, Makoto; Baba, Shinichi

1991-11-01

When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary cooling system reduces the pressure. After the pressures are balanced between the reactor and the container, air is expected to enter into the reactor core from the breach. It seems to be probable that the graphite structures is oxidized by air. Hence, it is necessary to investigate the air ingress process and the behavior of the generating gases by the oxidation reactions. The previous experimental study is performed on the molecular diffusion and natural convection of the two component gas mixtures using a test model simulating simply the reactor. Objective of the study was to investigate the air ingress process during the early stage of the primary pipe rupture accident. However, since the model did not have any kind of graphite components, the reaction between graphite and oxygen was not simulated. The present model includes the reactor core and the high temperature plenum simulators made of graphite. The major results obtained in the present study are summarized in the followings: (1) The air ingress process with graphite oxidation reaction is similar to that without the reaction qualitatively. (2) When the reactor core simulator is maintained at low temperatures (lower than 450degC), the initiation time of the natural circulation of air is almost equal to that of the natural circulation of nitrogen. On the other hand, when the temperature of the reactor core simulator is high (more than 500degC), the initiation time of the natural circulation of air is earlier than that of nitrogen. (3) When the temperature of the reactor core simulator is higher than 600degC, oxygen is almost dissipated by the graphite structures. When the temperature of the reactor core simulator is below 700degC, carbon dioxide mainly is generated by the oxidation reactions. (author)
BROOKHAVEN

International Nuclear Information System (INIS)

Anon.

1993-01-01

Neutrino physics has been an integral part of the Brookhaven research programme for much of the Laboratory's 46-year history. Milestones have been the determination of the helicity of neutrinos (1958), the establishment of the existence of two kinds of neutrinos (1962) for which Leon Lederman, Mel Schwartz and Jack Steinberger were awarded the 1988 Nobel Physics Prize, the discovery of charmed baryons in the 7' Bubble Chamber in 1975, and the ongoing measurements by Ray Davis and collaborators of solar neutrinos, first reported in 1968. There have also been significant contributions to the understanding of neutral currents in exclusive hadron and electron channels. In addition some of the earliest, and to date best, accelerator limits on electron muon neutrino oscillations are from Brookhaven experiments. The Laboratory is also the 'B' in the IMB underground experiment, built to search for proton decay and which caught neutrinos from the SN 1987a supernova. At present Brookhaven is heavily involved in the Gallex project in the Gran Sasso and recently a new collaboration has received scientific approval for a long baseline experiment to search for muon neutrino oscillations via muon neutrino disappearance
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.

Nuclear graphite ageing and turnaround

International Nuclear Information System (INIS)

Marsden, B.J.; Hall, G.N.; Smart, J.

2001-01-01

Graphite moderated reactors are being operated in many countries including, the UK, Russia, Lithuania, Ukraine and Japan. Many of these reactors will operate well into the next century. New designs of High Temperature Graphite Moderated Reactors (HTRS) are being built in China and Japan. The design life of these graphite-moderated reactors is governed by the ageing of the graphite core due to fast neutron damage, and also, in the case of carbon dioxide cooled reactors by the rate of oxidation of the graphite. Nuclear graphites are polycrystalline in nature and it is the irradiation-induced damage to the individual graphite crystals that determines the material property changes with age. The life of a graphite component in a nuclear reactor can be related to the graphite irradiation induced dimensional changes. Graphites typically shrink with age, until a point is reached where the shrinkage stops and the graphite starts to swell. This change from shrinkage to swelling is known as ''turnaround''. It is well known that pre-oxidising graphite specimens caused ''turnaround'' to be delayed, thus extending the life of the graphite, and hence the life of the reactor. However, there was no satisfactory explanation of this behaviour. This paper presents a numerical crystal based model of dimensional change in graphite, which explains the delay in ''turnaround'' in the pre-oxidised specimens irradiated in a fast neutron flux, in terms of crystal accommodation and orientation and change in compliance due to radiolytic oxidation. (author)
Studies on the behavior of graphite structures irradiated in the Dragon Reactor. Dragon Project report

Energy Technology Data Exchange (ETDEWEB)

Everett, M. R.; Graham, L. W.; Ridealgh, F.

1971-11-15

Design data for the physical and mechanical property changes which occur in graphite structural and fuel body components irradiated in an HTR are largely obtained from small specimens tested in the laboratory and in materials test reactors. A brief data summary is given. This graphite physics data can be used to predict dimensional changes, internal stress generation and strength changes in the graphite materials of HTR fuel elements irradiated in the Dragon Reactor. In this paper, the results which have been obtained from post-irradiation examination of a number of fuel pins, are compared with prediction.
Research reactors: design, safety requirements and applications

International Nuclear Information System (INIS)

Hassan, Abobaker Mohammed Rahmtalla

2014-09-01

There are two types of reactors: research reactors or power reactors. The difference between the research reactor and energy reactor is that the research reactor has working temperature and fuel less than the power reactor. The research reactors cooling uses light or heavy water and also research reactors need reflector of graphite or beryllium to reduce the loss of neutrons from the reactor core. Research reactors are used for research training as well as testing of materials and the production of radioisotopes for medical uses and for industrial application. The difference is also that the research reactor smaller in terms of capacity than that of power plant. Research reactors produce radioactive isotopes are not used for energy production, the power plant generates electrical energy. In the world there are more than 284 reactor research in 56 countries, operates as source of neutron for scientific research. Among the incidents related to nuclear reactors leak radiation partial reactor which took place in three mile island nuclear near pennsylvania in 1979, due to result of the loss of control of the fission reaction, which led to the explosion emitting hug amounts of radiation. However, there was control of radiation inside the building, and so no occurred then, another accident that lead to radiation leakage similar in nuclear power plant Chernobyl in Russia in 1986, has led to deaths of 4000 people and exposing hundreds of thousands to radiation, and can continue to be effect of harmful radiation to affect future generations. (author)
Graphites and composites irradiations for gas cooled reactor core structures

International Nuclear Information System (INIS)

Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

2008-01-01

Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space
Graphite-moderated and heavy water-moderated spectral shift controlled reactors

International Nuclear Information System (INIS)

Alcala Ruiz, F.

1984-01-01

It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs
Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

2013-01-01

Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite
Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

2013-07-15

Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.
Temperature control of the graphite stack of the reactor RBMK-1500

International Nuclear Information System (INIS)

Lesnoj, S.

1998-01-01

The paper includes general information about RBMK-1500 reactor, construction features and main technical data; graphite moderator stack, temperature channel, thermocouple TXA-1379, its basic technical and metrologic parameters as well as its advantages and disadvantages
Disintegration of graphite matrix from the simulative high temperature gas-cooled reactor fuel element by electrochemical method

International Nuclear Information System (INIS)

Tian Lifang; Wen Mingfen; Li Linyan; Chen Jing

2009-01-01

Electrochemical method with salt as electrolyte has been studied to disintegrate the graphite matrix from the simulative high temperature gas-cooled reactor fuel elements. Ammonium nitrate was experimentally chosen as the appropriate electrolyte. The volume average diameter of disintegrated graphite fragments is about 100 μm and the maximal value is less than 900 μm. After disintegration, the weight of graphite is found to increase by about 20% without the release of a large amount of CO 2 probably owing to the partial oxidation to graphite in electrochemical process. The present work indicates that the improved electrochemical method has the potential to reduce the secondary nuclear waste and is a promising option to disintegrate graphite matrix from high temperature gas-cooled reactor spent fuel elements in the head-end of reprocessing.
Deuterium migration in nuclear graphite: Consequences for the behavior of tritium in CO{sub 2}-cooled reactors and for the decontamination of irradiated graphite waste

Energy Technology Data Exchange (ETDEWEB)

Le Guillou, M. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Agence nationale pour la gestion des déchets radioactifs, DRD/CM – 1-7, rue Jean Monnet, Parc de la Croix-Blanche, F-92298 Châtenay-Malabry cedex (France); Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); CEA/DEN – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France); Pipon, Y. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Institut Universitaire Technologique, Université Claude Bernard Lyon 1, Université de Lyon – 43, boulevard du 11 novembre 1918, F-69622 Villeurbanne cedex (France); Moncoffre, N. [Institut de Physique Nucléaire de Lyon, CNRS/IN2P3 UMR 5822, Université Claude Bernard Lyon 1, Université de Lyon – 4, rue Enrico Fermi, F-69622 Villeurbanne cedex (France); Khodja, H. [Laboratoire d’Etude des Eléments Légers, CEA/DSM/IRAMIS/NIMBE, UMR 3299 SIS2M – Centre de Saclay, F-91191 Gif-sur-Yvette cedex (France)

2015-06-15

In this paper, we aim at understanding tritium behavior in the graphite moderator of French CO{sub 2}-cooled nuclear fission reactors (called UNGG for “Uranium Naturel-Graphite-Gaz”) to get information on its distribution and inventory in the irradiated graphite waste after their dismantling. These findings should be useful both to improve waste treatment processes and to foresee tritium behavior during reactor decommissioning and waste disposal operations. The purpose of the present work is to elucidate the effects of temperature on the behavior of tritium during reactor operation. Furthermore, it aims at exploring options of thermal decontamination. For both purposes, annealing experiments were carried out in inert atmosphere as well as in thermal conditions as close as possible to those encountered in UNGG reactors and in view of a potential decontamination in humid gas. D{sup +} ions were implanted into virgin nuclear graphite in order to simulate tritium displaced from its original structural site through recoil during reactor operation. The effect of thermal treatments on the mobility of the implanted deuterium was then investigated at temperatures ranging from 200 to 1200 °C, in inert atmosphere (vacuum or argon), in a gas simulating the UNGG coolant gas (mainly CO{sub 2}) or in humid nitrogen. Deuterium was analyzed by Nuclear Reaction Analysis (NRA) both at millimetric and micrometric scales. We have identified three main stages for the deuterium release. The first one corresponds to deuterium permeation through graphite open pores. The second and third ones are controlled by the progressive detrapping of deuterium located at different trapping sites and its successive migration through the crystallites and along crystallites and coke grains edges. Extrapolating the thermal behavior of deuterium to tritium, the results show that the release becomes significant above the maximum UNGG reactor temperature of 500 °C and should be lower than 30% of the
Graphite and carbonaceous materials in a molten salt nuclear reactor

International Nuclear Information System (INIS)

Rousseau, Ginette; Lecocq, Alfred; Hery, Michel.

1982-09-01

A project for a molten salt 1000 MWe reactor is studied by EDF-CEA teams. The design provides for a chromesco 3 vessel housing graphite structures in which the salt circulates. The salt (Th, U, Be and Li fluorides) is cooled by direct contact with lead. The graphites and carbonated materials, inert with respect to lead and the fuel salt, are being considered not only as moderators, but as reflectors and in the construction of the sections where the heat exchange takes place. On the basis of the problems raised in the operation of the reactor, a study programme on French experimental materials (Le Carbone Lorraine, SERS, SEP) has been defined. Hence, depending on the function or functions that the material is to ensure in the structure, the criteria of choice which follow will have to be examined: behaviour under irradiation, insertion of a fluid in the material, thermal properties required, mechanical properties required, utilization [fr
Acceptance test for graphite components and construction status of HTTR

International Nuclear Information System (INIS)

Iyoku, T.; Ishihara, M.; Maruyama, S.; Shiozawa, S.; Tsuji, N.; Miki, T.

1996-01-01

In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs
Graphite moderator annealing of the experimental reactor for irradiation (0.5 MW)

International Nuclear Information System (INIS)

Oliveira Avila, Carlos Alberto de; Pires, Luis Fernando Goncalves

1995-01-01

This work describes an operational procedure for the annealing of the graphite moderator in the 0,5 MW Experimental Reactor for Irradiation. A theoretical methodology has been developed for calculating the temperature field during the annealing process. The equations for mass, momentum, and energy conservation for the coolant as well as for the energy conservation in the moderator are solved numerically. The energy stored in the graphite and released in the annealing is accounted for by the use of a modified source term in the energy conservation equation for the moderator. A good agreement has been found for comparisons of the calculations with annealing data from the BEPO reactor. The major parameters affecting annealing have also been determined. (author). 8 refs, 11 figs
Mechanical, chemical and radiological characterization of the graphite of the UNGG reactors type

International Nuclear Information System (INIS)

Bresard, I.; Bonal, J.P.

2000-01-01

In the framework of UNGG reactors type dismantling procedures, the characterization of the graphite, used as moderator, has to be realized. This paper presents the mechanical, chemical and radiological characterizations, the properties measured and gives some results in the case of the Bugey 1 reactor. (A.L.B.)
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)
Examination of Surface Deposits on Oldbury Reactor Core Graphite to Determine the Concentration and Distribution of 14C.

Directory of Open Access Journals (Sweden)

Liam Payne

Full Text Available Pile Grade A graphite was used as a moderator and reflector material in the first generation of UK Magnox nuclear power reactors. As all of these reactors are now shut down there is a need to examine the concentration and distribution of long lived radioisotopes, such as 14C, to aid in understanding their behaviour in a geological disposal facility. A selection of irradiated graphite samples from Oldbury reactor one were examined where it was observed that Raman spectroscopy can distinguish between underlying graphite and a surface deposit found on exposed channel wall surfaces. The concentration of 14C in this deposit was examined by sequentially oxidising the graphite samples in air at low temperatures (450°C and 600°C to remove the deposit and then the underlying graphite. The gases produced were captured in a series of bubbler solutions that were analysed using liquid scintillation counting. It was observed that the surface deposit was relatively enriched with 14C, with samples originating lower in the reactor exhibiting a higher concentration of 14C. Oxidation at 600°C showed that the remaining graphite material consisted of two fractions of 14C, a surface associated fraction and a graphite lattice associated fraction. The results presented correlate well with previous studies on irradiated graphite that suggest there are up to three fractions of 14C; a readily releasable fraction (corresponding to that removed by oxidation at 450°C in this study, a slowly releasable fraction (removed early at 600°C in this study, and an unreleasable fraction (removed later at 600°C in this study.
Analytical and numerical study of graphite IG110 parts in advanced reactor under high temperature and irradiation

Energy Technology Data Exchange (ETDEWEB)

Gao, Jinling, E-mail: Jinling_Gao@yeah.net; Yao, Wenjuan, E-mail: wj_yao@yeah.net; Ma, Yudong

2016-08-15

Graphical abstract: An analytical model and a numerical procedure are developed to study the mechanical response of IG-110 graphite bricks in HTGR subjected to high temperature and irradiation. The calculation results show great accordance with each other. Rational suggestions on the calculation and design of the IG-110 graphite structure are proposed based on the sensitivity analyses including temperature, irradiation dimensional change, creep and Poisson’s ratio. - Highlights: • Analytical solution of stress and displacement of IG-110 graphite components in HTGR. • Finite element procedure developed for stress analysis of HTGR graphite component. • Parameters analysis of mechanical response of graphite components during the whole life of the reflector. - Abstract: Structural design of nuclear power plant project is an important sub-discipline of civil engineering. Especially after appearance of the fourth generation advanced high temperature gas cooled reactor, structural mechanics in reactor technology becomes a popular subject in structural engineering. As basic ingredients of reflector in reactor, graphite bricks are subjected to high temperature and irradiation and the stress field of graphite structures determines integrity of reflector and makes a great difference to safety of whole structure. In this paper, based on assumptions of elasticity, side reflector is regarded approximately as a straight cylinder structure and primary creep strain is ignored. An analytical study on stress of IG110 graphite parts is present. Meanwhile, a finite element procedure for calculating stresses in the IG110 graphite structure exposed in the high temperature and irradiation is developed. Subsequently, numerical solution of stress in IG110 graphite structure is obtained. Analytical solution agrees well with numerical solution, which indicates that analytical derivation is accurate. Finally, influence of temperature, irradiation dimensional change, creep and Poisson�
University of Missouri research reactor exhaust ventilation/laboratory fume hood upgrade

International Nuclear Information System (INIS)

Edwards, C.B. Jr.; McKibben, J.C.; McCracken, C.B.

1989-01-01

The University of Missouri research reactor (MURR) facility is located in Research Park, 1 mile south of the Columbia campus. The reactor is a 10-MW pressurized loop, in-pool-type, light-water-moderated, beryllium-and-graphite-reflected core, serviced by six radial beam tubes for research, and has sample irradiation facilities in both a flux trap and in the graphite region. The reactor operates at full power 150 h/week, 52 week/yr, making it one of the best operating schedules and the most extensively used of any university research reactor. This extensive utilization includes many programs, such as radioisotope applications, neutron activation analysis, etc., that depend heavily on fume hoods, glove boxes, and hot cells that put a tremendous demand on the exhaust system. The exhaust system is required to be operable whenever the reactor is operating and must have the capability of being operated from an emergency electrical generator on loss of site electrical power. The originally installed exhaust ventilation system was below needed capacity and, with increased program requirements and system age, the necessity to upgrade the system was paramount. The challenge was to complete the upgrade construction while continuing to operate the reactor and maintain all the other ongoing programs, rather than take the easy way of an extended shutdown. This paper discusses how MURR met this challenge and solved these problems, problems that are similarly experienced by almost all research reactors to some degree when major work is required on critical systems
ACCELERATED SITE TECHNOLOGY DEPLOYMENT COST AND PERFORMANCE REPORT COMPARABILITY OF ISOCS INSTRUMENT IN RADIONUCLIDE CHARACTERICATION AT BROOKHAVEN NATIONAL LABORATORY

Energy Technology Data Exchange (ETDEWEB)

KALB,P.; LUCKETT,L.; MILLER,K.; GOGOLAK,C.; MILIAN,L.

2001-03-01

This report describes a DOE Accelerated Site Technology Deployment project being conducted at Brookhaven National Laboratory to deploy innovative, radiological, in situ analytical techniques. The technologies are being deployed in support of efforts to characterize the Brookhaven Graphite Research Reactor (BGRR) facility, which is currently undergoing decontamination and decommissioning. This report focuses on the deployment of the Canberra Industries In Situ Object Counting System (ISOCS) and assesses its data comparability to baseline methods of sampling and laboratory analysis. The battery-operated, field deployable gamma spectrometer provides traditional spectra of counts as a function of gamma energy. The spectra are then converted to radionuclide concentration by applying innovative efficiency calculations using monte carlo statistical methods and pre-defined geometry templates in the analysis software. Measurement of gamma emitting radionuclides has been accomplished during characterization of several BGRR components including the Pile Fan Sump, Above Ground Ducts, contaminated cooling fans, and graphite pile internals. Cs-137 is the predominant gamma-emitting radionuclide identified, with smaller quantities of Co-60 and Am-241 detected. The Project used the Multi-Agency Radiation Survey and Site Investigation Manual guidance and the Data Quality Objectives process to provide direction for survey planning and data quality assessment. Analytical results have been used to calculate data quality indicators (DQI) for the ISOCS measurements. Among the DQIs assessed in the report are sensitivity, accuracy, precision, bias, and minimum detectable concentration. The assessment of the in situ data quality using the DQIs demonstrates that the ISOCS data quality can be comparable to definitive level laboratory analysis when the field instrument is supported by an appropriate Quality Assurance Project Plan. A discussion of the results obtained by ISOCS analysis of
ACCELERATED SITE TECHNOLOGY DEPLOYMENT COST AND PERFORMANCE REPORT COMPARABILITY OF ISOCS INSTRUMENT IN RADIONUCLIDE CHARACTERICATION AT BROOKHAVEN NATIONAL LABORATORY

International Nuclear Information System (INIS)

KALB, P.; LUCKETT, L.; MILLER, K.; GOGOLAK, C.; MILIAN, L.

2001-01-01

This report describes a DOE Accelerated Site Technology Deployment project being conducted at Brookhaven National Laboratory to deploy innovative, radiological, in situ analytical techniques. The technologies are being deployed in support of efforts to characterize the Brookhaven Graphite Research Reactor (BGRR) facility, which is currently undergoing decontamination and decommissioning. This report focuses on the deployment of the Canberra Industries In Situ Object Counting System (ISOCS) and assesses its data comparability to baseline methods of sampling and laboratory analysis. The battery-operated, field deployable gamma spectrometer provides traditional spectra of counts as a function of gamma energy. The spectra are then converted to radionuclide concentration by applying innovative efficiency calculations using monte carlo statistical methods and pre-defined geometry templates in the analysis software. Measurement of gamma emitting radionuclides has been accomplished during characterization of several BGRR components including the Pile Fan Sump, Above Ground Ducts, contaminated cooling fans, and graphite pile internals. Cs-137 is the predominant gamma-emitting radionuclide identified, with smaller quantities of Co-60 and Am-241 detected. The Project used the Multi-Agency Radiation Survey and Site Investigation Manual guidance and the Data Quality Objectives process to provide direction for survey planning and data quality assessment. Analytical results have been used to calculate data quality indicators (DQI) for the ISOCS measurements. Among the DQIs assessed in the report are sensitivity, accuracy, precision, bias, and minimum detectable concentration. The assessment of the in situ data quality using the DQIs demonstrates that the ISOCS data quality can be comparable to definitive level laboratory analysis when the field instrument is supported by an appropriate Quality Assurance Project Plan. A discussion of the results obtained by ISOCS analysis of

Landmarks in particle physics at Brookhaven National Laboratory: Brookhaven Lecture Series, Number 238

International Nuclear Information System (INIS)

Adair, R.K.

1987-01-01

Robert Adair's lecture on Landmarks in Particle Physics at Brookhaven National Laboratory (BNL) is a commemoration of the 40th Anniversary of Brookhaven National Laboratory. Adair describes ten researches in elementary particle physics at Brookhaven that had a revolutionary impact on the understanding of elementary particles. Two of the discoveries were made in 1952 and 1956 at the Cosmotron, BNL's first proton accelerator. Four were made in 1962 and 1964 at the Alternating Gradient Synchrotron, the Cosmotron's replacement. Two other discoveries in 1954 and 1956 were theoretical, and strong focusing (1952) is the only technical discovery. One discovery (1958) happened in an old barrack. Four of the discoveries were awarded the Nobel prize in Physics. Adair believes that all of the discoveries are worthy of the Nobel prize. 14 figs
Fuel elements for high temperature reactors having special suitability for reuse of the structural graphite

International Nuclear Information System (INIS)

Huschka, H.; Herrmann, F.J.

1976-01-01

There are prepared fuel elements for high temperature reactors from which the fuel zone can be removed from the structural graphite after the burnup of the fissile material has taken place so that the fuel element can be filled with new fuel and again placed in the reactor by having the strength of the matrix in the fuel zone sufficient for binding the embedded coated fuel particles but substantially less than the strength of the structural graphite whereby by the action of force it can be easily split up without destroying the particles
Oxidation damage evaluation by non-destructive method for graphite components in high temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shibata, Taiju; Tada, Tatsuya; Sumita, Junya; Sawa, Kazuhiro

2008-01-01

To develop non-destructive evaluation methods for oxidation damage on graphite components in High Temperature Gas-cooled Reactors (HTGRs), the applicability of ultrasonic wave and micro-indentation methods were investigated. Candidate graphites, IG-110 and IG-430, for core components of Very High Temperature Reactor (VHTR) were used in this study. These graphites were oxidized uniformly by air at 500degC. The following results were obtained from this study. (1) Ultrasonic wave velocities with 1 MHz can be expressed empirically by exponential formulas to burn-off, oxidation weight loss. (2) The porous condition of the oxidized graphite could be evaluated with wave propagation analysis with a wave-pore interaction model. It is important to consider the non-uniformity of oxidized porous condition. (3) Micro-indentation method is expected to determine the local oxidation damage. It is necessary to assess the variation of the test data. (author)
Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4
BROOKHAVEN: Looking towards heavy ion physics

International Nuclear Information System (INIS)

Anon.

1988-01-01

July 11-22 were busy days at Brookhaven with a two-week Summer Institute on Relativistic Heavy Ion Physics. After an intensive first week designed to introduce young physicists to high energy heavy ion research, the second week was a workshop on detector technology for Brookhaven's proposed Relativistic Heavy Ion Collider (RHIC), attended by some 150 physicists
Graphite materials testing in the ATR for lifetime management of Magnox reactors

International Nuclear Information System (INIS)

Grover, S.B.; Metcalfe, M.P.

2002-01-01

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on their graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment. (author)
Graphite Materials Testing in the ATR for Lifetime Management of Magnox Reactors

International Nuclear Information System (INIS)

Grover, S.B.; Metcalfe, M.P.

2002-01-01

A major feature of the Magnox gas cooled reactor design is the graphite core, which acts as the moderator but also provides the physical structure for fuel, control rods, instrumentation and coolant gas channels. The lifetime of a graphite core is dependent upon two principal aging processes: irradiation damage and radiolytic oxidation. Irradiation damage from fast neutrons creates lattice defects leading to changes in physical and mechanical properties and the accumulation of stresses. Radiolytic oxidation is caused by the reaction of oxidizing species from the carbon dioxide coolant gas with the graphite, these species being produced by gamma radiation. Radiolytic oxidation reduces the density and hence the moderating capability of the graphite, but also reduces strength affecting the integrity of core components. In order to manage continued operation over the planned lifetimes of their power stations, BNFL needed to extend their database of the effects of these two phenomena on the ir graphite cores through an irradiation experiment. This paper will discuss the background, purpose, and the processes taken and planned (i.e. post irradiation examination) to ensure meaningful data on the graphite core material is obtained from the irradiation experiment
Brookhaven Linac Isotope Producer

Data.gov (United States)

Federal Laboratory Consortium — The Brookhaven Linac Isoptope Producer (BLIP)—positioned at the forefront of research into radioisotopes used in cancer treatment and diagnosis—produces commercially...
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.
Development and testing of nuclear graphite for the German pebble-bed high temperature reactor

International Nuclear Information System (INIS)

Haag, G.; Delle, W.; Nickel, H.; Theymann, W.; Wilhelmi, G.

1987-01-01

Several types of high temperature reactors have been developed in the Federal Republic of Germany. They are all based on spherical fuel elements being surrounded by graphite as reflector material. As an example, HTR-500 developed by the Hochtemperatur Reaktorbau GmbH is shown. The core consists of the top reflector, the side reflector with inner and outer parts, the bottom reflector and the core support columns. The most serious problem with respect to fast neutron radiation damage had to be solved for the materials of those parts near the pebble bed. Regarding the temperature profile in the core, the top reflector is at 300 deg C, and as cooling gas flows from the top downward, the temperature of the inner side reflector rises to about 700 deg C at the bottom. Fortunately, the highest fast neutron load accumulated during the life time of a reactor corresponds to the lowest temperature. This makes graphite components easier to survive neutron exposure without being mechanically damaged, although the maximum fast neutron fluence is as high as 4 x 10 22 /cm 2 at about 400 deg C. HTR graphite components are divided into four classes according to loading. The raw materials for nuclear graphite, the development of pitch coke nuclear graphite, the irradiation behavior of ATR-2E and ASR-IRS and others are reported. (Kako, I.)
Dr. Praveen Chaudhari named director of Brookhaven National Laboratory

CERN Multimedia

2003-01-01

"Brookhaven Science Associates announced today the selection of Dr. Praveen Chaudhari as Director of the U.S. Department of Energy's Brookhaven National Laboratory. Dr. Chaudhari, who will begin his new duties on April 1, joins Brookhaven Lab after 36 years of distinguished service at IBM as a scientist and senior manager of research" (1 page).
Related activities on management of ageing of Dalat Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Lam, Pham Van [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

1998-10-01

The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)
Brookhaven highlights, October 1978-September 1979. [October 1978 to September 1979

Energy Technology Data Exchange (ETDEWEB)

1979-01-01

These highlights present an overview of the major research and development achievements at Brookhaven National Laboratory from October 1978 to September 1979. Specific areas covered include: accelerator and high energy physics programs; high energy physics research; the AGS and improvements to the AGS; neutral beam development; heavy ion fusion; superconducting power cables; ISABELLE storage rings; the BNL Tandem accelerator; heavy ion experiments at the Tandem; the High Flux Beam Reactor; medium energy physics; nuclear theory; atomic and applied physics; solid state physics; neutron scattering studies; x-ray scattering studies; solid state theory; defects and disorder in solids; surface physics; the National Synchrotron Light Source ; Chemistry Department; Biology Department; Medical Department; energy sciences; environmental sciences; energy technology programs; National Center for Analysis of Energy Systems; advanced reactor systems; nuclear safety; National Nuclear Data Center; nuclear materials safeguards; Applied Mathematics Department; and support activities. (GHT)
Radionuclide characterization of graphite stacks from plutonium production reactors of the Siberian group of chemical enterprises

International Nuclear Information System (INIS)

Bushuev, A.V.; Verzilov, Yu.M.; Zubarev, V.N.

2001-01-01

The residual radionuclide concentrations and distributions in graphite from moderator stack of plutonium production reactors at Tomsk-7 have been investigated. It was found that the dominant activity of graphite is 14 C. To gain information on surface and volume contamination of graphite blocks from the moderator stack, the special sets of samples were collected and assayed. The schemes are proposed for evaluation of individual radionuclide inventories together with results of the evaluations performed. (author)
Characterization of graphite-matrix pulsed reactor fuels

International Nuclear Information System (INIS)

Karnes, C.H.; Marion, R.H.

1976-01-01

The performance of the Annular Core Pulsed Reactor (ACPR) is being upgraded in order to accommodate higher fluence experiments for fast reactor fuel element transient and safety studies. The increased fluence requires a two-zone core with the inner zone containing fuel having a high enthalpy and the capability of withstanding very high temperatures during both pulsed and steady state operation. Because the fuel is subjected to a temperature risetime of 2 to 5 ms and to a large temperature difference across the diameter, fracture due to thermal stresses is the primary failure mode. One of the fuels considered for the high enthalpy inner region is a graphite-matrix fuel containing a dispersion of uranium--zirconium carbide solid solution particles. A program was initiated to optimize the development of this class of fuel. This summary presents results on formulations of fuel which have been fabricated by the Materials Technology Group of the Los Alamos Scientific Laboratory
Inhibition of oxidation in nuclear graphite

International Nuclear Information System (INIS)

Winston, Philip L.; Sterbentz, James W.; Windes, William E.

2015-01-01

Graphite is a fundamental material of high-temperature gas-cooled nuclear reactors, providing both structure and neutron moderation. Its high thermal conductivity, chemical inertness, thermal heat capacity, and high thermal structural stability under normal and off-normal conditions contribute to the inherent safety of these reactor designs. One of the primary safety issues for a high-temperature graphite reactor core is the possibility of rapid oxidation of the carbon structure during an off-normal design basis event where an oxidising atmosphere (air ingress) can be introduced to the hot core. Although the current Generation IV high-temperature reactor designs attempt to mitigate any damage caused by a postulated air ingress event, the use of graphite components that inhibit oxidation is a logical step to increase the safety of these reactors. Recent experimental studies of graphite containing between 5.5 and 7 wt% boron carbide (B 4 C) indicate that oxidation is dramatically reduced even at prolonged exposures at temperatures up to 900 deg. C. The proposed addition of B 4 C to graphite components in the nuclear core would necessarily be enriched in B-11 isotope in order to minimise B-10 neutron absorption and graphite swelling. The enriched boron can be added to the graphite during billet fabrication. Experimental oxidation rate results and potential applications for borated graphite in nuclear reactor components will be discussed. (authors)
Graphite-moderated and heavy water-moderated spectral shift controlled reactors; Reactores de moderador solido controlados por desplazamiento espectral

Energy Technology Data Exchange (ETDEWEB)

Alcala Ruiz, F

1984-07-01

It has been studied the physical mechanisms related with the spectral shift control method and their general positive effects on economical and non-proliferant aspects (extension of the fuel cycle length and low proliferation index). This methods has been extended to non-hydrogenous fuel cells of high moderator/fuel ratio: heavy water cells have been con- trolled by graphite rods graphite-moderated and gas-cooled cells have been controlled by berylium rods and graphite-moderated and water-cooled cells have been controlled by a changing mixture of heavy and light water. It has been carried out neutron and thermal analysis on a pre design of these types of fuel cells. We have studied its neutron optimization and their fuel cycles, temperature coefficients and proliferation indices. Finally, we have carried out a comparative analysis of the fuel cycles of conventionally controlled PWRs and graphite-moderated, water-cooled and spectral shift controlled reactors. (Author) 71 refs.
Organic free radicals and micropores in solid graphitic carbonaceous matter at the Oklo natural fission reactors, Gabon

International Nuclear Information System (INIS)

Rigali, M.J.; Nagy, B.

1997-01-01

The presence, concentration, and distribution of organic free radicals as well as their association with specific surface areas and microporosities help characterize the evolution and behavior of the Oklo carbonaceous matter. Such information is necessary in order to evaluate uranium mineralization, liquid bitumen solidification, and radio nuclide containment at Oklo. In the Oklo ore deposits and natural fission reactors carbonaceous matter is often referred to as solid graphitic bitumen. The carbonaceous parts of the natural reactors may contain as much as 65.9% organic C by weight in heterogeneous distribution within the clay-rich matrix. The solid carbonaceous matter immobilized small uraninite crystals and some fission products enclosed in this uraninite and thereby facilitated radio nuclide containment in the reactors. Hence, the Oklo natural fission reactors are currently the subjects of detailed studies because they may be useful analogues to support performance assessment of radio nuclide containment at anthropogenic radioactive waste repository sites. Seven carbonaceous matter rich samples from the 1968 ± 50 Ma old natural fission reactors and the associated Oklo uranium ore deposit were studied by electron spin resonance (ESR) spectroscopy and by measurements of specific surface areas (BET method). Humic acid, fulvic acid, and fully crystalline graphite standards were also examined by ESR spectroscopy for comparison with the Oklo solid graphitic bitumens. With one exception, the ancient Oklo bitumens have higher organic free radical concentrations than the modem humic and fulvic acid samples. The presence of carbon free radicals in the graphite standard could not be determined due to the conductivity of this material. 72 refs., 7 figs., 1 tab
Study of new structures adapted to gas-graphite and gas-heavy water reactors; Etude de structures nouvelles adaptees aux reacteurs graphite-gaz et eau lourde-gaz

Energy Technology Data Exchange (ETDEWEB)

Martin, R; Roche, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The experience acquired as a result of the operation of the Marcoule reactors and of the construction and start-up of the E.D.F. reactors on the one hand, and the conclusions of research and tests carried out out-of-pile on the other hand, lead to a considerable change in the general design of reactors of the gas-graphite type. The main modifications envisaged are analysed in the paper. The adoption of an annular fuel element and of a down-current cooling will make it possible to increase considerably the specific power and the power output of each channel; as a result there will be a considerable reduction in the number of the channels and a corresponding increase in the size of the unit cell. The graphite stack will have to be adapted to there new conditions. For security reasons, the use of prestressed concrete for the construction of the reactor vessel is becoming more widespread; they could lead to the exchangers and the fuel-handling apparatus becoming integrated inside the vessel (the so-called 'attic' device). A full-size mode) of this attic has been built at Saclay with the participation of EURATOM; the operational results obtained are presented as well as a new original design for the control rods. As for as the gas-heavy-water system is concerned, the research is carried out on two points of design; the first, which retains the use of horizontal pressure tubes, takes into account the experience acquired during the construction of the EL 4 reactor of which it will constitute an extrapolation; the second, arising from the research carried out on the gas-graphite system, will use a pre-stressed concrete vessel for holding the pressure, the moderator being almost at the same pressure as the cooling fluid and the fuel being placed in vertical channels. The relative merits of these two variants are analysed in the present paper. (authors) [French] L'experience acquise par l'exploitation des reacteurs de MARCOULE, la construction et le demarrage des reacteurs d
Jordan Research and Training Reactor (JRTR) Utilization Facilities

International Nuclear Information System (INIS)

Xoubi, N.

2013-01-01

Jordan Research and Training Reactor (JRTR) is a 5 MW light water open pool multipurpose reactor that serves as the focal point for Jordan National Nuclear Centre, and is designed to be utilized in three main areas: Education and training, nuclear research, and radioisotopes production and other commercial and industrial services. The reactor core is composed of 18 fuel assemblies, MTR plate type 19.75% enriched uranium silicide (U 3 Si 2 ) in aluminium matrix, and is reflected on all sides by beryllium and graphite. The reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45×10 14 cm -2 s -1 , and is controlled by a Hafnium control absorber rod and B 4 C shutdown rod. The reactor is designed to include laboratories and classrooms that will support the establishment of a nuclear reactor school for educating and training students in disciplines like nuclear engineering, reactor physics, radiochemistry, nuclear technology, radiation protection, and other related scientific fields where classroom instruction and laboratory experiments will be related in a very practical and realistic manner to the actual operation of the reactor. JRTR is designed to support advanced nuclear research as well as commercial and industrial services, which can be preformed utilizing any of its 35 experimental facilities. (author)

Brookhaven National Laboratory

Science.gov (United States)

... For Vendors Departments Public Events Newsroom Technology Licensing Stakeholder Relations Students & Educators Sustainability Privacy and security notice Brookhaven Science Associates Brookhaven Science Associates manages ...
Structural characteristics of a graphite moderated critical assembly for a Zero Power reactor at IEA (Brazil)

International Nuclear Information System (INIS)

Almeida Ferreira, A.C. de; Hukai, R.Y.

1975-01-01

The structural characteristics of a graphite moderated core of a critical assembly to be installed in the Zero Power Reactor of IEA have been defined. These characteristics are the graphite block dimensions, the number and dimensions of the holes in the graphite, the pitch, the dimensions of the sticks of fuel and graphite to be inserted in the holes, and the mechanical reproducibility of the system. The composition of the fuel and moderator sticks were also defined. The main boundary conditions were the range of the relation C/U and C/TH used in commercial HTGR and the neutronics homogeneity
Calculation of radiation heat generation on a graphite reflector side of IAN-R1 Reactor

International Nuclear Information System (INIS)

Duque O, J.; Velez A, L.H.

1987-01-01

Calculation methods for radiation heat generation in nuclear reactor, based on the point kernel approach are revisited and applied to the graphite reflector of IAN-R1 reactor. A Fortran computer program was written for the determination of total heat generation in the reflector, taking 1155 point in it
Design and fabrication of irradiation testing capsule for research reactor materials

International Nuclear Information System (INIS)

Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu

2012-01-01

Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed
Design and fabrication of irradiation testing capsule for research reactor materials

Energy Technology Data Exchange (ETDEWEB)

Yang, Seong Woo; Kim, Bong Goo; Park, Seung Jae; Cho, Man Soon; Choo, Kee Nam; Oh, Jong Myeong; Choi, Myeong Hwan; Lee, Byung Chul; Kang, Suk Hoon; Kim, Dae Jong; Chun, Young Bum; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

Recently, the demand of research reactors is increasing because there are many ageing research reactors in the world. Also, the production of radioisotope related with the medical purpose is very important. Korea Atomic Energy Research Institute (KAERI) is designing and licensing for Jordan Research and Training Reactor (JRTR) and new type research reactor for export which will be constructed in Amman, Jordan and Busan, Korea, respectively. Thus, It is expected that more research reactors will be designed and constructed by KAERI. To design the research reactor, the irradiation performance and behavior of core structure material are necessary. However, the irradiation behavior of these materials is not yet investigated. Therefore, the irradiation performance must be verified by irradiation test. 11M 20K and 11M 21K irradiation capsules were designed and fabricated to conduct the irradiation test for some candidate core materials, Zircaloy 4, beryllium, and graphite, at HANARO. In this paper, the design and fabrication features of 11M 20K and 11M 21K were discussed.
Status of the French research in the field of molten salt nuclear reactors

International Nuclear Information System (INIS)

Hery, M.; Israel, M.; Fauger, P.; Lecocq, A.

1977-01-01

The research program of the CEA in the field of molten salt nuclear reactors has been concerned with MSBR type reactors (Molten Salt Breeder Reactor). The papers written after having performed the theoretical analysis are entitled: core, circuits, chemistry and economy; they include some criticisms and suggestions. The experimental studies consisted in: graphite studies, chemical studies of the salt, metallic materials, the salt loop and the lead loop [fr
Developments in natural uranium - graphite reactors; Developpement des reacteurs a graphite et uranium naturel

Energy Technology Data Exchange (ETDEWEB)

Bourgeois, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Saitcevsky, B. [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

The French natural uranium-graphite power-reactor programme has been developing - from EDF 1 to EDF 4 - in the direction of an increase of the unit power of the installations, of the specific and volume powers, and of an improvement in the operational security conditions. The high power of EDF 4 (500 MWe) and the integration of the primary circuit into the reactor vessel, which is itself made of pre-stressed concrete, make it possible to make the most of the annular fuel elements already in use in EDF 1, and to arrive thus at a very satisfactory solution. The use of an internally cooled fuel element (an annular element) has led to a further step forward: it now becomes possible to increase the pressure of the cooling gas without danger of causing creep in the uranium tube. The use of a pre-stressed concrete vessel makes this pressure increase possible, and the integration of the primary circuit avoids the risk of a rapid depressurization which would be in this case a major danger. This report deals with the main problems presented by this new type of nuclear power station, and gives the main lines of research and studies now being carried out in France. - Neutronic and thermal research has made it possible to consider using large size fuel elements (internal diameter = 77 mm, external diameter 95 mm) while still using natural uranium. - The problems connected with the production of these elements and with their in pile behaviour are the subject of a large programme, both out of pile and in power reactors (EDF 2) and test reactors (Pegase). - The increase in the size of the element leads to a large lattice pitch (35 to 40 cm). This makes it possible to consider having one charging aperture per channel or for a small number of channels, whether the charge machine be inside or outside the pressure vessel. In conclusion are given the main characteristics of a project for a 500 MWe power station using such a fuel element. In particular this project is compared to EDF 4
Development of a Low Temperature Irradiation Capsule for Research Reactor Materials

International Nuclear Information System (INIS)

Choo, Kee Nam; Cho, Man Soon; Lee, Cheol Yong; Yang, Sung Woo; Shin, Yoon Taek; Park, Seng Jae; Kang, Suk Hoon; Kang, Young Hwan; Park, Sang Jun

2013-01-01

A new capsule design was prepared and tested at HANARO for a neutron irradiation of core materials of research reactors as a part of the research reactor development project. Irradiation testing of the materials including graphite, beryllium, and zircaloy-4 that are supposed to be used as core materials in research reactors was required for irradiation at up to 8 reactor operation cycles at low temperature (<100 .deg. C). Therefore, three instrumented capsules were designed and fabricated for an evaluation of the neutron irradiation properties of the core materials (Graphite, Be, Zircaloy-4) of research reactors. The capsules were first designed and fabricated to irradiate materials at low temperature (<100 .deg. C) for a long cycle of 8 irradiation cycles at HANARO. Therefore, the safety of the new designed capsule should be fully checked before irradiation testing. Out-pile performance and endurance testing before HANARO irradiation testing was performed using a capsule under a 110% condition of a reactor coolant flow amount. The structural integrity of the capsule was analyzed in terms of a vibration-induced fatigue cracking of a rod tip of the capsule that is suspected to be the most vulnerable part of a capsule. Another two capsules were irradiated at HANARO for 4 cycles, and one capsule was transferred to a hot cell to examine the integrity of the rod tip of the capsule. After confirming the soundness of the 4 cycle-irradiated capsule, the remaining capsule was irradiated at up to 8 cycles at HANARO. Based on the structural integrity analysis of the capsule, an improved capsule design will be suggested for a longer irradiation test at HANARO
EFFECTS OF REACTOR CONDITIONS ON ELECTROCHEMICAL DECHLORINATION OF TRICHLOROETHYLENE USING GRANULAR-GRAPHITE ELECTRODE.

Science.gov (United States)

Trichloroethylene (TCE) was electrochemically dechlorinated in aqueous environments using granular graphite cathode in a mixed reactor. Effects of pH, current, electrolyte type, and flow rate on TCE dechlorination rate were evaluated. TCE dechlorination rate constant and gas pr...
Brookhaven lecture series No. 227: The Chernobyl accident

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

This lecture discusses the events leading to, during, and following the Chernobyl Reactor number 4 accident. A description of the light water cooled, graphite moderated reactor, the reactor site conditions leading to meltdown is presented. The emission of radioactive effluents and the biological radiation effects is also discussed. (FI)
Fracture criteria of reactor graphite under multiaxial stresses

International Nuclear Information System (INIS)

Sato, S.; Kawamata, K.; Kurumada, A.; Oku, T.

1987-01-01

New fracture criteria for graphite under multiaxial stresses are presented for designing core and support materials of a high temperature gas cooled reactor. Different kinds of fracture strength tests are carried out for a near isotropic graphite IG-11. Results show that, under the stress state in which tensile stresses are predominant, the maximum principal stress theory is seen as applicable for brittle fracture. Under the stress state in which compressive stresses are predominant there may be two fracture modes for brittle fracture, namely, slipping fracture and mode II fracture. For the former fracture mode the maximum shear stress criterion is suitable, but for the latter fracture mode a new mode II fracture criterion including a restraint effect for cracks is verified to be applicable. Also a statistical correction for brittle fracture criteria under multiaxial stresses is discussed. By considering the allowable stress values for safe design, the specified minimum ultimate strengths corresponding to a survival probability of 99% at the 95% confidence level are presented. (orig./HP)
The in-core experimental program at the MIT Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kohse, G.E.; Hu, L-W., E-mail: kohse@mit.edu [Massachusetts Inst. of Technology, Nuclear Reactor Lab., Cambridge, Massachusetts (United States)

2014-07-01

This paper describes the program of in-core experiments at the Massachusetts Institute of Technology Research Reactor (MITR), a 6 MW research reactor. The MITR has a neutron flux and spectrum similar to those in water-cooled power reactors and therefore provides a useful test environment for materials and fuels research. In-core facilities include: a water loop operating at pressurized water or boiling water reactor conditions, an inert gas irradiation facility operating at temperature up to 850 {sup o}C and special purpose facilities including fuel irradiation experiments. Recent and ongoing tests include: water loop investigations of corrosion and thermal and mechanical property evolution of SiC/SiC composites for fuel cladding, irradiation of advanced materials and in-core sensors at elevated temperatures, irradiation in molten fluoride salt at 700 {sup o}C of metal alloy, graphite and composite materials for power reactor applications and instrumented irradiations of metal-bonded hydride fuel. (author)
Oxidation Resistant Graphite Studies

Energy Technology Data Exchange (ETDEWEB)

W. Windes; R. Smith

2014-07-01

The Very High Temperature Reactor (VHTR) Graphite Research and Development Program is investigating doped nuclear graphite grades exhibiting oxidation resistance. During a oxygen ingress accident the oxidation rates of the high temperature graphite core region would be extremely high resulting in significant structural damage to the core. Reducing the oxidation rate of the graphite core material would reduce the structural effects and keep the core integrity intact during any air-ingress accident. Oxidation testing of graphite doped with oxidation resistant material is being conducted to determine the extent of oxidation rate reduction. Nuclear grade graphite doped with varying levels of Boron-Carbide (B4C) was oxidized in air at nominal 740°C at 10/90% (air/He) and 100% air. The oxidation rates of the boronated and unboronated graphite grade were compared. With increasing boron-carbide content (up to 6 vol%) the oxidation rate was observed to have a 20 fold reduction from unboronated graphite. Visual inspection and uniformity of oxidation across the surface of the specimens were conducted. Future work to determine the remaining mechanical strength as well as graphite grades with SiC doped material are discussed.
Non-destructive evaluation on mechanical properties of nuclear graphite with porous structure

International Nuclear Information System (INIS)

Shibata, Taiju; Hanawa, Satoshi; Sumita, Junya; Tada, Tatsuya; Sawa, Kazuhiro; Iyoku, Tatsuo

2005-01-01

As a research subjects of 'Research and development for advanced high temperature gas cooled reactor fuels and graphite components,' we started the study of development of non-destructive evaluation methods for mechanical properties of graphite components. The micro-indentation and ultrasonic wave methods are focused to evaluate the degradation of graphite components in VHTR core. For the micro-indentation method, the test apparatus was designed for the indentation test on graphite specimens with some stress levels. It is expected the stress condition is evaluated by the indentation load-depth characteristics and hardness. For the ultrasonic wave method, ultrasonic wave testing machine and probes were prepared for experiments. It is expected that the stress and inner porous conditions are evaluated by the wave propagation characteristics with wave-pore interaction model. R and D plan to develop the non-destructive evaluation method for graphite is presented in this paper. (This study is the result of contract research in the fiscal year of 2004, Research and development for advanced high temperature gas cooled reactor fuels and graphite components,' which is entrusted to the Japan Atomic Energy Research Institute from the Ministry of Education, Culture, Sports, Science and Technology of Japan.) (author)
The Phase I/II BNCT Trials at the Brookhaven medical research reactor: Critical considerations

International Nuclear Information System (INIS)

Diaz, A.Z.

2001-01-01

A phase I/II clinical trial of boronophenylalanine-fructose (BPA-F) mediated boron neutron capture therapy (BNCT) for Glioblastoma Multiforme (GBM) was initiated at Brookhaven National Laboratory (BNL) in 1994. Many critical issues were considered during the design of the first of many sequential dose escalation protocols. These critical issues included patient selection criteria, boron delivery agent, dose limits to the normal brain, dose escalation schemes for both neutron exposure and boron dose, and fractionation. As the clinical protocols progressed and evaluation of the tolerance of the central nervous system (CNS) to BPA-mediated BNCT at the BMRR continued new specifications were adopted. Clinical data reflecting the progression of the protocols will be presented to illustrate the steps taken and the reasons behind their adoption. (author)
Derivation of a radionuclide inventory for irradiated graphite-chlorine-36 inventory determination

International Nuclear Information System (INIS)

Brown, F.J.; Palmer, J.D.; Wood, P.

2001-01-01

The irradiation of materials in nuclear reactors results in neutron activation of component elements. Irradiated graphite wastes arise from their use in UK gas-cooled research and commercial reactor cores, and in fuel element components, where the graphite has acted as the neutron moderator. During irradiation the residual chlorine, which was used to purify the graphite during manufacture, is activated to chlorine-36. This isotope is long-lived and poorly retarded by geological barriers, and may therefore be a key radionuclide with respect to post-closure disposal facilities performance. United Kingdom Nirex Limited, currently responsible for the development of a disposal route for intermediate-level radioactive wastes in the UK, carried out a major research programme to support an overall assessment of the chlorine-36 activity of all wastes including graphite reactor components. The various UK gas cooled reactors reactors have used a range of graphite components made from diverse graphite types; this has necessitated a systematic programme to cover the wide range of graphite and production processes. The programme consisted of: precursor measurements - on the surface and/or bulk of representative samples of relevant materials, using specially developed methods; transfer studies - to quantify the potential for transfer of Cl-36 into and between waste streams during irradiation of graphite; theoretical assessments - to support the calculational methodology; actual measurements - to confirm the modelling. For graphite, a total of 458 measurements on samples from 57 batches were performed, to provide a detailed understanding of the composition of nuclear graphite. The work has resulted in the generation of probability density functions (PDF) for the mean chlorine concentration of three classes of graphite: fuel element graphite; Magnox moderator and reflector graphite and AGR reflector graphite; AGR moderator graphite. Transfer studies have shown that a significant
Temperature and radiolytic corrosion effects on the chlorine behaviour in nuclear graphite: consequences for the disposable of irradiated graphite from UNGG reactors

International Nuclear Information System (INIS)

Vaudey, C.E.

2010-10-01

This work concerns the dismantling of the UNGG reactor which have produced around 23 000 t of graphite wastes that ave to be disposed of according to the French law of June 206. These wastes contain two long-lived radionuclides ( 14 C and 36 Cl) which are the main long term dose contributors. In order to get information about their inventory and their long term behaviour in case of water ingress into the repository, it is necessary to determine their location and speciation in the irradiated graphite after the reactor shutdown. This work concerns the study of 36 Cl. The main objective is to reproduce its behaviour during reactor operation. For that purpose, we have studied the effects of temperature and radiolytic corrosion independently. Our results show a rapid release of around 20% 36 Cl during the first hours of reactor operation whereas a much slower release occurs afterwards. We have put in evidence two types of chlorine corresponding to two different chemical forms (of different thermal stabilities) or to two locations (of different accessibilities). We have also shown that the radiolytic corrosion seems to enhance chlorine release, whatever the irradiation dose. Moreover, the major chemical form of chlorine is inorganic. (author)
An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

International Nuclear Information System (INIS)

Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

1991-05-01

The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)
Present status of research reactor decommissioning programme in Indonesia

International Nuclear Information System (INIS)

Suripto, A.; Mulyanto, N.

2002-01-01

At present Indonesia has 3 research reactors, namely the 30 MW MTR-type multipurpose reactor at Serpong Site, two TRIGA-type research reactors, the first one being 1 MW located at Bandung Site and the second one a small reactor of 100 kW at Yogyakarta Site. The TRIGA Reactor at the Bandung Site reached its first criticality at 250 kW in 1964, and then was operated at 1000 kW since 1971. In October 2000 the reactor power was successfully upgraded to 2 MW. This reactor has already been operated for 38 years. There is not yet any decision for the decommissioning of this reactor. However it will surely be an object for the near future decommissioning programme and hence anticipation for the above situation becomes necessary. The regulation on decommissioning of research reactor is already issued by the independent regulatory body (BAPETEN) according to which the decommissioning permit has to be applied by the BATAN. For Indonesia, an early decommissioning strategy for research reactor dictates a restricted re-use of the site for other nuclear installation. This is based on high land price, limited availability of radwaste repository site, and other cost analysis. Spent graphite reflector from the Bandung TRIGA reactor is recommended for a direct disposal after conditioning, without any volume reduction treatment. Development of human resources, technological capability as well as information flow from and exchange with advanced countries are important factors for the future development of research reactor decommissioning programme in Indonesia. (author)
Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

Energy Technology Data Exchange (ETDEWEB)

Romano, T.

1997-09-29

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required.

Safety evaluation for packaging (onsite) plutonium recycle test reactor graphite cask

International Nuclear Information System (INIS)

Romano, T.

1997-01-01

This safety evaluation for packaging (SEP) provides the evaluation necessary to demonstrate that the Plutonium Recycle Test Reactor (PRTR) Graphite Cask meets the requirements of WHC-CM-2-14, Hazardous Material Packaging and Shipping, for transfer of Type B, fissile, non-highway route controlled quantities of radioactive material within the 300 Area of the Hanford Site. The scope of this SEP includes risk, shieldling, criticality, and.tiedown analyses to demonstrate that onsite transportation safety requirements are satisfied. This SEP also establishes operational and maintenance guidelines to ensure that transport of the PRTR Graphite Cask is performed safely in accordance with WHC-CM-2-14. This SEP is valid until October 1, 1999. After this date, an update or upgrade to this document is required
The use and evolution of the CEA research reactors

International Nuclear Information System (INIS)

Rossillon, F.; Chauvez, C.

1964-01-01

The authors successively examine the different research reactors in use in the French C.E.A. Nuclear Centres. They trace briefly their histories, describing how they have been used up to the present, and how they have been adapted to changes in programme by means of certain modifications. They also describe the reasons which have led to the elaboration of the project for the new reactor Osiris. Zoe, the oldest reactor in the CEA, has been in service in the Centre de Fontenay-aux-Roses since 1948. It is used mainly for measurements of absorption cross-sections in graphite, and for various short irradiations which do not require high fluxes. The reactor EL 2, in service since 1952, was used for the first studies on gas cooling. It has also been widely used for the production of radioisotopes and for a large number of experiments in the fields of physics, metallurgy and physical chemistry. The ageing of certain elements of the reactor has led to the decision to close it down in the near future The reactor EL 3 has been widely used for experiments in physics and in the investigation of fuels. The possibilities of the reactor in fast neutron irradiations will be considerably improved by the adoption of a new type of core (the 'snow crystal' structure). Triton-I, a 2 MW swimming-pool reactor, is used for the most part for fast neutron and gamma irradiations. The modifications being carried out on it at present should result in an increase in the power of the reactor up to 4 or 5 MW. In a neighbouring compartment is housed Triton-II which is of the same general structure, as Triton-I, but whose maximum power is 100 kW. Triton-II is used solely for studies on shielding. Melusine, a 2 MW swimming-pool reactor, has been in use in the Centre d'Etudes Nucleaires de Grenoble since 1959. It has supported a very high programme concerned mainly with solid state physics, fundamental research into refractory fissile materials and special graphites, and the study of the behaviour of
Method to Assess the Radionuclide Inventory of Irradiated Graphite from Gas-Cooled Reactors - 13072

Energy Technology Data Exchange (ETDEWEB)

Poncet, Bernard [EDF-CIDEN, 154 Avenue Thiers, CS 60018, F-69458 LYON cedex 06 (France)

2013-07-01

About 17,000 t of irradiated graphite waste will be produced from the decommissioning of the six French gas-cooled nuclear reactors. Determining the radionuclide (RN) content of this waste is of relevant importance for safety reasons and in order to determine the best way to manage them. For many reasons the impurity content that gave rise to the RNs in irradiated graphite by neutron activation during operation is not always well known and sometimes actually unknown. So, assessing the RN content by the use of traditional calculation activation, starting from assumed impurity content, leads to a false assessment. Moreover, radiochemical measurements exhibit very wide discrepancies especially on RN corresponding to precursor at the trace level such as natural chlorine corresponding to chlorine 36. This wide discrepancy is unavoidable and is due to very simple reasons. The level of impurity is very low because the uranium fuel used at that very moment was not enriched, so it was a necessity to have very pure nuclear grade graphite and the very low size of radiochemical sample is a simple technical constraint because device size used to get mineralization product for measurement purpose is limited. The assessment of a radionuclide inventory only based on few number of radiochemical measurements lead in most cases, to a gross over or under-estimation that is detrimental for graphite waste management. A method using an identification calculation-measurement process is proposed in order to assess a radiological inventory for disposal sizing purpose as precise as possible while guaranteeing its upper character. This method present a closer approach to the reality of the main phenomenon at the origin of RNs in a reactor, while also incorporating the secondary effects that can alter this result such as RN (or its precursor) release during reactor operation. (authors)
Brookhaven highlights, 1986-1987

International Nuclear Information System (INIS)

Rowe, M.S.

1988-01-01

The highlights of research conducted between October 1985 and September 1987 at Brookhaven National Laboratory are reviewed in this publication. Also covered are the administrative and financial status of the laboratory and a brief mention of meetings held and honors received. (FI)
Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France

International Nuclear Information System (INIS)

Gaussens, J.; Tanguy, P.

1964-01-01

The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable
Role of nuclear grade graphite in controlling oxidation in modular HTGRs

Energy Technology Data Exchange (ETDEWEB)

Windes, Willaim [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kane, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-11-01

The passively safe High Temperature Gas-cooled Reactor (HTGR) design is one of the primary concepts considered for Generation IV and Small Modular Reactor (SMR) programs. The helium cooled, nuclear grade graphite moderated core achieves extremely high operating temperatures allowing either industrial process heat or electricity generation at high efficiencies. In addition to their neutron moderating properties, nuclear grade graphite core components provide excellent high temperature stability, thermal conductivity, and chemical compatibility with the high temperature nuclear fuel form. Graphite has been continuously used in nuclear reactors since the 1940’s and has performed remarkably well over a wide range of core environments and operating conditions. Graphite moderated, gas-cooled reactor designs have been safely used for research and power production purposes in multiple countries since the inception of nuclear energy development. However, graphite is a carbonaceous material, and this has generated a persistent concern that the graphite components could actually burn during either normal or accident conditions [ , ]. The common assumption is that graphite, since it is ostensibly similar to charcoal and coal, will burn in a similar manner. While charcoal and coal may have the appearance of graphite, the internal microstructure and impurities within these carbonaceous materials are very different. Volatile species and trapped moisture provide a source of oxygen within coal and charcoal allowing them to burn. The fabrication process used to produce nuclear grade graphite eliminates these oxidation enhancing impurities, creating a dense, highly ordered form of carbon possessing high thermal diffusivity and strongly (covalently) bonded atoms.
Characterization of Ignalina NPP RBMK Reactors Graphite

International Nuclear Information System (INIS)

Hacker, P.J.; Neighbour, G.B.; Levinskas, R.; Milcius, D.

2001-01-01

The paper concentrates on the investigations of the initial physical properties of graphite used in production of graphite bricks of Ignalina NPP. These graphite bricks are used as nuclear moderator and major core structural components. Graphite bulk density is calculated by mensuration, pore volumes are measured by investigation of helium gas penetration in graphite pore network, the Young's modulus is determined using an ultrasonic time of flight method, the coefficient of thermal expansion is determined using a Netzsch dilatometer 402C, the fractured and machined graphite surfaces are studied using SEM, impurities are investigated qualitatively by EDAX, the degree of graphitization of the material is tested using X-ray diffraction. (author)
Solid State Reactor Final Report

Energy Technology Data Exchange (ETDEWEB)

Mays, G.T.

2004-03-10

The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas
Diagnosis of electric equipment at the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Nguyen Truong Sinh

1999-01-01

The Dalat Nuclear Research Reactor (DNRR) is a pool type of its kind in the world: Soviet-designed core and control system harmoniously integrated into the left-over infrastructure of the former American-made TRIGA MARK II reactor, which includes the reactor tank and shielding, graphite reflector, beam tubes and thermal column. The reactor is mainly used for radioisotope and radiopharmaceutical production, elemental analysis using neutron activation techniques, neutron beam exploitation, silicon doping, and reactor physics experimentation. For safe operation of the reactor maintenance work has been carried out for the reactor control and instrumentation, reactor cooling, ventilation, radiomonitoring, mechanical, normal electric supply systems as well as emergency electric diesel generators and the water treatment station. Technical management of the reactor includes periodical maintenance as required by technical specifications, training, re-training and control of knowledge for reactor staff. During recent years, periodic preventive maintenance (PPM) has been carried out for the electric machines of the technological systems. (author)
Uranium Oxide Aerosol Transport in Porous Graphite

Energy Technology Data Exchange (ETDEWEB)

Blanchard, Jeremy; Gerlach, David C.; Scheele, Randall D.; Stewart, Mark L.; Reid, Bruce D.; Gauglitz, Phillip A.; Bagaasen, Larry M.; Brown, Charles C.; Iovin, Cristian; Delegard, Calvin H.; Zelenyuk, Alla; Buck, Edgar C.; Riley, Brian J.; Burns, Carolyn A.

2012-01-23

The objective of this paper is to investigate the transport of uranium oxide particles that may be present in carbon dioxide (CO2) gas coolant, into the graphite blocks of gas-cooled, graphite moderated reactors. The transport of uranium oxide in the coolant system, and subsequent deposition of this material in the graphite, of such reactors is of interest because it has the potential to influence the application of the Graphite Isotope Ratio Method (GIRM). The GIRM is a technology that has been developed to validate the declared operation of graphite moderated reactors. GIRM exploits isotopic ratio changes that occur in the impurity elements present in the graphite to infer cumulative exposure and hence the reactor’s lifetime cumulative plutonium production. Reference Gesh, et. al., for a more complete discussion on the GIRM technology.
An automatic regulating control system for a graphite moderated reactor using digital techniques

International Nuclear Information System (INIS)

Carvalho Goncalves Junior, J. de.

1989-01-01

The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)
Brookhaven Highlights, January 1982-March 1983

Energy Technology Data Exchange (ETDEWEB)

Kuper, J.B.H.; Rustad, M.C. (eds.)

1983-01-01

Research at Brookhaven National Laboratory is summarized. Major headings are high energy physics, physics and chemistry, life sciences, applied energy science, support activities and administration. (GHT)
Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-08-01

The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs.
Graphite moderator lifecycle behaviour. Proceedings of a specialists meeting

International Nuclear Information System (INIS)

1996-08-01

The meeting provided the forum for graphite specialists representing operating and research organizations worldwide to exchange information in the following areas: the status of graphite development; operation and safety procedures for existing and future graphite moderated reactors; graphite testing techniques; review of the experiences gained and data acquired on the influence of neutron irradiation and oxidizing conditions on key graphite properties; and to exchange information useful for decommissioning activities. The participants provided twenty-seven papers on behalf of their countries and respective technical organizations. An open discussion followed each of the presentations. A consistently reoccurring theme throughout the specialists meeting was the noticeable reduction in the number of graphite experts remaining the nuclear power industry. Graphite moderated power reactors have provided a significant contribution to the generation of electricity throughout the past forty years and will continue to be a prominent energy source for the future. Yet, many of the renowned experts in the field of nuclear graphites are nearing the end of their careers without apparent replacement. This, coupled with changes in the focus on nuclear power by some industrialized countries, has prompted the IAEA to initiate an evaluation on the feasibility and interest by Member States of establishing a central archive facility for the storage of data on irradiated graphites. Refs, figs, tabs
Pyrolysis and its potential use in nuclear graphite disposal

International Nuclear Information System (INIS)

Mason, J.B.; Bradbury, D.

2001-01-01

Graphite is used as a moderator material in a number of nuclear reactor designs, such as MAGNOX and AGR gas cooled reactors in the United Kingdom and the RBMK design in Russia. During construction the moderator of the reactor is usually installed as an interlocking structure of graphite bricks. At the end of reactor life the graphite moderator, weighing typically 2,000 tonnes, is a radioactive waste which requires eventual management. Radioactive graphite disposal options conventionally include: In-situ SAFESTORE for extended periods to permit manual disassembly of the graphite moderator through decay of short-lived radionuclides. Robotic or manual disassembly of the reactor core followed by disposal of the graphite blocks. Robotic or manual disassembly of the reactor core followed by incineration of the graphite and release of the resulting carbon dioxide Studsvik, Inc. is a nuclear waste management and waste processing company organised to serve the US nuclear utility and government facilities. Studsvik's management and technical staff have a wealth of experience in processing liquid, slurry and solid low level radioactive waste using (amongst others) pyrolysis and steam reforming techniques. Bradtec is a UK company specialising in decontamination and waste management. This paper describes the use of pyrolysis and steam reforming techniques to gasify graphite leading to a low volume off-gas product. This allows the following options/advantages. Safe release of any stored Wigner energy in the graphite. The process can accept small pieces or a water-slurry of graphite, which enables the graphite to be removed from the reactor core by mechanical machining or water cutting techniques, applied remotely in the reactor fuel channels. In certain situations the process could be used to gasify the reactor moderator in-situ. The low volume of the off-gas product enables non-carbon radioactive impurities to be efficiently separated from the off-gas. The off-gas product can
Description of the intense, low energy, monoenergetic positron beam at Brookhaven

International Nuclear Information System (INIS)

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

1985-01-01

An intense (4 x 10 7 s -1 ), low energy (approx. =1.0 eV), monoenergetic (ΔE approx. = 75 MeV) beam of positrons has been built at the Brookhaven National Laboratory. This flux is more than 10 times greater than any existing beam from radioactive sources. Plans are underway to increase further the flux by more than an order of magnitude. The intense low energy positron beam is made by utilizing the High Flux Beam Reactor at Brookhaven to produce the isotope 64 Cu with an activity of 40 curies of positrons. Source moderation techniques are utilized to produce the low energy positron beam from the high energy positrons emitted from 64 Cu. 31 refs., 7 figs
Brookhaven fastbus/unibus interface

International Nuclear Information System (INIS)

Benenson, G.; Bauernfeind, J.; Larsen, R.C.

1983-01-01

A typical high energy physics experiment requires both a high speed data acquisition and processing system, for data collection and reduction; and a general purpose computer to handle further reduction, bookkeeping and mass storage. Broad differences in architecture, format or technology, will often exist between these two systems, and interface design can become a formidable task. The PDP-11 series minicomputer is widely used in physics research, and the Brookhaven FASTBUS is the only standard high speed data acquisition system which is fully implemented in a current high energy physics experiment. This paper will describe the design and operation of an interface between these two systems. The major issues are elucidated by a preliminary discussion on the basic principles of Bus Systems, and their application to Brookhaven FASTBUS and UNIBUS
QUARTERLY PROGRESS REPORT JANUARY, FEBRUARY, MARCH, 1968 REACTOR FUELS AND MATERIALS DEVELOPMENT PROGRAMS FOR FUELS AND MATERIALS BRANCH OF USAEC DIVISION OF REACTOR DEVELOPMENT AND TECHNOLOGY

Energy Technology Data Exchange (ETDEWEB)

Cadwell, J. J.; de Halas, D. R.; Nightingale, R. E.; Worlton, D. C.

1968-06-01

Progress is reported in these areas: nuclear graphite; fuel development for gas-cooled reactors; HTGR graphite studies; nuclear ceramics; fast-reactor nitrides research; non-destructive testing; metallic fuels; basic swelling studies; ATR gas and water loop operation and maintenance; reactor fuels and materials; fast reactor dosimetry and damage analysis; and irradiation damage to reactor metals.
BROOKHAVEN: Booster commissioned

Energy Technology Data Exchange (ETDEWEB)

Bleser, Ed

1992-03-15

The construction and first commissioning phase of the Booster synchrotron to inject into Brookhaven's veteran Alternating Gradient Synchrotron (AGS) were completed last year. Scheduled to come into operation this year, the new Booster will extend the research capabilities AGS, and with its ability to accelerate partially stripped heavy ions will play an essential role in the chain of accelerators serving the Relativistic Heavy Ion Collider (RHIC)
A reverse method for the determination of the radiological inventory of irradiated graphite at reactor scale

Energy Technology Data Exchange (ETDEWEB)

Nicaise, Gregory [Institut de Radioprotection et de Surete Nucleaire, Fontenay-aux-roses (France); Poncet, Bernard [EDF-DP2D, Lyon (France)

2016-11-15

Irradiated graphite waste will be produced from the decommissioning of the six gas-cooled nuclear reactors operated by Electricite De France (EDF). Determining the radionuclide content of this waste is an important legal commitment for both safety reasons and in order to determine the best suited management strategy. As evidenced by numerous studies nuclear graphite is a very pure material, however, it cannot be considered from an analytical viewpoint as a usual homogeneous material. Because of graphite high purity, radionuclide measurements in irradiated graphite exhibit very high discrepancies especially when corresponding to precursors at trace level. Therefore the assessment of a radionuclide inventory only based on few number of radiochemical measurements leads in most of cases to a gross over or under-estimation that can be detrimental to graphite waste management. A reverse method using an identification calculation-measurement process is proposed in order to assess the radionuclide inventory as precisely as possible.

Nuclear graphite development, operational problems, and resolution of these problems at the Hanford production reactors

International Nuclear Information System (INIS)

Morgan, W.C.

1996-01-01

This paper chronicles the history of the Hanford Production Reactor, from the initial design considerations for B, D, and F Reactors through the selection of the agreed method for safe disposal of the decommissioned reactors. The operational problems that challenged the operations and support staff of each new generation of production reactors, the engineering actions an operational changes that alleviated or resolved the immediate problems, the changes in reactor design and design-bases for the next generation of production reactors, and the changes in manufacturing variables that resulted in new ''improved'' grades of nuclear graphites for use in the moderators of the Hanford Production Reactors are reviewed in the context of the existing knowledge-base and the mission-driven priorities on the time. 14 refs, 6 figs, 3 tabs
Research on neutron radiography in Research Reactor Institute, Kyoto University and activities related to it

International Nuclear Information System (INIS)

Fujine, Shigenori; Yoneda, Kenji

1994-01-01

The research on neutron radiography in Research Reactor Institute, Kyoto University was begun in 1974 using the E-2 experimental hole which was designed for neutron irradiation. It was reconstructed for the excellent performance as neutron radiography facility by fixing aluminum plugs, a collimator and so on. The research activities thereafter are briefly described. In 1989, the cold neutron facility was installed in the graphite thermal neutron facility, and the experiment on cold neutron radiography became feasible. The reactor in Kyoto University is of the thermal output of 5 MW, and is put to the joint utilization by universities and research institutes in whole Japan. The experimental items carried out so far are enumerated. At present, the main subjects of research are the development of the standard for establishing image evaluation method, the analysis of gas-liquid two-phase flow, the construction of the data base for the literatures and images of neutron radiography, the application of cold neutron radiography, the development of the imaging method using fast neutrons and so on. The thermal neutron radiography and the cold neutron radiography facilities of Kyoto University research reactor are described. The research and activities at Kyoto University research reactor and the investigation of problems are reported. (K.I.) 56 refs
Research program of the high temperature engineering test reactor for upgrading the HTGR technology

International Nuclear Information System (INIS)

Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

1997-07-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)
Graphite for fusion energy applications

International Nuclear Information System (INIS)

Eatherly, W.P.; Clausing, R.E.; Strehlow, R.A.; Kennedy, C.R.; Mioduszewski, P.K.

1987-03-01

Graphite is in widespread and beneficial use in present fusion energy devices. This report reflects the view of graphite materials scientists on using graphite in fusion devices. Graphite properties are discussed with emphasis on application to fusion reactors. This report is intended to be introductory and descriptive and is not intended to serve as a definitive information source
Graphite selection for the PBMR reflector

International Nuclear Information System (INIS)

Marsden, B.J.; Preston, S.D.

2000-01-01

A high temperature, direct cycle gas turbine, graphite moderated, helium cooled, pebble-bed reactor (PBMR) is being designed and constructed in South Africa. One of the major components in the PBMR is the graphite reflector, which must be designed to last thirty-five full power years. Fast neutron irradiation changes the dimensions and material properties of reactor graphite, thus for design purposes a suitable graphite database is required. Data on the effect of irradiation on nuclear graphites has been gathered for many years, at considerable financial cost, but unfortunately these graphites are no longer available due to rationalization of the graphite industry and loss of key graphite coke supplies. However, it is possible, using un-irradiated graphite materials properties and knowledge of the particular graphite microstructure, to determine the probable irradiation behaviour. Three types of nuclear graphites are currently being considered for the PBMR reflector: an isostatically moulded, fine grained, high strength graphite and two extruded medium grained graphites of moderately high strength. Although there is some irradiation data available for these graphites, the data does not cover the temperature and dose range required for the PBMR. The available graphites have been examined to determine their microstructure and some of the key material properties are presented. (authors)
Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

International Nuclear Information System (INIS)

2005-03-01

The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)
Graphite reactor physics; Physique des piles a graphite

Energy Technology Data Exchange (ETDEWEB)

Bacher, P; Cogne, F [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Noc, B [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

The study of graphite-natural uranium power reactor physics, undertaken ten years ago when the Marcoule piles were built, has continued to keep in step with the development of this type of pile. From 1960 onwards the critical facility Marius has been available for a systematic study of the properties of lattices as a function of their pitch, of fuel geometry and of the diameter of cooling channels. This study has covered a very wide field: lattice pitch varying from 19 to 38 cm. uranium rods and tubes of cross-sections from 6 to 35 cm{sup 2}, channels with diameters between 70 and 140 mm. The lattice calculation methods could thus be checked and where necessary adapted. The running of the Marcoule piles and the experiments carried out on them during the last few years have supplied valuable information on the overall evolution of the neutronic properties of the fuel as a function of irradiation. More detailed experiments have also been performed in Marius with plutonium-containing fuels (irradiated or synthetic fuels), and will be undertaken at the beginning of 1965 at high temperature in the critical facility Cesar, which is just being completed at Cadarache. Spent fuel analyses complement these results and help in their interpretation. The thermalization and spectra theories developed in France can thus be verified over the whole valid temperature range. The efficiency of control rods as a function of their dimensions, the materials of which they are made and the lattices surrounding them has been measured in Marius, and the results compared with calculation on the one hand and with the measurements carried out in EDF 1 on the other. Studies on the control proper of graphite piles were concerned essentially with the risks of spatial instability and the means of detecting and controlling them, and with flux distortions caused by the control rods. (authors) [French] Entreprise il y a dix ans a l'occasion de la construction des piles de Marcoule, l'etude de la
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.
Radiation damage and life-time evaluation of RBMK graphite stack

Energy Technology Data Exchange (ETDEWEB)

Platonov, P A; Chugunov, O K; Manevsky, V N; Karpukhin, V I [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation). Reactor Material Div.

1996-08-01

At the present time there are 11 NPP units with RBMK reactors in operation in Russia, with the oldest now in operation 22 years. Design life-time of the RBMK-1000 reactor is 30 years. This paper addresses the evaluation of RBMK graphite stack life-time. It is the practice in Russia to evaluate the reliability of the channel reactor graphite stack using at least three criteria: degradation of physical-mechanical properties of graphite, preservation of the graphite brick integrity, and degradation of the graphite stack as a structure. Stack life-time evaluation by different criteria indicates that the most realistic approach may be realized on the basis of the criteria of brick cracking and degradation of the graphite stack as a structure. The RBMK reactor graphite stack life-time depends on its temperature and for different units it may be different. (author). 2 refs, 10 figs.
A solution to level 3 dismantling of gas-cooled reactors: Graphite incineration

International Nuclear Information System (INIS)

Dubourg, M.

1993-01-01

This paper presents an approach developed to solve the specific decommissioning problems of the G2 and G3 gas cooled reactors at Marcoule and the strategy applied with emphasis in incinerating the graphite core components, using a fluidized-bed incinerator developed jointly between the CEA and FRAMATOME. The incineration option was selected over subsurface storage for technical and economic reasons. Studies have shown that gaseous incineration releases are environmentally acceptable
Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

Energy Technology Data Exchange (ETDEWEB)

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael; Papin, Pallas; Nelson, Andrew; Hunter, James

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.
Characteristics of first loaded IG-110 graphite in HTTR core

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Iyoku, Tatsuo; Sawa, Kazuhiro; Hanawa, Satoshi; Ishihara, Masahiro

2006-10-01

IG-110 graphite is a fine-grained isotropic and nuclear-grade graphite with excellent resistivity on both irradiation and corrosion and with high strength. The IG-110 graphite is used for the graphite components of High Temperature Engineering Test Reactor (HTTR) such as fuel and control rod guide blocks and support posts. In order to design and fabricate the graphite components in the HTTR, the Japan Atomic Energy Research Institute (the Japan Atomic Energy Agency at present) had established the graphite structural design code and design data on the basis of former research results. After the design code establishment, the IG-110 graphite components were fabricated and loaded in the HTTR core. This report summarized the characteristics of the first loaded IG-110 graphite as basic data for surveillance test, measuring material characteristics changed by neutron irradiation and oxidation. By comparing the design data, it was shown that the first loaded IG-110 graphite had excellent strength properties and enough safety margins to the stress limits in the design code. (author)
Recuperation of the energy released in the G-1, an air-cooled graphite reactor core

International Nuclear Information System (INIS)

Chambadal, P.; Pascal, M.

1955-01-01

The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [fr
Researchers solve big mysteries of pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Shams, Afaque; Roelofs, Ferry; Komen, E.M.J. [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Baglietto, Emilio [Massachusetts Institute of Technology, Cambridge, MA (United States). Dept. of Nuclear Science and Engineering; Sgro, Titus [CD-adapco, London (United Kingdom). Technical Marketing

2014-03-15

The PBR is one type of High Temperature Reactors, which allows high temperature work while preventing the fuel from melting (bringing huge safety margins to the reactor) and high electricity efficiency. The design is also highly scalable; a plant could be designed to be as large or small as needed, and can even be made mobile, allowing it to be used onboard a ship. In a PBR, small particles of nuclear fuel, embedded in a moderating graphite pebble, are dropped into the reactor as needed. At the bottom, the pebbles can be removed simply by opening a small hatch and letting gravity pull them down. To cool the reactor and create electricity, helium gas is pumped through the reactor to pull heat out which is then run through generators. One of the most difficult problems to deal with has been the possible appearance of local temperature hotspots within the pebble bed heating to the point of melting the graphite moderators surrounding the fuel. Obviously, constructing a reactor and experimenting to investigate this possibility is out of the question. Instead, nuclear engineers have been attempting to simulate a PBR with various CFD codes. The thermo-dynamic analysis to simulate realistic conditions in a pebble bed are described and the results are shown. (orig.)
AGC-2 Graphite Preirradiation Data Package

Energy Technology Data Exchange (ETDEWEB)

David Swank; Joseph Lord; David Rohrbaugh; William Windes

2012-10-01

The NGNP Graphite R&D program is currently establishing the safe operating envelope of graphite core components for a Very High Temperature Reactor (VHTR) design. The program is generating quantitative data necessary for predicting the behavior and operating performance of the new nuclear graphite grades. To determine the in-service behavior of the graphite for pebble bed and prismatic designs, the Advanced Graphite Creep (AGC) experiment is underway. This experiment is examining the properties and behavior of nuclear grade graphite over a large spectrum of temperatures, neutron fluences and compressive loads. Each experiment consists of over 400 graphite specimens that are characterized prior to irradiation and following irradiation. Six experiments are planned with the first, AGC-1, currently being irradiated in the Advanced Test Reactor (ATR) and pre-irradiation characterization of the second, AGC-2, completed. This data package establishes the readiness of 512 specimens for assembly into the AGC-2 capsule.
High temperature graphite irradiation creep experiment in the Dragon Reactor. Dragon Project report

Energy Technology Data Exchange (ETDEWEB)

Manzel, R.; Everett, M. R.; Graham, L. W.

1971-05-15

The irradiation induced creep of pressed Gilsocarbon graphite under constant tensile stress has been investigated in an experiment carried out in FE 317 of the OECD High Temperature Gass Cooled Reactor ''Dragon'' at Winfrith (England). The experiment covered a temperature range of 850 dec C to 1240 deg C and reached a maximum fast neutron dose of 1.19 x 10²¹ n cm^-2 NDE (Nickel Dose DIDO Equivalent). Irradiation induced dimensional changes of a string of unrestrained graphite specimens are compared with the dimensional changes of three strings of restrained graphite specimens stressed to 40%, 58%, and 70% of the initial ultimate tensile strength of pressed Gilsocarbon graphite. Total creep strains ranging from 0.18% to 1.25% have been measured and a linear dependence of creep strain on applied stress was observed. Mechanical property measurements carried out before and after irradiation demonstrate that Gilsocarbon graphite can accommodate significant creep strains without failure or structural deterioration. Total creep strains are in excellent agreement with other data, however the results indicate a relatively large temperature dependent primary creep component which at 1200 deg C approaches a value which is three times larger than the normally assumed initial elastic strain. Secondary creep constants derived from the experiment show a temperature dependence and are in fair agreement with data reported elsewhere. A possible determination of the results is given.
Preparation of pyrolytic carbon coating on graphite for inhibiting liquid fluoride salt and Xe135 penetration for molten salt breeder reactor

International Nuclear Information System (INIS)

Song, Jinliang; Zhao, Yanling; He, Xiujie; Zhang, Baoliang; Xu, Li; He, Zhoutong; Zhang, DongSheng; Gao, Lina; Xia, Huihao; Zhou, Xingtai; Huai, Ping; Bai, Shuo

2015-01-01

Highlights: • Rough laminar pyrolytic carbon coating (RLPyC) is prepared by a fixed-bed method. • The salt-infiltration into IG-110 is 13.5%, less than 0.01% of RLPyC under 1.5 atm. • The helium diffusion coefficient of RLPyC coated graphite is 2.16 × 10 −8 cm 2 /s. • The coated graphite can inhibit the liquid fluoride salt and Xe 135 penetration. - Abstract: A fixed-bed deposition method was used to prepare rough laminar pyrolytic carbon coating (RLPyC) on graphite for inhibiting liquid fluoride salt and Xe 135 penetration during use in molten salt breeder reactor. The RLPyC coating possessed a graphitization degree of 44% and had good contact with graphite substrate. A high-pressure reactor was constructed to evaluate the molten salt infiltration in the isostatic graphite (IG-110, TOYO TANSO CO., LTD.) and RLPyC coated graphite under 1.01, 1.52, 3.04, 5.07 and 10.13 × 10 5 Pa for 12 h. Mercury injection and molten-salt infiltration experiments indicated the porosity and the salt-infiltration amount of 18.4% and 13.5 wt% under 1.52 × 10 5 Pa of IG-110, which was much less than 1.2% and 0.06 wt% under 10.13 × 10 5 Pa of the RLPyC, respectively. A vacuum device was constructed to evaluate the Xe 135 penetration in the graphite. The helium diffusion coefficient of RLPyC coated graphite was 2.16 × 10 −12 m 2 /s, much less than 1.21 × 10 −6 m 2 /s of the graphite. Thermal cycle experiment indicated the coatings possessed excellent thermal stability. The coated graphite could effectively inhibit the liquid fluoride salt and Xe 135 penetration
Progress in radioactive graphite waste management

International Nuclear Information System (INIS)

2010-07-01

Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management
Nuclear graphite waste management. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

2001-05-01

The purpose of the seminar was to bring together the specialists dealing with various aspects of radioactive graphite waste management to exchange and review information on the decommissioning, characterisation, processing and disposal of irradiated graphite from reactor cores and other graphite waste associated with reactor operation. The seminar covered radioactive graphite characterisation, the effect of irradiation on graphite components, Wigner energy, radioactive graphite waste treatment, conditioning, interim storage and long term disposal options. Individual papers presented at the seminar were indexed separately
Brookhaven at 40 - looking forward as well as back

International Nuclear Information System (INIS)

Anon.

1987-01-01

In 1947, the famous Camp Upton Army Base on New York's Long Island switched to a new career as Brookhaven National Laboratory. The reputation the Laboratory has established as a world-class research centre and its continued attraction for scientists looking for exciting possibilities were highlighted on 9-11 September at a symposium and celebration marking forty years of Brookhaven and its parent organization, AUI (Associated Universities Inc)

IEA-R1 research reactor: operational life extension and considerations regarding future decommissioning

International Nuclear Information System (INIS)

Frajndlich, Roberto

2009-01-01

The IEA-R1 reactor is a pool type research reactor moderated and cooled by light water and uses graphite and beryllium reflectors. The reactor is located at the Instituto de Pesquisas Energeticas e Nucleares (IPEN-CNEN/SP), in the city of Sao Paulo, Brazil. It is the oldest research reactor in the southern hemisphere and one of the oldest of this kind in the world. The first criticality of the reactor was obtained on September 16, 1957. Given the fact that Brazil does not have yet a definitive radioactive waste repository and a national policy establishing rules for the spent fuel storage, the institutions which operate the research reactors for more than 50 years in the country have searched internal solutions for continued operation. This paper describes the spent fuel assemblies and radioactive waste management process for the IEA-R1 reactor and the refurbishment and modernization program adopted to extend its lifetime. Some considerations about the future decommissioning of the reactor are also discussed which, in my opinion, might help the operating organization to make decisions about financial, legal and technical aspects of the decommissioning procedures in a time frame of 10-15 years(author)
Characterisation of Chlorine Behavior in French Graphite

International Nuclear Information System (INIS)

Blondel, A.; Moncoffre, N.; Toulhoat, N.; Bererd, N.; Petit, L.; Laurent, G.; Lamouroux, C.

2016-01-01

Chlorine 36 is one of the main radionuclides of concern for French graphite waste disposal. In order to help the understanding of its leaching behaviour under disposal conditions, the respective impact of temperature, irradiation and gas radiolysis on chlorine release in reactor has been studied. Chlorine 36 has been simulated through chlorine 37 ion implantation in virgin nuclear graphite samples. Results show that part of chlorine is highly mobile in graphite in the range of French reactors operating temperatures in relation with graphite structural recovering. Ballistic damage generated by irradiation also promotes chlorine release whereas no clear impact of the coolant gas radiolysis was observed in the absence of graphite radiolytic corrosion. (author)
Reactor safety research and safety technology. Pt. 2

International Nuclear Information System (INIS)

Theenhaus, R.; Wolters, J.

1987-01-01

The state of HTR safety research work reached permits a comprehensive and reliable answer to be given to questions which have been raised by the reactor accident at Chernobyl, regarding HTR safety. Together with the probability safety analyses, the way to a safety concept suitable for an HTR is cleared; instructions are given for design optimisation with regard to safety technique and economy. The consequences of a graphite fire, the neutron physics design and the consequenes of the lack of a safety containment are briefly described. (DG) [de
Ion irradiation to simulate neutron irradiation in model graphites: Consequences for nuclear graphite

Science.gov (United States)

Galy, N.; Toulhoat, N.; Moncoffre, N.; Pipon, Y.; Bérerd, N.; Ammar, M. R.; Simon, P.; Deldicque, D.; Sainsot, P.

2017-10-01

Due to its excellent moderator and reflector qualities, graphite was used in CO2-cooled nuclear reactors such as UNGG (Uranium Naturel-Graphite-Gaz). Neutron irradiation of graphite resulted in the production of 14C which is a key issue radionuclide for the management of the irradiated graphite waste. In order to elucidate the impact of neutron irradiation on 14C behavior, we carried out a systematic investigation of irradiation and its synergistic effects with temperature in Highly Oriented Pyrolitic Graphite (HOPG) model graphite used to simulate the coke grains of nuclear graphite. We used 13C implantation in order to simulate 14C displaced from its original structural site through recoil. The collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic damage. However, a part of the recoil carbon atom energy is also transferred to the graphite lattice through electronic excitation. The effects of the different irradiation regimes in synergy with temperature were simulated using ion irradiation by varying Sn(nuclear)/Se(electronic) stopping power. Thus, the samples were irradiated with different ions of different energies. The structure modifications were followed by High Resolution Transmission Electron Microscopy (HRTEM) and Raman microspectrometry. The results show that temperature generally counteracts the disordering effects of irradiation but the achieved reordering level strongly depends on the initial structural state of the graphite matrix. Thus, extrapolating to reactor conditions, for an initially highly disordered structure, irradiation at reactor temperatures (200 - 500 °C) should induce almost no change of the initial structure. On the contrary, when the structure is initially less disordered, there should be a "zoning" of the reordering: In "cold" high flux irradiated zones where the ballistic damage is important, the structure should be poorly reordered; In "hot" low flux irradiated zones where the ballistic
Brookhaven FASTBUS ADC's

International Nuclear Information System (INIS)

Schmidt, M.P.; Black, J.K.; Blatt, S.R.

1983-01-01

A high energy physics experiment has been performed at Brookhaven National Laboratory on k/sub L/ 0 → 2π 0 decays employing a large (> 200 element) lead glass array as an electromagnetic calorimeter. To acquire pulse height information from the detector we have constructed ADC modules in the context of the Brookhaven Fastbus data acquisition system. Digitization (8 bits) and encoding, including pedestal subtraction and data sparsing, is achieved for each 16 channel module in 6 μsec
Brookhaven highlights. [Fiscal year 1992, October 1, 1991--September 30, 1992

Energy Technology Data Exchange (ETDEWEB)

Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L. [eds.

1992-12-31

This publication provides a broad overview of the research programs and efforts being conducted, built, designed, and planned at Brookhaven National Laboratory. This work covers a broad range of scientific disciplines. Major facilities include the Alternating Gradient Synchrotron (AGS), with its newly completed booster, the National Synchrotron Light Source (NSLS), the High Flux Beam Reactor (HFBR), and the RHIC, which is under construction. Departments within the laboratory include the AGS department, accelerator development, physics, chemistry, biology, NSLS, medical, nuclear energy, and interdepartmental research efforts. Research ranges from the pure sciences, in nuclear physics and high energy physics as one example, to environmental work in applied science to study climatic effects, from efforts in biology which are a component of the human genome project to the study, production, and characterization of new materials. The paper provides an overview of the laboratory operations during 1992, including staffing, research, honors, funding, and general laboratory plans for the future.
Basic data for surveillance test on core support graphite structures for the high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Kikuchi, Takayuki; Iyoku, Tatsuo; Fujimoto, Nozomu; Ishihara, Masahiro; Sawa, Kazuhiro

2007-02-01

Both of the visual inspection by a TV camera and the measurement of material properties by surveillance test on core support graphite structures are planned for the High Temperature Engineering Test Reactor (HTTR) to confirm their structural integrity and characteristics. The surveillance test is aimed to investigate the change of material properties by aging effects such as fast neutron irradiation and oxidation. The obtained data will be used not only for evaluating the structural integrity of the core support graphite structures of the HTTR but also for design of advanced Very High Temperature Reactor (VHTR) discussed at generation IV international forum. This report describes the initial material properties of surveillance specimens before installation and installed position of surveillance specimens in the HTTR. (author)
Graphite oxidation and structural strength of graphite support column in VHTR

International Nuclear Information System (INIS)

Park, Byung Ha; No, Hee Cheno; Kim, Eung Soo; Oh, Chang H.

2009-01-01

The air-ingress event by a large pipe break is an important accident considered in design of very high-temperature gas-cooled reactors (VHTR). Core-collapse prediction is a main safety issue. Structural failure model are technically required. The objective of this study is to develop structural failure model for the supporting graphite material in the lower plenum of the GT-MHR (gas-turbine-modular high temperature reactor). Graphite support column is important for VHTR structural integrity. Graphite support columns are under the axial load. Critical strength of graphite column is related to slenderness ratio and bulk density. Through compression tests for fresh and oxidized graphite columns we show that compressive strength of IG-110 was 79.46 MPa. And, the buckling strength of IG-110 column was expressed by the empirical formula: σ 0 =σ straight-line - C L/r, σ straight-line =91.31 MPa, C=1.01. The results of uniform and non-uniform oxidation tests show that the strength degradation of oxidized graphite column is expressed in the following non-dimensional form: σ/σ 0 =exp(-kd), k=0.111. Also, from the results of the uniform oxidation test with a complicated-shape column, we found out that the above non-dimensional equation obtained from the uniform oxidation test is applicable to a uniform oxidation case with a complicated-shape column. (author)
Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)
Irradiation damage in graphite. The works of Professor B.T. Kelly

International Nuclear Information System (INIS)

Marsden, B.J.

1996-01-01

The irradiation damage produced in graphite by energetic neutrons (>100eV) has been extensively studied because of the use of graphite as a moderator in thermal nuclear reactors. In recent times, graphite has been adopted as the protective tiling of the inner wall of experimental fusion systems and property changes due to fusion neutrons have become important. The late Professor B.T. Kelly reviewed the work carried out on the irradiation behaviour of graphite since the 1940s. This work is particularly timely as the scale of research into the effects of fission neutrons has been greatly reduced and many of the active researchers have retired. In recent years, new programmes of work are being formulated for the use of graphite in both the field of high temperature reactor systems and fusion systems. It is therefore important that the knowledge gained by Professor Kelly and other workers is not lost but passed on to future generations of nuclear scientists and engineers. This paper reviews Professor Kelly's last work, it also draws on the experience gained during many long discussions with Brian during the years he worked closely with the present graphite team at AEA Technology. It is hoped to publish his work in full in the near future. (author). 13 refs, 14 figs, 3 tabs
Proton irradiated graphite grades for a long baseline neutrino facility experiment

International Nuclear Information System (INIS)

Simos, N.; Nocera, P.; Zwaska, R.; Mokhov, N.

2017-01-01

In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ~6.1×10"2"0 p/cm"2 and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ~10"2"0 cm"-"2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in
Proton irradiated graphite grades for a long baseline neutrino facility experiment

Directory of Open Access Journals (Sweden)

N. Simos

2017-07-01

Full Text Available In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF of the Deep Underground Neutrino Experiment (DUNE four graphite grades were irradiated with protons in the energy range of 140–180 MeV, to peak fluence of ∼6.1×10^{20} p/cm^{2} and irradiation temperatures between 120–200 °C. The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young’s modulus. The proton fluence level of ∼10^{20} cm^{−2} where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite
Proton irradiated graphite grades for a long baseline neutrino facility experiment

Science.gov (United States)

Simos, N.; Nocera, P.; Zhong, Z.; Zwaska, R.; Mokhov, N.; Misek, J.; Ammigan, K.; Hurh, P.; Kotsina, Z.

2017-07-01

In search of a low-Z pion production target for the Long Baseline Neutrino Facility (LBNF) of the Deep Underground Neutrino Experiment (DUNE) four graphite grades were irradiated with protons in the energy range of 140-180 MeV, to peak fluence of ˜6.1 ×1020 p /cm2 and irradiation temperatures between 120 - 200 °C . The test array included POCO ZXF-5Q, Toyo-Tanso IG 430, Carbone-Lorraine 2020 and SGL R7650 grades of graphite. Irradiation was performed at the Brookhaven Linear Isotope Producer. Postirradiation analyses were performed with the objective of (a) comparing their response under the postulated irradiation conditions to guide a graphite grade selection for use as a pion target and (b) understanding changes in physical and mechanical properties as well as microstructure that occurred as a result of the achieved fluence and in particular at this low-temperature regime where pion graphite targets are expected to operate. A further goal of the postirradiation evaluation was to establish a proton-neutron correlation damage on graphite that will allow for the use of a wealth of available neutron-based damage data in proton-based studies and applications. Macroscopic postirradiation analyses as well as energy dispersive x-ray diffraction of 200 KeV x rays at the NSLS synchrotron of Brookhaven National Laboratory were employed. The macroscopic analyses revealed differences in the physical and strength properties of the four grades with behavior however under proton irradiation that qualitatively agrees with that reported for graphite under neutrons for the same low temperature regime and in particular the increase of thermal expansion, strength and Young's modulus. The proton fluence level of ˜1020 cm-2 where strength reaches a maximum before it begins to decrease at higher fluences has been identified and it agrees with neutron-induced changes. X-ray diffraction analyses of the proton irradiated graphite revealed for the first time the similarity in
A woman like you: Women scientists and engineers at Brookhaven National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Benkovitz, Carmen; Bernholc, Nicole; Cohen, Anita; Eng, Susan; Enriquez-Leder, Rosario; Franz, Barbara; Gorden, Patricia; Hanson, Louise; Lamble, Geraldine; Martin, Harriet; Mastrangelo, Iris; McLane, Victoria; Villela, Maria-Alicia; Vivirito, Katherine; Woodhead, Avril

1991-01-01

This publication by the women in Science and Engineering introduces career possibilities in science and engineering. It introduces what work and home life are like for women who have already entered these fields. Women at Brookhaven National Laboratory work in a variety of challenging research roles -- from biologist and environmental scientist to safety engineer, from patent lawyer to technician. Brookhaven National Laboratory is a multi-program laboratory which carries out basic and applied research in the physical, biomedical and environmental sciences and in selected energy technologies. The Laboratory is managed by Associated University, Inc., under contract with the US Department of Energy. Brookhaven and the other national laboratories, because of their enormous research resources, can play a critical role in a education and training of the workforce.
Short-lived radionuclide production capability at the Brookhaven Linac Isotope Producer

International Nuclear Information System (INIS)

Mausner, L.F.; Richards, P.

1985-01-01

The Brookhaven National Linac Isotope Producer is the first facility to demonstrate the capability of a large linear accelerator for efficient and economical production of difficult-to-make, medically useful radionuclides. The linac provides a beam of 200-MeV protons at an integrated beam current of up to 60 μA. The 200-MeV proton energy is very suitable for isotope production because the spallation process can create radionuclides unavailable at lower energy accelerators or reactors. Several medically important short-lived radionuclides are presently being prepared for on-site and off-site collaborative research programs. These are iodine-123, iron-52, manganese-52m, ruthenium-97, and the rubidium-81-krypton-81m system. The production parameters for these are summarized
Review: BNL Tokamak graphite blanket design concepts

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1976-01-01

The BNL minimum activity graphite blanket designs are reviewed, and three are discussed in the context of an experimental power reactor (EPR) and commercial power reactor. Basically, the three designs employ a 30 cm or thicker graphite screen. Bremsstrahlung energy is deposited on the graphite surface and re-radiated away as thermal radiation. Fast neutrons are slowed down in the graphite, depositing most of their energy, which is then radiated to a secondary blanket with coolant tubes, as in types A and B, or removed by intermittent direct gas cooling (type C). In types A and B, radiation damage to the coolant tubes in the secondary blanket is reduced by one or two orders of magnitude, while in type C, the blanket is only cooled when the reactor is shut down, so that coolant cannot quench the plasma. (Auth.)
Preparation of pyrolytic carbon coating on graphite for inhibiting liquid fluoride salt and Xe{sup 135} penetration for molten salt breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Song, Jinliang [Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Zhao, Yanling, E-mail: jlsong1982@yeah.net [School of Materials Science and Engineering, University of Jinan, Jinan 250022 (China); He, Xiujie; Zhang, Baoliang [Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Xu, Li [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); He, Zhoutong; Zhang, DongSheng; Gao, Lina; Xia, Huihao; Zhou, Xingtai; Huai, Ping [Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Bai, Shuo [Shenyang National Laboratory for Materials Science, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China)

2015-01-15

Highlights: • Rough laminar pyrolytic carbon coating (RLPyC) is prepared by a fixed-bed method. • The salt-infiltration into IG-110 is 13.5%, less than 0.01% of RLPyC under 1.5 atm. • The helium diffusion coefficient of RLPyC coated graphite is 2.16 × 10{sup −8} cm{sup 2}/s. • The coated graphite can inhibit the liquid fluoride salt and Xe{sup 135} penetration. - Abstract: A fixed-bed deposition method was used to prepare rough laminar pyrolytic carbon coating (RLPyC) on graphite for inhibiting liquid fluoride salt and Xe{sup 135} penetration during use in molten salt breeder reactor. The RLPyC coating possessed a graphitization degree of 44% and had good contact with graphite substrate. A high-pressure reactor was constructed to evaluate the molten salt infiltration in the isostatic graphite (IG-110, TOYO TANSO CO., LTD.) and RLPyC coated graphite under 1.01, 1.52, 3.04, 5.07 and 10.13 × 10{sup 5} Pa for 12 h. Mercury injection and molten-salt infiltration experiments indicated the porosity and the salt-infiltration amount of 18.4% and 13.5 wt% under 1.52 × 10{sup 5} Pa of IG-110, which was much less than 1.2% and 0.06 wt% under 10.13 × 10{sup 5} Pa of the RLPyC, respectively. A vacuum device was constructed to evaluate the Xe{sup 135} penetration in the graphite. The helium diffusion coefficient of RLPyC coated graphite was 2.16 × 10{sup −12} m{sup 2}/s, much less than 1.21 × 10{sup −6} m{sup 2}/s of the graphite. Thermal cycle experiment indicated the coatings possessed excellent thermal stability. The coated graphite could effectively inhibit the liquid fluoride salt and Xe{sup 135} penetration.
Brookhaven highlights, October 1, 1989--September 30, 1990

Energy Technology Data Exchange (ETDEWEB)

Rowe, M.S.; Cohen, A.; Greenberg, D.; Seubert, L.; Kuper, J.B.H. (eds.)

1990-01-01

This report discusses research being conducted at Brookhaven National Laboratory. Highlights from all the department are illustrated. The main topics are on accelerator development and applications. (LSP)
Irradiation Creep in Graphite

Energy Technology Data Exchange (ETDEWEB)

Ubic, Rick; Butt, Darryl; Windes, William

2014-03-13

An understanding of the underlying mechanisms of irradiation creep in graphite material is required to correctly interpret experimental data, explain micromechanical modeling results, and predict whole-core behavior. This project will focus on experimental microscopic data to demonstrate the mechanism of irradiation creep. High-resolution transmission electron microscopy should be able to image both the dislocations in graphite and the irradiation-induced interstitial clusters that pin those dislocations. The team will first prepare and characterize nanoscale samples of virgin nuclear graphite in a transmission electron microscope. Additional samples will be irradiated to varying degrees at the Advanced Test Reactor (ATR) facility and similarly characterized. Researchers will record microstructures and crystal defects and suggest a mechanism for irradiation creep based on the results. In addition, the purchase of a tensile holder for a transmission electron microscope will allow, for the first time, in situ observation of creep behavior on the microstructure and crystallographic defects.
Experimental research of the yielding behavior of a graphite cylinder subjected to line loading

International Nuclear Information System (INIS)

Liu Hetong; Ma Qinwei; Ma Shaopeng; Wang Hongtao

2014-01-01

The graphite material cylinders are widely used in High-temperature gas-cooled reactor as connecting components. For engineering design, the deformation behavior, especially the yielding process of the graphite cylinder should be investigated in order to evaluate the carrying capacity of the cylinder. The yielding formation and propagation of a graphite cylinder subjected to line loading, which corresponds to the global behavior of the structure, was experimentally studied and evaluated by measuring the strain fields on the end of the cylinder using Digital Image Correlation. The global behavior of the structure is expressed by a relationship between the average stress (load divided by contact area) and the equivalent strain (ratio of half width of contact area to radius of the cylinder), the contact area was measured by identifying the color area of the pressure film in a new experiment which graphite component is loaded and unloaded continuously. A correspondence between the yielding state and the nonlinearity of the global behavior was constructed, as loading was increased, the cylinder was found to first yield at a specific point after which a yielding core formed and propagated. Before the yielding core propagated to the surface of the cylinder, the global behavior of the structure remained linear. After the yielding core propagated to the surface of the cylinder, the global behavior became nonlinear. The correspondence constructed in the paper will be helpful to understand the failure process and to evaluate the carrying capacity of a graphite cylinder subjected to line loading in reactors. (author)

Methods and codes for neutronic calculations of the MARIA research reactor

International Nuclear Information System (INIS)

Andrzejewski, K.; Kulikowska, T.; Bretscher, M.M.; Hanan, N.A.; Matos, J.E.

1998-01-01

The core of the MARIA high flux multipurpose research reactor is highly heterogeneous. It consists of beryllium blocks arranged in 6x8 matrix, tubular fuel assemblies, control rods and irradiation channels. The reflector is also heterogeneous and consists of graphite blocks clad with aluminium. Its structure is perturbed by the experimental beam tubes. This paper presents methods and codes used to calculate the MARIA reactor neutronics characteristics and experience gained thus far at IAE and ANL. At ANL the methods of MARIA calculations were developed in connection with RERTR program. At IAE the package of programs was developed to help its operator in optimization of fuel utilization. (author)
BROOKHAVEN: Getting a boost

International Nuclear Information System (INIS)

Anon.

1992-01-01

Some recent scenarios for the future of US high energy physics have glossed over the ongoing high energy physics role of Brookhaven's Alternating Gradient Synchrotron (AGS), seeing it being relegated essentially to a service function as the injector for the RHIC heavy ion collider. However this view is not shared by Brookhaven, now enjoying the benefits of a Booster to inject into the AGS
Graphite moderated 252Cf source

International Nuclear Information System (INIS)

Sajo B, L.; Barros, H.; Greaves, E. D.; Vega C, H. R.

2014-08-01

The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a 252 Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the 252 Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)
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.
Research Reactors Types and Utilization

International Nuclear Information System (INIS)

Abdelrazek, I.D.

2008-01-01

A nuclear reactor, in gross terms, is a device in which nuclear chain reactions are initiated, controlled, and sustained at a steady rate. The nuclei of fuel heavy atoms (mostly 235 U or 239 Pu), when struck by a slow neutron, may split into two or more smaller nuclei as fission products,releasing energy and neutrons in a process called nuclear fission. These newly-born fast neutrons then undergo several successive collisions with relatively low atomic mass material, the moderator, to become thermalized or slow. Normal water, heavy water, graphite and beryllium are typical moderators. These neutrons then trigger further fissions, and so on. When this nuclear chain reaction is controlled, the energy released can be used to heat water, produce steam and drive a turbine that generates electricity. The fission process, and hence the energy release, are controlled by the insertion (or extraction) of control rods through the reactor. These rods are strongly neutron absorbents, and thus only enough neutrons to sustain the chain reaction are left in the core. The energy released, mostly in the form of heat, should be continuously removed, to protect the core from damage. The most significant use of nuclear reactors is as an energy source for the generation of electrical power and for power in some military ships. This is usually accomplished by methods that involve using heat from the nuclear reaction to power steam turbines. Research reactors are used for radioisotope production and for beam experiments with free neutrons. Historically, the first use of nuclear reactors was the production of weapons grade plutonium for nuclear weapons. Currently all commercial nuclear reactors are based on nuclear fission. Fusion power is an experimental technology based on nuclear fusion instead of fission.
Reactor Operations informal monthly report December 1994

International Nuclear Information System (INIS)

1994-12-01

Reactor operations at the MRR and HFBR reactors at Brookhaven National Laboratory are presented for December 1994. Reactor run-time and power levels, instrumentation, mechanical maintenance, occurrence reports, and safety information are included
Applied programs at Brookhaven National Laboratory

Energy Technology Data Exchange (ETDEWEB)

1991-09-01

This document overviews the areas of current research at Brookhaven National Laboratory. Technology transfer and the user facilities are discussed. Current topics are presented in the areas of applied physics, chemical science, material science, energy efficiency and conservation, environmental health and mathematics, biosystems and process science, oceanography, and nuclear energy. (GHH)
Graphite Oxidation Simulation in HTR Accident Conditions

Energy Technology Data Exchange (ETDEWEB)

El-Genk, Mohamed

2012-10-19

Massive air and water ingress, following a pipe break or leak in steam-generator tubes, is a design-basis accident for high-temperature reactors (HTRs). Analysis of these accidents in both prismatic and pebble bed HTRs requires state-of-the-art capability for predictions of: 1) oxidation kinetics, 2) air helium gas mixture stratification and diffusion into the core following the depressurization, 3) transport of multi-species gas mixture, and 4) graphite corrosion. This project will develop a multi-dimensional, comprehensive oxidation kinetics model of graphite in HTRs, with diverse capabilities for handling different flow regimes. The chemical kinetics/multi-species transport model for graphite burning and oxidation will account for temperature-related changes in the properties of graphite, oxidants (O2, H2O, CO), reaction products (CO, CO2, H2, CH4) and other gases in the mixture (He and N2). The model will treat the oxidation and corrosion of graphite in geometries representative of HTR core component at temperatures of 900°C or higher. The developed chemical reaction kinetics model will be user-friendly for coupling to full core analysis codes such as MELCOR and RELAP, as well as computational fluid dynamics (CFD) codes such as CD-adapco. The research team will solve governing equations for the multi-dimensional flow and the chemical reactions and kinetics using Simulink, an extension of the MATLAB solver, and will validate and benchmark the model's predictions using reported experimental data. Researchers will develop an interface to couple the validated model to a commercially available CFD fluid flow and thermal-hydraulic model of the reactor , and will perform a simulation of a pipe break in a prismatic core HTR, with the potential for future application to a pebble-bed type HTR.
Use of research reactors for training and teaching nuclear sciences

International Nuclear Information System (INIS)

Safieh, J.; Gless, B.

2002-01-01

Training activities on reactors are organized by Cea on 2 specially dedicated reactors Ulysse (Saclay) and Siloette (Grenoble) and 2 research reactors Minerve (Cadarache) and Azur (Cadarache, facility managed by Technicatome). About 4000 students have been trained on Ulysse since its commissioning more than 40 years ago. The concept that led to the design of Ulysse was to build a true reactor dedicated to teaching and training activities and that was able to operate with great flexibility and under high conditions of safety, this reactor is inspired from the Argonaut-type reactor. The main specificities of Ulysse are: a nominal power of 100 kW, a maximal thermal neutron flux of 1.4 10 12 n.cm -2 .s -1 , a 90 % enriched fuel, a graphite reflector, the use of water as coolant and moderator, and 6 cadmium plates as control rods. Ulysse allows students to get practical experience on a large range of topics: approach to criticality, effect of the starting neutron source, calibration of control rods, distribution of the neutron flux in the thermal column, temperature coefficient, radiation detectors, neutron activation analysis, and radioprotection. (A.C.)
Activities of the cross-section compilation and evaluation centers at the Brookhaven National Laboratory

International Nuclear Information System (INIS)

Chernick, J.

1967-01-01

The growth of the compilation and evaluation efforts at the Brookhaven National Laboratory are reviewed. The current work of the Sigma Center is discussed, including the status of the publication of supplements to BNL-325 and the current state of the SCISRS-I tape. Future needs for BNL-325 type publications and SCISRS-II cross-section tapes are outlined. The history of the Cross-Section Evaluation Center at the Brookhaven National Laboratory is similarly reviewed. The status of current work is discussed, including the growth of the ENDF/A tape. The status of US efforts to produce a cross-section tape (ENDF7B) at an early date to satisfy the needs of US reactor designers is discussed. The continued importance of integral experiments and their accurate analysis to provide checks of the cross-section tapes is pointed out. The role of the Brookhaven National Laboratory in collaboration on an international basis is reviewed, including its current relationship to the ENEA Neutron Data Compilation Centre, the International Atomic Energy Agency and other nuclear centres. (author)
Analysis of Wigner energy release process in graphite stack of shut-down uranium-graphite reactor

OpenAIRE

Bespala, E. V.; Pavliuk, A. O.; Kotlyarevskiy, S. G.

2015-01-01

Data, which finding during thermal differential analysis of sampled irradiated graphite are presented. Results of computational modeling of Winger energy release process from irradiated graphite staking are demonstrated. It's shown, that spontaneous combustion of graphite possible only in adiabatic case.
A preliminary definition of the parameters of an experimental natural - uranium, graphite - moderated, helium - cooled power reactor

International Nuclear Information System (INIS)

Baltazar, O.

1978-01-01

A preliminary study of the technical characteristic of an experiment at 32 MWe power with a natural uconium, graphite-moderated, helium cooled reactor is described. The national participation and the use of reactor as an instrument for the technological development of future high temperature gas cooled reactor is considered in the choice of the reactor type. Considerations about nuclear power plants components based in extensive bibliography about similar english GCR reactor is presented. The main thermal, neutronic an static characteristic and in core management of the nuclear fuel is stablished. A simplified scheme of the secondary system and its thermodynamic performance is determined. A scheme of parameters calculation of the reactor type is defined based in the present capacity of calculation developed by Coordenadoria de Engenharia Nuclear and Centro de Processamento de Dados, IEA, Brazil [pt
Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt
Effect of total pressure on graphite oxidation

International Nuclear Information System (INIS)

Burnette, R.D.; Hoot, C.G.

1983-04-01

Graphite corrosion in the high-temperature gas-cooled reactor (HTGR) is calculated using two key assumptions: (1) the kinetic, catalysis, and transport characteristics of graphite determined by bench-scale tests apply to large components at reactor conditions and (2) the effects of high pressure and turbulent flow are predictable. To better understand the differences between laboratory tests and reactor conditions, a high-pressure test loop (HPTL) has been constructed and used to perform tests at reactor temperature, pressure, and flow conditions. The HPTL is intended to determine the functional dependence of oxidation rate and characteristics on total pressure and gas velocity and to compare the oxidation results with calculations using models and codes developed for the reactor
Development of in-service inspection system for core support graphite structures in the high temperature engineering test reactor (HTTR)

Energy Technology Data Exchange (ETDEWEB)

Sumita, Junya; Hanawa, Satoshi; Kikuchi, Takayuki; Ishihara, Masahiro [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2003-03-01

Visual inspection of core support graphite structures using TV camera as in-service inspection and measurement of material characteristics using surveillance test specimens are planned in the High Temperature Engineering Test Reactor (HTTR) to confirm structural integrity of the core support graphite structures. For the visual inspection, in-service inspection system developed from September 1996 to June 1998, and pre-service inspection using the system was carried out. As the result of the pre-service inspection, it was validated that high quality of visual inspection with TV camera can be carried out, and also structural integrity of the core support graphite structures at the initial stage of the HTTR operation was confirmed. (author)
Production of nuclear graphite in France; Production de graphite nucleaire en France

Energy Technology Data Exchange (ETDEWEB)

Legendre, P; Mondet, L [Societe Pechiney, 74 - Chedde (France); Arragon, Ph; Cornuault, P; Gueron, J; Hering, H [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The graphite intended for the construction of the reactors is obtained by the usual process: confection of a cake from coke of oil and tar, cooked (in a electric oven) then the product of cook is graphitized, also by electric heating. The use of the air transportation and the control of conditions cooking and graphitization have permitted to increase the nuclear graphite production as well as to better control their physical and mechanical properties and to reduce to the minimum the unwanted stains. (M.B.) [French] Le graphite destine a la construction des reacteurs est obtenu par le procede usuel: confection d'une pate a partir de coke de petrole et de brai, cuisson de cette pate (au four electrique) puis graphitation du produit cuit, egalement par chauffage electrique. L'usage du transport pneumatique et le controle des conditions cuisson et de graphitation ont permit d'augmenter la production de graphite nucleaire ainsi que de mieux controler ses proprietes physiques et mecaniques et de reduire au minimum les souillures accidentelles. (M.B.)
Change in physical properties of high density isotropic graphites irradiated in the ?JOYO? fast reactor

Science.gov (United States)

Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

1995-08-01

Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor "JOYO" to fluences from 2.11 to 2.86 × 10 26 n/m 2 ( E > 0.1 MeV) at temperatures from 549 to 597°C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens.
Change in physical properties of high density isotropic graphites irradiated in the ''JOYO'' fast reactor

International Nuclear Information System (INIS)

Maruyama, T.; Kaito, T.; Onose, S.; Shibahara, I.

1995-01-01

Thirteen kinds of isotropic graphites with different density and maximum grain size were irradiated in the experimental fast reactor ''JOYO'' to fluences from 2.11 to 2.86x10 26 n/m 2 (E>0.1 MeV) at temperatures from 549 to 597 C. Postirradiation examination was carried out on the dimensional changes, elastic modulus, and thermal conductivity of these materials. Dimensional change results indicate that the graphites irradiated at lower fluences showed shrinkage upon neutron irradiation followed by increase with increasing neutron fluences, irrespective of differences in material parameters. The Young's modulus and Poisson's ratio increased by two to three times the unirradiated values. The large scatter found in Poisson's ratio of unirradiated materials became very small and a linear dependence on density was obtained after irradiation. The thermal conductivity decreased to one-fifth to one-tenth of unirradiated values, with a negligible change in specific heat. The results of postirradiation examination indicated that the changes in physical properties of high density, isotropic graphites were mainly dominated by the irradiation condition rather than their material parameters. Namely, the effects of irradiation induced defects on physical properties of heavily neutron-irradiated graphites are much larger than that of defects associated with as-fabricated specimens. (orig.)
Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors

International Nuclear Information System (INIS)

Boudouresque, B.; Courcon, P.; Lestiboubois, G.

1964-01-01

The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm 2 gas pressure, should remain in contact with the fuel. (authors) [fr
Progress in radioactive graphite waste management. Additional information

International Nuclear Information System (INIS)

2010-06-01

Radioactive graphite constitutes a major waste stream which arises during the decommissioning of certain types of nuclear installations. Worldwide, a total of around 250 000 tonnes of radioactive graphite, comprising graphite moderators and reflectors, will require management solutions in the coming years. 14 C is the radionuclide of greatest concern in nuclear graphite; it arises principally through the interaction of reactor neutrons with nitrogen, which is present in graphite as an impurity or in the reactor coolant or cover gas. 3 H is created by the reactions of neutrons with 6 Li impurities in graphite as well as in fission of the fuel. 36 Cl is generated in the neutron activation of chlorine impurities in graphite. Problems in the radioactive waste management of graphite arise mainly because of the large volumes requiring disposal, the long half-lives of the main radionuclides involved and the specific properties of graphite - such as stored Wigner energy, graphite dust explosibility and the potential for radioactive gases to be released. Various options for the management of radioactive graphite have been studied but a generally accepted approach for its conditioning and disposal does not yet exist. Different solutions may be appropriate in different cases. In most of the countries with radioactive graphite to manage, little progress has been made to date in respect of the disposal of this material. Only in France has there been specific thinking about a dedicated graphite waste-disposal facility (within ANDRA): other major producers of graphite waste (UK and the countries of the former Soviet Union) are either thinking in terms of repository disposal or have no developed plans. A conference entitled 'Solutions for Graphite Waste: a Contribution to the Accelerated Decommissioning of Graphite Moderated Nuclear Reactors' was held at the University of Manchester 21-23 March 2007 in order to stimulate progress in radioactive graphite waste management

Impact-Contact Analysis of Prismatic Graphite Blocks Using Abaqus

International Nuclear Information System (INIS)

Kang, Ji Ho; Kim, Gyeong Ho; Choi, Woo Seok

2010-12-01

Graphite blocks are the important core components of the high temperature gas-cooled reactor. As these blocks are simply stacked in array, collisions among neighboring components may occur during earthquakes or accidents. The final objective of the research project is to develop a reliable seismic model of the stacked graphite blocks from which their behavior can be predicted and, thus, they are designed to have sufficient strength to maintain their structural integrity during the anticipated occurrences. The work summarized in this report is a first step toward the big picture and is dedicated to build a realistic impact-contact dynamics model of the graphite block using a commercial FEM package, Abaqus. The developed model will be further used to assist building a reliable lumped dynamics model of these stacked graphite components
BROOKHAVEN: Japanese collaboration

International Nuclear Information System (INIS)

Thieberger, Peter

1995-01-01

Full text: The Japanese RIKEN Laboratory is contributing $20 million to help construct the RHIC Relativistic Heavy Ion Collider now being built at Brookhaven and due to be completed in 1999. In return, RIKEN will participate in research at RHIC. RHIC is being built to collide beams of heavy ions at energies of about 100 GeV per nucleon to explore hot and dense states of nuclear matter, with the ultimate aim of finding the quark-gluon plasma, the medium which existed in the fiery aftermath of the Big Bang before subsequently 'freezing' into nucleons. However another long-time Brookhaven speciality is handling beams of polarized (spin-oriented) protons in the 30 GeV AGS Alternating Gradient Synchrotron, which will act as the injector for RHIC. With the involvement of RIKEN, the RHIC programme now expands to cover polarized protons. Half of the RIKEN support will be used to build and install the special hardware needed to handle the polarized beams in RHIC. This includes 'Siberian Snakes' to negotiate depolarizing resonances which would otherwise mar beam acceleration (September 1994, page 27). The remaining RIKEN funding will go towards additional equipment for the PHENIX detector (May 1992, page 10) to enable it to cover spin physics. This equipment includes a second muon arm, with a magnet and tracking chamber. A multidisciplinary laboratory, RIKEN - Rikagaku Kenkyusho, or the Institute of Physical and Chemical Research - near Tokyo is currently the scene of construction of an 8 GeV synchrotron X-ray source
From Core to Capture: Graphite Management by Gasification and Carbon Capture & Storage (CCS)

International Nuclear Information System (INIS)

Goodwin, J.; Bradbury, D.; Black, S.; Tomlinson, T.; Livesey, B.; Robinson, J.; Lindberg, M.; Newton, C.; Jones, A.; Wickham, A.

2016-01-01

Radioactive graphite waste arises principally from the moderators of graphite/gas-cooled reactors at the end of life of the reactors. Commercial power producing reactors (for example, Magnox, AGR and RBMK) have graphite moderators, each containing several thousand tonnes of graphite, with the UK having the largest inventory of over 90,000 tonnes. Additionally, there are smaller quantities of graphite arising from other sources such as fuel element components. The current long term strategy for management of reactor graphite in the UK is for these wastes to be conditioned for disposal followed by transfer to a geological disposal facility (GDF). With this baseline position, these wastes will account for about 30% of the ILW inventory in a GDF. As the volume of the graphite waste is so large, it is not currently economic to retrieve and process the graphite in advance of the availability of a geological disposal facility. Recent work by the NDA has ascribed a much smaller “incremental” volume of 2% due to graphite, calculated on the basis that the GDF has to be a certain size anyway in order to dissipate the decay heat from high level waste
Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor
Nondestructive evaluation of nuclear-grade graphite

Science.gov (United States)

Kunerth, D. C.; McJunkin, T. R.

2012-05-01

The material of choice for the core of the high-temperature gas-cooled reactors being developed by the U.S. Department of Energy's Next Generation Nuclear Plant Program is graphite. Graphite is a composite material whose properties are highly dependent on the base material and manufacturing methods. In addition to the material variations intrinsic to the manufacturing process, graphite will also undergo changes in material properties resulting from radiation damage and possible oxidation within the reactor. Idaho National Laboratory is presently evaluating the viability of conventional nondestructive evaluation techniques to characterize the material variations inherent to manufacturing and in-service degradation. Approaches of interest include x-ray radiography, eddy currents, and ultrasonics.
Release from control of inactive material from decommissioning the ASTRA research reactor

International Nuclear Information System (INIS)

Brandl, A.; Hrnecek, E.; Steger, F.; Kurz, H.; Meyer, F.; Karacson, P.

2003-01-01

The Austrian Research Centers Seibersdorf have been operating a 10 MW ASTRA research reactor from 1960 until 1999. After that date, a submission of the intention to decommission the reactor has been provided to the Competent Authorities. After completion of an Environmental Impact Study by the Competent Authorities and modification of the Permissions for Site Use, the reactor finally entered the decommissioning phase in 2003. Inactive materials from the decommissioning site are expected to be released from control. The procedure for such a release from control agreed upon between the Competent Authorities and ARC Seibersdorf involves a four-step measurement, verification, and certification process detailed in this paper. By September 2003, this four-step procedure has been completed for 16500 kg of steel re-enforced concrete and for 5500 kg of other materials; the release from control of 3000 kg of paraffin and 10000 kg of graphite from the thermal column are planned for the near future. (author)
The Dragon reactor experiment

International Nuclear Information System (INIS)

Anon.

1976-01-01

The concept on which the Dragon Reactor Experiment was based was evolved at the Atomic Energy Research Establishment at Harwell in 1956, and in February of that year a High Temperature Gas- cooled Reactor Project Group was set up to study the feasibility of a helium-cooled reactor with a graphite or beryllium moderator, and with the emphasis on the thorium fuel cycle [af
Graphite moderated {sup 252}Cf source

Energy Technology Data Exchange (ETDEWEB)

Sajo B, L.; Barros, H.; Greaves, E. D. [Universidad Simon Bolivar, Nuclear Physics Laboratory, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of); Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2014-08-15

The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a {sup 252}Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the {sup 252}Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)
A review of the behaviour of graphite under the conditions appropriate for protection of the first wall of a fusion reactor

International Nuclear Information System (INIS)

Birch, M.; Brocklehurst, J.E.

1987-12-01

The material used as a first wall protection in fusion reactor systems will be exposed to 14 MeV neutrons from the fusion reaction and suffer surface bombardment by other energetic particles in the plasma. Graphite is a potential candidate for the first wall material. Calculations are performed of the damaging power of 14 MeV neutrons so that existing graphite irradiation data can be utilised. Such data at high irradiation temperatures are reviewed for a wide range of graphite types, characterised by specific examples, and the application of the data to design calculations is discussed. The erosion/corrosion effect of the plasma at the graphite surface is also considered. Limitations in the state of knowledge are identified, and particular areas of further work are recommended. (author)
Assessments of the stresses and deformations in an RBMK graphite moderator brick

International Nuclear Information System (INIS)

Jones, C.J.; Davies, M.A.; Marsden, B.J.; Bougaenko, S.E.; Baldin, V.D.; Demintievski, V.N.; Rodtchenkov, B.S.; Sinitsyn, E.N.

1996-01-01

The RBMK reactors, designed by RDIPE (Moscow), are graphite moderated and cooled by light water. Graphite dimensions and thermo-mechanical properties change significantly in a complex manner during reactor life due to fast neutron damage and these changes have implications on the safe operation of all graphite moderated reactors. A joint programme of work is being carried out between AEA Technology (UK) and RDIPE (Russia) to assess the life of the RBMK graphite stack under normal operating conditions. The programme has included the modelling of graphite dimensional changes due to irradiation through reactor life and the assessment of the implications of these changes on the stresses and deformations in the graphite stack. Calculations have been carried out to assess the deformations of a moderator brick over a period from start of life up to 30 years of operation. The assessment have also included an analysis of the stresses in the bricks so that the time to brick failure could be determined. This paper describes the RBMK core design, the data and assessment methodology used in the analysis of the RBMK core and presents some results from analyses of the Leningrad Unit 1 RBMK reactor. (author). 2 refs, 8 figs
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
A plasma arc reactor for fullerene research

Science.gov (United States)

Anderson, T. T.; Dyer, P. L.; Dykes, J. W.; Klavins, P.; Anderson, P. E.; Liu, J. Z.; Shelton, R. N.

1994-12-01

A modified Krätschmer-Huffman reactor for the mass production of fullerenes is presented. Fullerene mass production is fundamental for the synthesis of higher and endohedral fullerenes. The reactor employs mechanisms for continuous graphite-rod feeding and in situ slag removal. Soot collects into a Soxhlet extraction thimble which serves as a fore-line vacuum pump filter, thereby easing fullerene separation from soot. Thermal gravimetric analysis (TGA) for yield determination is reported. This TGA method is faster and uses smaller samples than Soxhlet extraction methods which rely on aromatic solvents. Production of 10 g of soot per hour is readily achieved utilizing this reactor. Fullerene yields of 20% are attained routinely.
Safeguarding research reactors

International Nuclear Information System (INIS)

Powers, J.A.

1983-03-01

The report is organized in four sections, including the introduction. The second section contains a discussion of the characteristics and attributes of research reactors important to safeguards. In this section, research reactors are described according to their power level, if greater than 25 thermal megawatts, or according to each fuel type. This descriptive discussion includes both reactor and reactor fuel information of a generic nature, according to the following categories. 1. Research reactors with more than 25 megawatts thermal power, 2. Plate fuelled reactors, 3. Assembly fuelled reactors. 4. Research reactors fuelled with individual rods. 5. Disk fuelled reactors, and 6. Research reactors fuelled with aqueous homogeneous fuel. The third section consists of a brief discussion of general IAEA safeguards as they apply to research reactors. This section is based on IAEA safeguards implementation documents and technical reports that are used to establish Agency-State agreements and facility attachments. The fourth and last section describes inspection activities at research reactors necessary to meet Agency objectives. The scope of the activities extends to both pre and post inspection as well as the on-site inspection and includes the examination of records and reports relative to reactor operation and to receipts, shipments and certain internal transfers, periodic verification of fresh fuel, spent fuel and core fuel, activities related to containment and surveillance, and other selected activities, depending on the reactor
Study on efficient methods for removal and treatment of graphite blocks in a gas cooled reactor

International Nuclear Information System (INIS)

Fujii, S.; Shirakawa, M.; Murakami, T.

2001-01-01

Tokai Power Station (GCR, 166 MWe) started its commercial operation on July 1966 and ceased activities at the end of March 1998 after 32 years of operation. The decommissioning plans are being developed, to prepare for near future dismantling. In the study, the methods for removal of the graphite blocks of about 1,600 ton have been developed to carrying it out safely and in a short period of time, and the methods of treatment of graphite have also been developed. All technological items have been identified for which R and D work will be required for removal from the core and treatment for disposal. (1) In order to reduce the programme required for the dismantling of reactor internals, an efficient method for removal of the graphite blocks is necessary. For this purpose the design of a dismantling machine has been investigated which can extract several blocks at a time. The conceptual design has being developed and the model has been manufactured and tested in a mock-up facility. (2) In order to reduce disposal costs, it will be necessary to segment the graphite blocks, maximising the packing density available in the disposal containers. Some of the graphite blocks will be cut into pieces longitudinally by a remote machine. Relevant technical matters have been identified, such as graphite cutting methods, the nature of fine particles arising from the cutting operation, the treatment of fine particles for disposal, and the method of mortar filling inside the waste container. (author)
Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

Zou, C.Y.; Cai, X.Z.; Jiang, D.Z.; Yu, C.G.; Li, X.X.; Ma, Y.W.; Han, J.L. [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China); Chen, J.G., E-mail: chenjg@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Center for Excellence in TMSR Energy System, Chinese Academy of Sciences, Shanghai 201800 (China); Key Laboratory of Nuclear Radiation and Nuclear Energy Technology, Chinese Academy of Sciences, Shanghai 201800 (China)

2015-01-15

Highlights: • The temperature feedback coefficient with different moderation ratios for TMSR in thermal neutron region is optimized. • The breeding ratio and doubling time of a thermal TMSR with three different reprocessing schemes are analyzed. • The smaller hexagon size and larger salt fraction with more negative feedback coefficient can better satisfy the safety demands. • A shorter reprocessing time can achieve a better breeding ratio in a thermal TMSR. • The graphite moderator lifespan is compared with other MSRs and discussed. - Abstract: Molten salt reactor (MSR) has fascinating features: inherent safety, no fuel fabrication, online fuel reprocessing, etc. However, the graphite moderated MSR may present positive feedback coefficient which has severe implications for the transient behavior during operation. In this paper, the feedback coefficient and the breeding ratio are optimized based on the fuel-to-graphite ratio variation for a thorium based MSR (TMSR). A certain thermal core with negative feedback coefficient and relative high initial breeding ratio is chosen for the reprocessing scheme analysis. The breeding performances for the TMSR under different online fuel reprocessing efficiencies and frequencies are evaluated and compared with other MSR concepts. The results indicate that the thermal TMSR can get a breeding ratio greater than 1.0 with appropriate reprocessing scheme. The low fissile inventory in thermal TMSR leads to a short doubling time and low transuranic (TRU) inventory. The lifetime of graphite used for the TMSR is also discussed.
Influence of irradiation on high-strength graphites

International Nuclear Information System (INIS)

Virgil'ev, Yu.S.; Grebennik, V.N.; Kalyagina, I.P.

1989-01-01

To ensure efficiency of the graphite elements of the construction of the masonry of reactors, the graphite must possess high radiation stability, strength, and heat resistance. In this connection, the physical properties of graphites based on uncalcined petroleum coke with a binder - high-temperature hard coal pitch - the amount of which reaches 40% are considered in this paper
Production of nuclear graphite in France

International Nuclear Information System (INIS)

Legendre, P.; Mondet, L.; Arragon, Ph.; Cornuault, P.; Gueron, J.; Hering, H.

1955-01-01

The graphite intended for the construction of the reactors is obtained by the usual process: confection of a cake from coke of oil and tar, cooked (in a electric oven) then the product of cook is graphitized, also by electric heating. The use of the air transportation and the control of conditions cooking and graphitization have permitted to increase the nuclear graphite production as well as to better control their physical and mechanical properties and to reduce to the minimum the unwanted stains. (M.B.) [fr
BROOKHAVEN: RHIC installation

International Nuclear Information System (INIS)

Anon.

1994-01-01

This summer, the first superconducting magnet was installed in 3.8 kilometre tunnel for Brookhaven's RHIC heavy ion collider (October, page 31). Manufactured by Northrop Grumman's Electronics and System Integration Division, the magnet is the first of RHIC's 373 dipoles. In addition to the dipoles, Northrop Grumman will also provide 432 RHIC quadrupoles. The first quadrupole was delivered on 8 April, a month before the first dipole arrived for onsite testing prior to installation. RHIC will need 1,700 superconducting magnets - dipoles, quadrupoles, sextupoles and correcting magnets, 1,200 of which will be built by industry and the rest built at Brookhaven. The 300 sextupoles are being supplied by Everson Electric
Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

International Nuclear Information System (INIS)

Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

2010-01-01

The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.
Very high temperature gas-cooled reactor critical facility for Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Ishihara, Noriyuki

1985-01-01

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

Research on the phenomenon of graphitization. Crystallographic study - Study of bromine sorption

International Nuclear Information System (INIS)

Maire, Jacques

1967-01-01

This research thesis reports the study of the mechanism of graphitization of carbon by using X-ray diffraction analysis and the physical and chemical study of lamellar reactions between carbon and bromine. The author first presents generalities and results of preliminary studies (meaning of graphitization, presentation of the various carbon groups and classes), and then reports the study of the graphitization of compact carbons (soft carbons). More precisely, he reports the crystallographic study of partially graphitized carbons: methods and principles, experimental results and their analysis, discussion of the graphitization mechanism. In the next part, the author reports the study of bromine sorption on carbons: experimental method, isotherms of a natural graphite and of a graphitized carbon, structure of carbon-bromine complexes, isotherms of graphitizable carbons and of all other carbons, distribution of bromine layers in partially graphitized carbons, bromine sorption and Fermi level
Protection of nuclear graphite toward fluoride molten salt by glassy carbon deposit

International Nuclear Information System (INIS)

Bernardet, V.; Gomes, S.; Delpeux, S.; Dubois, M.; Guerin, K.; Avignant, D.; Renaudin, G.; Duclaux, L.

2009-01-01

Molten salt reactor represents one of the promising future Generation IV nuclear reactors families where the fuel, a liquid molten fluoride salt, is circulating through the graphite reactor core. The interactions between nuclear graphite and fluoride molten salt and also the graphite surface protection were investigated in this paper by powder X-ray diffraction, micro-Raman spectroscopy and scanning electron microscopy coupled with X-ray microanalysis. Nuclear graphite discs were covered by two kinds of protection deposit: a glassy carbon coating and a double coating of pyrolitic carbon/glassy carbon. Different behaviours have been highlighted according to the presence and the nature of the coated protection film. Intercalation of molten salt between the graphite layers did not occur. Nevertheless the molten salt adhered more or less to the surface of the graphite disc, filled more or less the graphite surface porosity and perturbed more or less the graphite stacking order at the disc surface. The behaviour of unprotected graphite was far to be satisfactory after two days of immersion of graphite in molten salt at 500 deg. C. The best protection of the graphite disc surface, with the maximum of inertness towards molten salt, has been obtained with the double coating of pyrolitic carbon/glassy carbon
Failure Predictions for Graphite Reflector Bricks in the Very High Temperature Reactor with the Prismatic Core Design

Energy Technology Data Exchange (ETDEWEB)

Singh, Gyanender, E-mail: sing0550@umn.edu [Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States); Fok, Alex [Minnesota Dental Research in Biomaterials and Biomechanics, School of Dentistry, University of Minnesota, 515, Delaware St. SE, Minneapolis, MN 55455 (United States); Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States); Mantell, Susan [Department of Mechanical Engineering, University of Minnesota, 111, Church St. SE, Minneapolis, MN 55455 (United States)

2017-06-15

Highlights: • Failure probability of VHTR reflector bricks predicted though crack modeling. • Criterion chosen for defining failure strongly affects the predictions. • Breaching of the CRC could be significantly delayed through crack arrest. • Capability to predict crack initiation and propagation demonstrated. - Abstract: Graphite is used in nuclear reactor cores as a neutron moderator, reflector and structural material. The dimensions and physical properties of graphite change when it is exposed to neutron irradiation. The non-uniform changes in the dimensions and physical properties lead to the build-up of stresses over the course of time in the core components. When the stresses reach the critical limit, i.e. the strength of the material, cracking occurs and ultimately the components fail. In this paper, an explicit crack modeling approach to predict the probability of failure of a VHTR prismatic reactor core reflector brick is presented. Firstly, a constitutive model for graphite is constructed and used to predict the stress distribution in the reflector brick under in-reactor conditions of high temperature and irradiation. Fracture simulations are performed as part of a Monte Carlo analysis to predict the probability of failure. Failure probability is determined based on two different criteria for defining failure time: A) crack initiation and B) crack extension to near control rod channel. A significant difference is found between the failure probabilities based on the two criteria. It is predicted that the reflector bricks will start cracking during the time range of 5–9 years, while breaching of the control rod channels will occur during the period of 11–16 years. The results show that, due to crack arrest, there is a significantly delay between crack initiation and breaching of the control rod channel.
NDA Position on the UK Management of Waste Graphite (December 2013)

International Nuclear Information System (INIS)

Norris, S.

2016-01-01

The purpose of this paper is to summarise a number of pieces of work that have been undertaken by the Nuclear Decommissioning Authority (NDA) to better understand the challenges of managing radioactive graphite wastes, these have led to an updated strategic position on graphite waste management. The updated strategic position takes into consideration Government’s response to Recommendation 8 from the Committee on Radioactive Waste Management’s (CoRWM), and provides the current NDA strategic position alongside circumstances where this should be reviewed. Two studies that provided input to this position are: 1. Operational Graphite Management Strategy: Credible and Preferred Options (Gate A & B); 2. The Long-term Management of Reactor Core Graphite Waste: Credible Options (Gate A). The paper highlights the key findings from the following work that has been undertaken to better inform this position: • A review by the NDA Radioactive Waste Management Directorate (RWMD)1 of the current baseline for managing radioactive graphite in England and Wales of geological disposal. The review identified some areas for optimisation and provided clarification on some aspects of the baseline e.g. the assumed ‘footprint’ of graphite wastes for a future Geological Disposal Facility (GDF). • Investigations into suitability of near-surface disposal options for graphite wastes. This included a review of the Low Level Waste Repository (LLWR) Ltd's new Environmental Safety Case (ESC) to assess the potential for graphite disposal and a feasibility study into a near-surface disposal facility for Higher Activity Waste (HAW) graphite at the Hunterston A site. • Continued monitoring of potential future treatment options. • Detailed characterisation work under the NDA’s Direct Research Portfolio using computer modelling and sample analysis to better understand any limitations of the current inventory data for graphite wastes. • Graphite behaviour work under the NDA�
On estimating the fracture probability of nuclear graphite components

International Nuclear Information System (INIS)

Srinivasan, Makuteswara

2008-01-01

The properties of nuclear grade graphites exhibit anisotropy and could vary considerably within a manufactured block. Graphite strength is affected by the direction of alignment of the constituent coke particles, which is dictated by the forming method, coke particle size, and the size, shape, and orientation distribution of pores in the structure. In this paper, a Weibull failure probability analysis for components is presented using the American Society of Testing Materials strength specification for nuclear grade graphites for core components in advanced high-temperature gas-cooled reactors. The risk of rupture (probability of fracture) and survival probability (reliability) of large graphite blocks are calculated for varying and discrete values of service tensile stresses. The limitations in these calculations are discussed from considerations of actual reactor environmental conditions that could potentially degrade the specification properties because of damage due to complex interactions between irradiation, temperature, stress, and variability in reactor operation
A woman like you: Women scientists and engineers at Brookhaven National Laboratory. Careers in action

Energy Technology Data Exchange (ETDEWEB)

1991-12-31

This publication by the women in Science and Engineering introduces career possibilities in science and engineering. It introduces what work and home life are like for women who have already entered these fields. Women at Brookhaven National Laboratory work in a variety of challenging research roles -- from biologist and environmental scientist to safety engineer, from patent lawyer to technician. Brookhaven National Laboratory is a multi-program laboratory which carries out basic and applied research in the physical, biomedical and environmental sciences and in selected energy technologies. The Laboratory is managed by Associated University, Inc., under contract with the US Department of Energy. Brookhaven and the other national laboratories, because of their enormous research resources, can play a critical role in a education and training of the workforce.
Graphite waste incineration in a fluidized bed

International Nuclear Information System (INIS)

Guiroy, J.J.

1996-01-01

French gas-cooled reactors belonging to the Atomic Energy Commission (CEA), Electricite de France (EDF), Hifrensa (Spain), etc., commissioned between the 1950s and 1970s, have generated large quantities of graphite wastes, mainly in the form of spent fuel sleeves. Furthermore, some of these reactors scheduled for dismantling in the near future (such as the G2 and G3 reactors at Marcoule) have cores consisting of graphite blocks. Consequently, a fraction of the contaminated graphite, amounting to 6000 t in France for example, must be processed in the coming years. For this processing, incineration using a circulating fluidized bed combustor has been selected as a possible solution and validated. However, the first operation to be performed involves recovering this graphite waste, and particularly, first of all, the spent fuel sleeves that were stored in silos during the years of reactor operation. Subsequent to the final shutdown of the Spanish gas-cooled reactor unit, Vandellos 1, the operating utility Hifrensa awarded contracts to a Framatome Iberica SA/ENSA consortium for removing, sorting, and prepackaging of the waste stored in three silos on the Vandellos site, essentially graphite sleeves. On the other hand, a program to validate the Framatome fluidized bed incineration process was carried out using a prototype incinerator installed at Le Creusot, France. The validation program included 22 twelve-hour tests and one 120-hour test. Particular attention was paid to the safety aspects of this project. During the performance of the validation program, a preliminary safety assessment was carried out. An impact assessment was performed with the help of the French Institute for Protection and Nuclear Safety, taking into account the preliminary spectra supplied by the CEA and EDF, and the activities of the radionuclides susceptible of being released into the atmosphere during the incineration. (author). 4 refs, 11 figs, 1 tab
Brookhaven National Laboratory's Accelerator Test Facility: research highlights and plans

Science.gov (United States)

Pogorelsky, I. V.; Ben-Zvi, I.

2014-08-01

The Accelerator Test Facility (ATF) at Brookhaven National Laboratory has served as a user facility for accelerator science for over a quarter of a century. In fulfilling this mission, the ATF offers the unique combination of a high-brightness 80 MeV electron beam that is synchronized to a 1 TW picosecond CO2 laser. We unveil herein our plan to considerably expand the ATF's floor space with an upgrade of the electron beam's energy to 300 MeV and the CO2 laser's peak power to 100 TW. This upgrade will propel the ATF even further to the forefront of research on advanced accelerators and radiation sources, supporting the most innovative ideas in this field. We discuss emerging opportunities for scientific breakthroughs, including the following: plasma wakefield acceleration studies in research directions already active at the ATF; laser wakefield acceleration (LWFA), where the longer laser wavelengths are expected to engender a proportional increase in the beam's charge while our linac will assure, for the first time, the opportunity to undertake detailed studies of seeding and staging of the LWFA; proton acceleration to the 100-200 MeV level, which is essential for medical applications; and others.
Sealing nuclear graphite with pyrolytic carbon

International Nuclear Information System (INIS)

Feng, Shanglei; Xu, Li; Li, Li; Bai, Shuo; Yang, Xinmei; Zhou, Xingtai

2013-01-01

Pyrolytic carbon (PyC) coatings were deposited on IG-110 nuclear graphite by thermal decomposition of methane at ∼1830 °C. The PyC coatings are anisotropic and airtight enough to protect IG-110 nuclear graphite against the permeation of molten fluoride salts and the diffusion of gases. The investigations indicate that the sealing nuclear graphite with PyC coating is a promising method for its application in Molten Salt Reactor (MSR)
Argentina: Disposal aspects of RA-1 research reactor decommissioning waste

Energy Technology Data Exchange (ETDEWEB)

Harriague, S; Barberis, C; Cinat, E; Grizutti, C; Scolari, H [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

2007-12-15

The objective of the project is to analyze disposal aspects of waste from total dismantling of Argentinean research reactors, starting with the oldest one, 48 years old RA-1. In order to estimate decommissioning waste, data was collected from files, area monitoring, measurements, sampling to measure activity and composition, operational history and tracing of operational incidents. Measurements were complemented with neutron activation calculations. Decommissioning waste for RA-1 is estimated to be 71.5 metric tons, most of it concrete (57 tons), the rest being steels, lead and reflector graphite (4.8 tons). Due to their low specific activities, no disposal problems are foreseen in the case of metals and concrete. Disposal of aluminium, steel, lead and concrete is analyzed. On the contrary, as the country has no experience in managing graphite radioactive waste, work was concentrated on that material. Stored (Wigner) energy may exist in RA-1 graphite reflectors irradiated at room temperature. Evaluation of stored energy by calorimetric methods is proposed, and its annealing by inductive heating; HEPA filters should be used to deal with gaseous activity emissions, mainly Cl-36 and C-14. Galvanic corrosion, dust explosion, ignition and oxidation can be addressed and should not become disposal problems. Care must be taken with graphite dust generation and disposal, due to wetting and flotation problems. Lessons learned from the project are presented, and the benefits of sharing international experience are stressed. (author)
Brookhaven National Laboratory source water assessment for drinking water supply wells

International Nuclear Information System (INIS)

Bennett, D.B.; Paquette, D.E.; Klaus, K.; Dorsch, W.R.

2000-01-01

The BNL water supply system meets all water quality standards and has sufficient pumping and storage capacity to meet current and anticipated future operational demands. Because BNL's water supply is drawn from the shallow Upper Glacial aquifer, BNL's source water is susceptible to contamination. The quality of the water supply is being protected through (1) a comprehensive program of engineered and operational controls of existing aquifer contamination and potential sources of new contamination, (2) groundwater monitoring, and (3) potable water treatment. The BNL Source Water Assessment found that the source water for BNL's Western Well Field (comprised of Supply Wells 4, 6, and 7) has relatively few threats of contamination and identified potential sources are already being carefully managed. The source water for BNL's Eastern Well Field (comprised of Supply Wells 10, 11, and 12) has a moderate number of threats to water quality, primarily from several existing volatile organic compound and tritium plumes. The g-2 Tritium Plume and portions of the Operable Unit III VOC plume fall within the delineated source water area for the Eastern Well Field. In addition, portions of the much slower migrating strontium-90 plumes associated with the Brookhaven Graphite Research Reactor, Waste Concentration Facility and Building 650 lie within the Eastern source water area. However, the rate of travel in the aquifer for strontium-90 is about one-twentieth of that for tritium and volatile organic compounds. The Laboratory has been carefully monitoring plume migration, and has made adjustments to water supply operations. Although a number of BNL's water supply wells were impacted by VOC contamination in the late 1980s, recent routine analysis of water samples from BNL's supply wells indicate that no drinking water standards have been reached or exceeded. The high quality of the water supply strongly indicates that the operational and engineered controls implemented over the past
Change in properties of graphite on stake of Obninsk NPP

International Nuclear Information System (INIS)

Virgul'ev, Yu.S.; Gundorov, V.V.; Kalyagina, I.P.; Belinskaya, N.T.; Dolgov, V.V.; Komissarov, O.V.; Stuzhnev, Yu.A.

1997-01-01

The results of testing the graphite from the AM-1 reactor masonry at the Obninsk NPP for its operation period are discussed. It is shown that the masonry graphite state after 42 years of the reactor operation remains satisfactory in the most cells inspected. Separate cells requiring a repair resulted from oxidation are characterized by strength decreased by several times. The laws of radiation changes in graphite properties are analyzed. The conclusion on possibility of the further masonry operation is drawn
New insights into canted spiro carbon interstitial in graphite

Science.gov (United States)

EL-Barbary, A. A.

2017-12-01

The self-interstitial carbon is the key to radiation damage in graphite moderator nuclear reactor, so an understanding of its behavior is essential for plant safety and maximized reactor lifetime. The density functional theory is applied on four different graphite unit cells, starting from of 64 carbon atoms up to 256 carbon atoms, using AIMPRO code to obtain the energetic, athermal and mechanical properties of carbon interstitial in graphite. This study presents first principles calculations of the energy of formation that prove its high barrier to athermal diffusion (1.1 eV) and the consequent large critical shear stress (39 eV-50 eV) necessary to shear graphite planes in its presence. Also, for the first time, the gamma surface of graphite in two dimensions is calculated and found to yield the critical shear stress for perfect graphite. Finally, in contrast to the extensive literature describing the interstitial of carbon in graphite as spiro interstitial, in this work the ground state of interstitial carbon is found to be canted spiro interstitial.
Condensation nuclear power plants with water-cooled graphite-moderated channel type reactors and advances in their development

International Nuclear Information System (INIS)

Boldyrev, V.M.; Mikhaj, V.I.

1985-01-01

Consideration is being given to results of technical and economical investigations of advisability of increasing unit power by elevating steam generating capacity as a result of inserting numerous of stereotype sectional structural elements of the reactor with similar thermodynamic parameters. It is concluded that construction of power units of condensation nuclear power plants with water-cooled graphite-moderated channel type reactors of 2400-3200 MWe and higher unit power capacity represents the real method for sharp growth of efficiency and labour productivity in power industry. It can also provide the required increase of the rate of putting electrogenerating powers into operation
Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor
Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation
Impact on breeding rate of different Molten Salt reactor core structures

International Nuclear Information System (INIS)

Wang Haiwei; Mei Longwei; Cai Xiangzhou; Chen Jingen; Guo Wei; Jiang Dazhen

2013-01-01

Background: Molten Salt Reactor (MSR) has several advantages over the other Generation IV reactor. Referred to the French CNRS research and compared to the fast reactor, super epithermal neutron spectrum reactor type is slightly lower and beading rate reaches 1.002. Purpose: The aim is to explore the best conversion zone layout scheme in the super epithermal neutron spectrum reactor. This study can make nuclear fuel as one way to solve the energy problems of mankind in future. Methods: Firstly, SCALE program is used for molten salt reactor graphite channel, molten salt core structure, control rods, graphite reflector and layer cladding structure. And the SMART modules are used to record the important actinides isotopes and their related reaction values of each reaction channel. Secondly, the thorium-uranium conversion rate is calculated. Finally, the better molten salt reactor core optimum layout scheme is studied comparing with various beading rates. Results: Breading zone layout scheme has an important influence on the breading rate of MSR. Central graphite channels in the core can get higher neutron flux irradiation. And more 233 Th can convert to 233 Pa, which then undergoes beta decay to become 233 U. The graphite in the breading zone gets much lower neutron flux irradiation, so the life span of this graphite can be much longer than that of others. Because neutron flux irradiation in the uranium molten salt graphite has nearly 10 times higher than the graphite in the breading zone, it has great impact on the thorium-uranium conversion rates. For the super epithermal neutron spectrum molten salt reactors, double salt design cannot get higher thorium-uranium conversion rates. The single molten salt can get the same thorium-uranium conversion rate, meanwhile it can greatly extend the life of graphite in the core. Conclusions: From the analysis of calculation results, Blanket breeding area in different locations in the core can change the breeding rates of thorium
Dismantling the nuclear research reactor Thetis

Energy Technology Data Exchange (ETDEWEB)

Michiels, P. [Belgoprocess, 2480 Dessel (Belgium)

2013-07-01

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation analyses. The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3 m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space. The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW. Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined. Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were
Gravity Effects on the Free Vibration of a Graphite Column

International Nuclear Information System (INIS)

Ki, Dong-Ok; Kim, Jong-Bum; Park, Keun-Bae; Lee, Won-Jae

2006-01-01

The gravity effects on the free vibration of a graphite column are studied. Graphite block is a key component of a HTGR (High Temperature Gas Cooled Reactor). The major core elements, such as the fuel blocks and neutron reflector blocks, of HTTR (High Temperature Test Reactor, Japan) and GT-MHR (Gas Turbine- Modular Helium Reactor, USA) consist of stacked hexagonal graphite blocks forming a group of columns. The vibration of the columns induced by earthquakes may lead to solid impacts between graphite blocks and structural integrity problems. The study of free vibration characteristics of the graphite block column is the first step in the core internal structure dynamic analysis. Gravity force bring a negative stiffness term to the transverse vibration analysis of heavy long column structures, and results in natural frequency reductions. Generally it is not considered in the not so tall structure cases, because the gravity term makes the analysis and design complicated. Therefore it is important to check whether the gravity effect is severe or not
Operational safety and reactor life improvements of Kyoto University Reactor

International Nuclear Information System (INIS)

Utsuro, M.; Fujita, Y.; Nishihara, H.

1990-01-01

Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

Recuperation of the energy released in the G-1, an air-cooled graphite reactor core; Recuperation de l'energie degagee dans G 1 pile a graphite refroidie a l'air

Energy Technology Data Exchange (ETDEWEB)

Chambadal, P [Electricite de France (EDF), 75 - Paris (France); Pascal, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1955-07-01

The CEA (in his five-year setting plan) has objective among others, the realization of the two first french reactors moderated with graphite. The construction of the G-1 reactor in Marcoule, first french plutonic core, is achieved so that it will diverge in the beginning of 1956 and reach its full power in the beginning of the second semester of the same year. In this report we will detail the specificities of the reactor and in particular its cooling and energy recuperation system. The G-1 reactor being essentially intended to allow the french technicians to study the behavior of an energy installation supply taking its heat in a nuclear source as early as possible. (M.B.) [French] Le Commissariat a l'Energie Atomique (dans le cadre du plan quinquennal) a entre autres objectifs, la realisation des deux premiers reacteurs francais moderes au graphite. La construction du reacteur G-1 a Marcoule, premiere pile plutonigene francaise, est realise afin qu'il puisse diverger au debut de 1956 et atteindre sa pleine puissance au debut du second semestre de la meme annee. Dans ce rapport nous detaillerons les specificites du reacteur et en particulier son systeme de refroidissement et de recuperation d'energie. Le reacteur G-1 etant essentielement destine a permettre aux techniciens francais d'etudier le plus tot possible le comportement d'une installation productrice d'energie empruntant sa chaleur a une source nucleaire. (M.B.)
Graphite behaviour in relation to the fuel element design

Energy Technology Data Exchange (ETDEWEB)

Everett, M. R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Manzel, R. [OECD High Temperature Reactor Project Dragon, Winfrith (United Kingdom); Blackstone, R. [Reactor Centrum, Petten (Netherlands); Delle, W. [Kernforschungsanlage, Juelich (Germany); Lungagnani, V. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands); Krefeld, R. [Joint Nuclear Research Centre, Euratom, Petten (Netherlands)

1969-09-01

The first designs of H.T.R. power reactors will probably use a Gilsocarbon based graphite for both the moderator/carrier blocks and for the fuel tubes. The initial physical properties and changes of dimensions, thermal expansion coefficient, Young*s modulus, and thermal conductivity on irradiation of Gilsocarbon graphites to typical reactor dwell-time fast neutron doses of 4 * 10²¹ cm ^-2 Ni dose Dido equivalent are given and values for the irradiation creep constant are presented. The influence of these property changes and those of chemical corrosion are considered briefly in relation to the present fuel element designs. The selection of an eventual less costly replacement graphite for Gilsocarbon graphite is discussed in terms of materials properties.
Impact of radiolysis and radiolytic corrosion on the release of {sup 13}C and {sup 37}Cl implanted into nuclear graphite: Consequences for the behaviour of {sup 14}C and {sup 36}Cl in gas cooled graphite moderated reactors

Energy Technology Data Exchange (ETDEWEB)

Moncoffre, N., E-mail: nathalie.moncoffre@ipnl.in2p3.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Toulhoat, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Silbermann, G. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); EDF – DPI - DIN – CIDEN, DIE - Division Environnement, Lyon (France); and others

2016-04-15

Graphite finds widespread use in many areas of nuclear technology based on its excellent moderator and reflector qualities as well as its strength and high temperature stability. Thus, it has been used as moderator or reflector in CO{sub 2} cooled nuclear reactors such as UNGG, MAGNOX, and AGR. However, neutron irradiation of graphite results in the production of {sup 14}C (dose determining radionuclide) and {sup 36}Cl (long lived radionuclide), these radionuclides being a key issue regarding the management of the irradiated waste. Whatever the management option (purification, storage, and geological disposal), a previous assessment of the radioactive inventory and the radionuclide's location and speciation has to be made. During reactor operation, the effects of radiolysis are likely to promote the radionuclide release especially at the gas/graphite interface. Radiolysis of the coolant is mainly initiated through γ irradiation as well as through Compton electrons in the graphite pores. Radiolysis can be simulated in laboratory using γ irradiation or ion irradiation. In this paper, {sup 13}C, {sup 37}Cl and {sup 14}N are implanted into virgin nuclear graphite in order to simulate respectively the presence of {sup 14}C, {sup 36}Cl and nitrogen, a {sup 14}C precursor. Different irradiation experiments were carried out using different irradiation devices on implanted graphite brought into contact with a gas simulating the coolant. The aim was to assess the effects of gas radiolysis and radiolytic corrosion induced by γ or He{sup 2+} irradiation at the gas/graphite interface in order to evaluate their role on the radionuclide release. Our results allow inferring that radiolytic corrosion has clearly promoted the release of {sup 14}C, {sup 36}Cl and {sup 14}N located at the graphite brick/gas interfaces and open pores.
IDENTIFICATION OF CHLOROMETHANE FORMATION PATHS DURING ELECTROCHEMICAL DECHLORINATION OF TCE USING GRAPHITE ELECTRODES

Science.gov (United States)

The purpose of this research is to investigate the formation of chloromethane during TCE dechlorination in a mixed electrochemical reactor using graphite electrodes. Chloromethane was the major chlorinated organic compound detected in previous dechlorination experiments. In order...
IDENTIFICATION OF CHLOROMETHANE FROMATION PATHS DURING ELECTROCHEMICAL DECHLORINATION OF TCE USING GRAPHITE ELECTRODES

Science.gov (United States)

The purpose of this research is to investigate the formation of chloromethane during TCE dechlorination in a mixed electrochemical reactor using graphite electrodes. Chloromethane was the major chlorinated organic compound detected in previous dechlorination experiments. In order...
Evaluation of the significance of inverse oxidation for HTGR graphites

International Nuclear Information System (INIS)

Lee, B.S.; Heiser, J. III; Sastre, C.

1983-01-01

The inverse oxidation refers to a higher mass loss inside the graphite than the outside. In 1980, Wichner et al reported this phenomenon (referred to as inside/out corrosion) observed in some H451 graphites, and offered an explanation that a catalyst (almost certainly Fe) is activated by the progressively increasing reducing conditions found in the graphite interior. Recently, Morgan and Thomas (1982) investigated this phenomenon is PGX graphites, and agreed on the existing mechanism to explain this pheomenon. They also called for attention to the possibility that this phenomenon may occur under HTGR (High Temperature Gas-Cooled Reactor) operating conditions. The purpose of this paper is to confirm the above mentioned explanation for this phenomenon and to evaluate the significance of this effect for HTGR graphites under realistic reactor conditions
Feasibility study of fusion breeding blanket concept employing graphite reflector

International Nuclear Information System (INIS)

Cho, Seungyon; Ahn, Mu-Young; Lee, Cheol Woo; Kim, Eung Seon; Park, Yi-Hyun; Lee, Youngmin; Lee, Dong Won

2015-01-01

Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.
Feasibility study of fusion breeding blanket concept employing graphite reflector

Energy Technology Data Exchange (ETDEWEB)

Cho, Seungyon, E-mail: sycho@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Ahn, Mu-Young [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Cheol Woo; Kim, Eung Seon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

Highlights: • A Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept adopts graphite as a reflector material by reducing the amount of beryllium multiplier. • Its feasibility was investigated in view point of the nuclear performance as well as material-related issues. • A nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket. • Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions. • In conclusion, the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition. - Abstract: To obtain high tritium breeding performance with limited blanket thickness, most of solid breeder blanket concepts employ a combination of lithium ceramic as a breeder and beryllium as a multiplier. In this case, considering that huge amount of beryllium are needed in fusion power plants, its handling difficulty and cost can be a major factor to be accounted for commercial use. Korea has proposed a Helium-Cooled Ceramic Reflector (HCCR) breeding blanket concept relevant to fusion power plants. Here, graphite is used as a reflector material by reducing the amount of beryllium multiplier. Its feasibility has been investigated in view point of the nuclear performance as well as material-related issues. In this paper, a nuclear analysis is performed under the fusion reactor condition to address the feasibility of graphite reflector in breeding blanket, considering tritium breeding capability and neutron shielding and activation aspects. Also, the chemical stability of the graphite is investigated considering the chemical stability under accident conditions, resulting in that the adaptation of graphite reflector in breeding blanket is intrinsically safe and plausible under fusion reactor condition.
Impermeable Graphite: A New Development for Embedding Radioactive Waste

International Nuclear Information System (INIS)

Fachinger, Johannes

2016-01-01

Irradiated graphite has to be handled as radioactive waste after the operational period of the reactor. However, the waste management of irradiated graphite e.g. from the Spanish Vandellos reactor shows, that waste management of even low contaminated graphite could be expensive and requires special retrieval, treatment and disposal technologies for safe long term storage as low or medium radioactive waste. FNAG has developed an impermeable graphite matrix (IGM) as nuclear waste embedding material. This IGM provides a long term stable enclosure of radioactive waste and can reuse irradiated graphite as feedstock material. Therefore, no additional disposal volume is required if e.g. concrete waste packages were replaced by IGM waste packages. The variability of IGM as embedding has been summarized in the following paper usable for metal scraps, ion exchange resins or debris from buildings. Furthermore the main physical, chemical and structural properties are described. (author)
Mechanical strength parameters of cast iron with lamellar graphite and their significance for the design of pressure-carrying reactor components

International Nuclear Information System (INIS)

Janakiev, N.

1977-01-01

The tensile strength of thick-walled components in cast iron with lamellar graphite is lower by about 50 to 65% than that stated in DIN 1691. The usable compressive strength of this material under uni-axial load is about twice as high as its tensile strength. The graphite lamellae are not bonded into the metallic matrix. The width of the gaps between the graphite lamellae and the matrix increases with increasing wall thickness of the casting. In stress calculations for design purposes it is advisable to rely only on the permissible tensile stresses. It is shown that cast iron can be used as structural material for shieldings but is unsuitable for thick-walled reactor components carrying compressive and tensile stresses because its mechanical strength parameters decrease rapidly with increasing wall thickness. (orig.) [de
Spatial flux instabilities, and their control in the graphite gas power reactors; Les instabilites spatiales du flux et leur controle dans les reacteurs de puissance graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Cailly, J L [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

Radial-azimuthal and axial spatial flux instabilities in graphite-gas reactors are studied by means of an analytical approach. Results are checked with those which are given by two dimensional (r, z and r, {theta}) kinetic models programmed for an IBM 7094 computer. At least, conclusions on the control of instabilities obtained from these models are reported. (author) [French] Les instabilites spatiales du flux dans les reacteurs graphite-gaz, radiales et azimutales d'une part, axiales d'autre part, sont etudiees au moyen d'une formulation analytique. Les resultats sont confrontes avec ceux que fournissent des modeles cinetiques a deux dimensions (r, z et r, {theta}) programmes sur IBM 7094. On donne enfin les conclusions relatives au controle de ces instabilites que ces modeles ont permis de degager. (auteur)
Status of neutron beam utilization at the Dalat nuclear research reactor

International Nuclear Information System (INIS)

Dien, Nguyen Nhi; Hai, Nguyen Canh

2003-01-01

The 500-kW Dalat nuclear research reactor was reconstructed from the USA-made 250-kW TRIGA Mark II reactor. After completion of renovation and upgrading, the reactor has been operating at its nominal power since 1984. The reactor is used mainly for radioisotope production, neutron activation analysis, neutron beam researches and reactor physics study. In the framework of the reconstruction and renovation project of the 1982-1984 period, the reactor core, the control and instrumentation system, the primary and secondary cooling systems, as well as other associated systems were newly designed and installed by the former Soviet Union. Some structures of the reactor, such as the reactor aluminum tank, the graphite reflector, the thermal column, horizontal beam tubes and the radiation concrete shielding have been remained from the previous TRIGA reactor. As a typical configuration of the TRIGA reactor, there are four neutron beam ports, including three radial and one tangential. Besides, there is a large thermal column. Until now only two-neutron beam ports and the thermal column have been utilized. Effective utilization of horizontal experimental channels is one of the important research objectives at the Dalat reactor. The research program on effective utilization of these experimental channels was conducted from 1984. For this purpose, investigations on physical characteristics of the reactor, neutron spectra and fluxes at these channels, safety conditions in their exploitation, etc. have been carried out. The neutron beams, however, have been used only since 1988. The filtered thermal neutron beams at the tangential channel have been extracted using a single crystal silicon filter and mainly used for prompt gamma neutron activation analysis (PGNAA), neutron radiography (NR) and transmission experiments (TE). The filtered quasi-monoenergetic keV neutron beams using neutron filters at the piercing channel have been used for nuclear data measurements, study on
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)
An Overview of Ageing Management and Refurbishment of Research Reactors at Trombay

Energy Technology Data Exchange (ETDEWEB)

Sharma, R. C.; Raina, V. K. [Bhabha Atomic Research Centre, Mumbai (India)

2014-08-15

Three nuclear research reactors have been in operation at Bhabha Atomic Research Centre, Mumbai, India. India has a rich experience of about 120 research reactor operating years including ageing management. A well structured programme is in force for plant life management, refurbishment and upgrading reactors in operation. Apsara, commissioned in August 1956, was the first research reactor. Apsara is a 1 MW{sub th} swimming pool type of reactor with a movable core loaded with enriched uranium fuel and immersed in demineralized light water pool, which serves as coolant, moderator and reflector besides providing radiation shielding. Apsara was shut down during May 2009 for partial decommissioning and upgrading to a 2 MW reactor with several safety upgrades, e.g. a LEU based reactor core with higher neutron flux, a new reactor building meeting seismic qualification criteria and two independent shutdown devices. Cirus, a 40 MW{sub th} tank type reactor utilizing heavy water as moderator, graphite as reflector, demineralized light water as primary coolant and natural uranium metal as fuel; has been in operation since 1960. After about three decade of operation, the availability factor started declining mainly due to outage of equipment exhibiting signs of ageing. After ageing studies and performance review, refurbishment requirements were identified. A programme for refurbishment was drawn that included safety upgrades like civil repairs to the emergency storage reservoir to meet seismic qualification criteria and a new iodine removal system for better efficiency. The reactor was shut down during 1997 for execution of this refurbishment programme. After completion of refurbishment, the reactor was brought back into operation during 2003. It has completed about seven years of safe operation after refurbishment with a significant increase in availability factor from 70% to about 90%. The reactor was permanently shut down during December 2010. The reactor core was unloaded
Brookhaven segment interconnect

International Nuclear Information System (INIS)

Morse, W.M.; Benenson, G.; Leipuner, L.B.

1983-01-01

We have performed a high energy physics experiment using a multisegment Brookhaven FASTBUS system. The system was composed of three crate segments and two cable segments. We discuss the segment interconnect module which permits communication between the various segments
New about research reactors

International Nuclear Information System (INIS)

Egorenkov, P.M.

2001-01-01

The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru
Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)
A new neutron reflectometer at Australia's HIFAR research reactor

International Nuclear Information System (INIS)

James, M.; Nelson, A.; Schulz, J.C.; Jones, M.J.; Studer, A.J.; Hathaway, P.

2005-01-01

A new neutron reflectometer has been built at Australia's 10MW HIFAR research reactor. The X172 reflectometer operates in a monochromatic, angular dispersive mode collecting reflectivity data as a function of angle. The incident neutron beam is monochromated by a pair of pyrolytic graphite crystals (λ=2.43A) before being collimated using a pair of motorised sintered B 4 C slits. Detection of the reflected neutron beam is via a 10-atmosphere, helium-3, linear position sensitive detector. Examples of data collected using the X172 reflectometer at air-solid and solid-liquid interfaces are given. Neutron reflectivity values as low as 10 -5 have been measured on this instrument
Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs
IEA-R1 Nuclear Research Reactor: 58 Years of Operating Experience and Utilization for Research, Teaching and Radioisotopes Production

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Jose Patricio Nahuel; Filho, Tufic Madi; Saxena, Rajendra; Filho, Walter Ricci [Nuclear and Energy Research Institute, IPEN-CNEN/SP, Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242 Cid Universitaria CEP: 05508-000- Sao Paulo-SP (Brazil)

2015-07-01

IEA-R1 research reactor at the Instituto de Pesquisas Energeticas e Nucleares (Nuclear and Energy Research Institute) IPEN, Sao Paulo, Brazil is the largest power research reactor in Brazil, with a maximum power rating of 5 MWth. It is being used for basic and applied research in the nuclear and neutron related sciences, for the production of radioisotopes for medical and industrial applications, and for providing services of neutron activation analysis, real time neutron radiography, and neutron transmutation doping of silicon. IEA-R1 is a swimming pool reactor, with light water as the coolant and moderator, and graphite and beryllium as reflectors. The reactor was commissioned on September 16, 1957 and achieved its first criticality. It is currently operating at 4.5 MWth with a 60-hour cycle per week. In the early sixties, IPEN produced {sup 131}I, {sup 32}P, {sup 198}Au, {sup 24}Na, {sup 35}S, {sup 51}Cr and labeled compounds for medical use. During the past several years, a concerted effort has been made in order to upgrade the reactor power to 5 MWth through refurbishment and modernization programs. One of the reasons for this decision was to produce {sup 99}Mo at IPEN. The reactor cycle will be gradually increased to 120 hours per week continuous operation. It is anticipated that these programs will assure the safe and sustainable operation of the IEA-R1 reactor for several more years, to produce important primary radioisotopes {sup 99}Mo, {sup 125}I, {sup 131}I, {sup 153}Sm and {sup 192}Ir. Currently, all aspects of dealing with fuel element fabrication, fuel transportation, isotope processing, and spent fuel storage are handled by IPEN at the site. The reactor modernization program is slated for completion by 2015. This paper describes 58 years of operating experience and utilization of the IEA-R1 research reactor for research, teaching and radioisotopes production. (authors)

Research reactor support

International Nuclear Information System (INIS)

2005-01-01

Research reactors (RRs) have been used in a wide range of applications including nuclear power development, basic physics research, education and training, medical isotope production, geology, industry and other fields. However, many research reactors are fuelled with High Enriched Uranium (HEU), are underutilized and aging, and have significant quantities of spent fuel. HEU inventories (fresh and spent) pose security risks Unavailability of a high-density-reprocessable fuel hinders conversion and limits back-end options and represents a survival dilemma for many RRs. Improvement of interim spent fuel storage is required at some RRs. Many RRs are under-utilized and/or inadequately funded and need to find users for their services, or permanently shut down and eventually decommission. Reluctance to decommission affect both cost and safety (loss of experienced staff ) and many shut down but not decommissioned RR with fresh and/or spent fuel at the sites invoke serious concern. The IAEA's research reactor support helps to ensure that research reactors can be operated efficiently with fuels and targets of lower proliferation and security concern and that operators have appropriate technology and options to manage RR fuel cycle issues, especially on long term interim storage of spent research reactor fuel. Availability of a high-density-reprocessable fuel would expand and improve back end options. The International Atomic Energy Agency provides assistance to Member States to convert research reactors from High Enriched Uranium fuel and targets (for medical isotope production) to qualified Low Enriched Uranium fuel and targets while maintaining reactor performance levels. The assistance includes provision of handbooks and training in the performance of core conversion studies, advice for the procurement of LEU fuel, and expert services for LEU fuel acceptance. The IAEA further provides technical and administrative support for countries considering repatriation of its
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)
Effect of a Central Graphite Column on a Pebble Flow in a Pebble Bed Core

International Nuclear Information System (INIS)

In, W. K.; Lee, W. J.; Chang, J. H.

2006-01-01

A pebble bed reactor(PBR) uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. The pebble bed core is configured as cylindrical or annular depending on the reactor power. It is well known that an annular core can increase a cores' thermal power. The annular inner core zone is typically filled with movable graphite balls or a fixed graphite column. The first problem with this conventional annular core is that it is difficult to maintain a boundary between the central graphite ball zone and the outer fuel zone. The second problem is that it is expensive to replace the central fixed graphite column after several tens of years of reactor operation. In order to resolve these problems, a PBR with a central graphite column in a low core is invented. This paper presents the effect of the central graphite column on a pebble flow by using the computational fluid dynamics(CFD) code, CFX-10
Design and fabrication report on capsule (11M 19K for out of pile test) for irradiation testing of research reactor materials at HANARO

Energy Technology Data Exchange (ETDEWEB)

Kim, B.G.; Yang, S.W.; Park, S.J.; Shim, K.T.; Choo, K.N.; Oh, J.M.; Lee, B.C.; Choi, M.H.; Kim, D.J.; Kim, J.M.; Kang, S.H.; Chun, Y.B.; Kim, T.K.; Jeong, Y.H.

2012-05-15

As a part of the research reactor development project with a plate type fuel, the irradiation tests of graphite (Gr), beryllium (Be), and zircaloy 4 materials using the capsule have been investigating to obtain the mechanical characteristics such as an irradiation growth, hardness, swelling and tensile strength at the temperature below 100 .deg. C and the 30 MW reactor power. Then, A capsule to be able to irradiate materials(graphite, Be, zircaloy 4) under 100 .deg. C at the HANARO was designed and fabricated. After performing out of pile testing in single channel test loop by using the capsule, the final design of the capsules to be irradiated in CT and IR2 test hole of HANARO was approved, and 2 sets of capsule were fabricated. These capsules will be loaded in CT and IR2 test hole of HANARO, and be started the irradiation from the end of June, 2012. After performing the irradiation testing of 2 sets of capsule, PIE (Post Irradiation Examination) on irradiated specimens (Gr, Be, and zircaloy 4) will be carry out in IMEF (Irradiated Material Examination Facility). So, the irradiation testing will be contributed to obtain the characteristic data induced neutron irradiation on Gr, Be, and zircaloy 4. And then, it is convinced that these data will be also contributed to obtain the license for JRTR (Jordan Research and Training Reactor) and new research reactor in Korea, and export research reactors.
Department of Energy’s ARM Climate Research Facility External Data Center Operations Plan Located At Brookhaven National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Cialella, A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Gregory, L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Lazar, K. [Brookhaven National Lab. (BNL), Upton, NY (United States); Liang, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ma, L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Tilp, A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Wagener, R. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2015-05-01

The External Data Center (XDC) Operations Plan describes the activities performed to manage the XDC, located at Brookhaven National Laboratory (BNL), for the Department of Energy’s Atmospheric Radiation Measurement (ARM) Climate Research Facility. It includes all ARM infrastructure activities performed by the Data Management and Software Engineering Group (DMSE) at BNL. This plan establishes a baseline of expectation within the ARM Operations Management for the group managing the XDC.
The Amtex DAMA Project: The Brookhaven contribution

Energy Technology Data Exchange (ETDEWEB)

Peskin, A.M.

1995-01-01

The Amtex Partnership organized in 1993 as a Technology Transfer Collaboration among members of the integrated textile industry, the DOE National Laboratories, a number of universities, and several research/education/technology transfer organizations (RETTs). Under the Amtex umbrella organization, a number of technology areas were defined and individual projects were launched addressing various aspects of improving the health and competitiveness of the American textile industry. The first and, to date, the largest of these has been the computer-based Demand Activated Manufacturing Architecture (DAMA) project. Brookhaven National Laboratory became involved in DAMA beginning in March of 1993 and remained an active participant through January of 1995. It was staffed almost exclusively with personnel of the Computing and Communications Division. This document summarizes the activities and accomplishments of the Brookhaven team in working with the larger collaboration. Detailed information about the Amtex Partnership, the DAMA Project, and specific BNL contributions are documented elsewhere.
Source Term Analysis of the Irradiated Graphite in the Core of HTR-10

Directory of Open Access Journals (Sweden)

Xuegang Liu

2017-01-01

Full Text Available The high temperature gas-cooled reactor (HTGR has potential utilization due to its featured characteristics such as inherent safety and wide diversity of utilization. One distinct difference between HTGR and traditional pressurized water reactor (PWR is the large inventory of graphite in the core acting as reflector, moderator, or structure materials. Some radionuclides will be generated in graphite during the period of irradiation, which play significant roles in reactor safety, environmental release, waste disposal, and so forth. Based on the actual operation of the 10 MW pebble bed high temperature gas-cooled reactor (HTR-10 in Tsinghua University, China, an experimental study on source term analysis of the irradiated graphite has been done. An irradiated graphite sphere was randomly collected from the core of HTR-10 as sample in this study. This paper focuses on the analytical procedure and the establishment of the analytical methodology, including the sample collection, graphite sample preparation, and analytical parameters. The results reveal that the Co-60, Cs-137, Eu-152, and Eu-154 are the major γ contributors, while H-3 and C-14 are the dominating β emitting nuclides in postirradiation graphite material of HTR-10. The distribution profiles of the above four nuclides are also presented.
Intercomparison of graphite irradiations

Energy Technology Data Exchange (ETDEWEB)

Hering, H; Perio, P; Seguin, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

While fast neutrons only are effective in damaging graphite, results of irradiations are more or less universally expressed in terms of thermal neutron fluxes. This paper attempts to correlate irradiations made in different reactors, i.e., in fluxes of different spectral compositions. Those attempts are based on comparison of 1) bulk length change and volume expansion, and 2) crystalline properties (e.g., lattice parameter C, magnetic susceptibility, stored energy, etc.). The methods used by various authors for determining the lattice constants of irradiated graphite are discussed. (author)
Irradiation damage in graphite due to fast neutrons in fission and fusion systems

International Nuclear Information System (INIS)

2000-09-01

Gas cooled reactors have been in operation for the production of electricity for over forty years, encompassing a total of 56 units operated in seven countries. The predominant experience has been with carbon dioxide cooled reactors (52 units), with the majority operated in the United Kingdom. In addition, four prototype helium cooled power plants were operated in the United States and Germany. The United Kingdom has no plans for further construction of carbon dioxide units, and the last helium cooled unit was shutdown in 1990. However, there has been an increasing interest in modular helium cooled reactors during the 1990s as a possible future nuclear option. Graphite is a primary material for the construction of gas cooled reactor cores, serving as a low absorption neutron moderator and providing a high temperature, high strength structure. Commercial gas cooled reactor cores (both carbon dioxide cooled and helium cooled) utilise large quantities of graphite. The structural behaviour of graphite (strength, dimensional stability, susceptibility to cracking, etc.) is a complex function of the source material, manufacturing process, chemical environment, and temperature and irradiation history. A large body of data on graphite structural performance has accumulated from operation of commercial gas cooled reactors, beginning in the 1950s and continuing to the present. The IAEA is supporting a project to collect graphite data and archive it in a retrievable form as an International Database on Irradiated Nuclear Graphite Properties, with limited general access and more detailed access by participating Member States. Because of the large size of the database, the complexity of the phenomena and the number of variables involved, a general understanding of graphite behaviour is essential to the understanding and use of the data
Abrasion behavior of graphite pebble in lifting pipe of pebble-bed HTR

Energy Technology Data Exchange (ETDEWEB)

Shen, Ke; Su, Jiageng [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Zhou, Hongbo [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Chinergy Co., LTD., Beijing 100193 (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Yu, Suyun, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 10084 (China)

2015-11-15

Highlights: • Quantitative determination of abrasion rate of graphite pebbles in different lifting velocities. • Abrasion behavior of graphite pebble in helium, air and nitrogen. • In helium, intensive collisions caused by oscillatory motion result in more graphite dust production. - Abstract: A pebble-bed high-temperature gas-cooled reactor (pebble-bed HTR) uses a helium coolant, graphite core structure, and spherical fuel elements. The pebble-bed design enables on-line refueling, avoiding refueling shutdowns. During circulation process, the pebbles are lifted pneumatically via a stainless steel lifting pipe and reinserted into the reactor. Inevitably, the movement of the fuel elements as they recirculate in the reactor produces graphite dust. Mechanical wear is the primary source of graphite dust production. Specifically, the sources are mechanisms of pebble–pebble contact, pebble–wall (structural graphite) contact, and fuel handling (pebble–metal abrasion). The key contribution to graphite dust production is from the fuel handling system, particularly from the lifting pipe. During pneumatic lift, graphite pebbles undergo multiple collisions with the stainless steel lifting pipe, thereby causing abrasion of the graphite pebbles and producing graphite dust. The present work explored the abrasion behavior of graphite pebble in the lifting pipe by measuring the abrasion rate at different lifting velocities. The abrasion rate of the graphite pebble in helium was found much higher than those in air and nitrogen. This gas environment effect could be explained by either tribology behavior or dynamic behavior. Friction testing excluded the possibility of tribology reason. The dynamic behavior of the graphite pebble was captured by analysis of the audio waveforms during pneumatic lift. The analysis results revealed unique dynamic behavior of the graphite pebble in helium. Oscillation and consequently intensive collisions occur during pneumatic lift, causing
BROOKHAVEN NATIONAL LABORATORY INSTITUTIONAL PLAN FY2003-2007.

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-06-10

This document presents the vision for Brookhaven National Laboratory (BNL) for the next five years, and a roadmap for implementing that vision. Brookhaven is a multidisciplinary science-based laboratory operated for the U.S. Department of Energy (DOE), supported primarily by programs sponsored by the DOE's Office of Science. As the third-largest funding agency for science in the U.S., one of the DOE's goals is ''to advance basic research and the instruments of science that are the foundations for DOE's applied missions, a base for U.S. technology innovation, and a source of remarkable insights into our physical and biological world, and the nature of matter and energy'' (DOE Office of Science Strategic Plan, 2000 http://www.osti.gov/portfolio/science.htm). BNL shapes its vision according to this plan.
Advanced High-Temperature Reactor for Production of Electricity and Hydrogen: Molten-Salt-Coolant, Graphite-Coated-Particle-Fuel

International Nuclear Information System (INIS)

Forsberg, C.W.

2002-01-01

The objective of the Advanced High-Temperature Reactor (AHTR) is to provide the very high temperatures necessary to enable low-cost (1) efficient thermochemical production of hydrogen and (2) efficient production of electricity. The proposed AHTR uses coated-particle graphite fuel similar to the fuel used in modular high-temperature gas-cooled reactors (MHTGRs), such as the General Atomics gas turbine-modular helium reactor (GT-MHR). However, unlike the MHTGRs, the AHTR uses a molten salt coolant with a pool configuration, similar to that of the PRISM liquid metal reactor. A multi-reheat helium Brayton (gas-turbine) cycle, with efficiencies >50%, is used to produce electricity. This approach (1) minimizes requirements for new technology development and (2) results in an advanced reactor concept that operates at essentially ambient pressures and at very high temperatures. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for (1) exceptionally robust safety (including passive decay-heat removal) and (2) allows scaling to large reactor sizes [∼1000 Mw(e)] with passive safety systems to provide the potential for improved economics
Comparison of fast neutron spectra in graphite and FLINA salt inserted in well-defined core assembled in LR-0 reactor

International Nuclear Information System (INIS)

Košťál, Michal; Veškrna, Martin; Cvachovec, František; Jánský, Bohumil; Novák, Evžen; Rypar, Vojtěch; Milčák, Ján; Losa, Evžen; Mravec, Filip; Matěj, Zdeněk; Rejchrt, Jiří; Forget, Benoit; Harper, Sterling

2015-01-01

Highlights: • Neutron spectra measured in graphite and LiF + NaF. • Comparison of calculated and measured neutron spectra. • Effect of 19F on variation between various library calculated spectra. - Abstract: The present paper aims to compare the calculated and measured spectra after insertion of candidate materials for the Molten salt reactor/Fluoride cooled high temperature reactor system concept into the LR-0 reactor. The calculation is realized with MCNP6 code using ENDF/B-VII.0, JEFF-3.1, JENDL-3.3, JENDL-4, ROSFOND-2010 and CENDL-3.1 nuclear data libraries. Additionally, comparisons between the slowing down power of each media were performed. The slowing down properties are important parameters affecting the thickness of moderator media in a reactor
Obtention of nuclear grade graphite

International Nuclear Information System (INIS)

Ferreira, M.L.

1984-01-01

The impurity level of natural graphite found in some of the most important mines of the State of Minas Gerais - Brasil is determined. It is also concerned with the development and use of natural graphite in nuclear reactors. Standard methods for chemical and instrumentsal analysis such as Spectrografic Determination by Emission, Spectrografic Determination by X-Rays, Spectrografic Determination by Atomic Asorption, Photometric Determination, and also chemical and physical methods for separation of impurities as well standard method for Estimating the Thermal Neutron Absorption Cross Section of graphite were employed. Some aditionals methods of purification to the ordinary treatment such as the use of metanol and halogens are also described. (Author) [pt
Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

Energy Technology Data Exchange (ETDEWEB)

Zhong, Yajuan, E-mail: yajuan.zhong@gmail.com [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Zhang, Junpeng [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China); Lin, Jun, E-mail: linjun@sinap.ac.cn [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Xu, Liujun [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong [Center for Thorium Molten Salt Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Guo, Quangui [CAS Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences, Taiyuan 030001 (China)

2017-07-15

Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10{sup −6} K{sup −1} (α{sub ∥}) and 6.15 × 10{sup −6} K{sup −1} (α{sub ⊥}) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.
Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor

International Nuclear Information System (INIS)

Zhong, Yajuan; Zhang, Junpeng; Lin, Jun; Xu, Liujun; Zhang, Feng; Xu, Hongxia; Chen, Yu; Jiang, Haitao; Li, Ziwei; Zhu, Zhiyong; Guo, Quangui

2017-01-01

Mesocarbon microbeads (MCMB) and quasi-isostatic pressing method were used to prepare MCMB based graphite (MG) for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor (MSR). Characteristics of mercury infiltration and molten salt infiltration in MG were investigated and compared with A3-3 (graphite for spherical fuel element in high temperature gas cooled reactor) to identify the infiltration behaviors. The results indicated that MG had a low porosity about 14%, and an average pore diameter of 96 nm. Fluoride salt occupation of A3-3 (average pore diameter was 760 nm) was 10 wt% under 6.5 atm, whereas salt gain did not infiltrate in MG even up to 6.5 atm. It demonstrated that MG could inhibit the infiltration of liquid fluoride salt effectively. Coefficient of thermal expansion (CTE) of MG lies in 6.01 × 10 −6 K −1 (α ∥ ) and 6.15 × 10 −6 K −1 (α ⊥ ) at the temperature range of 25–700 °C. The anisotropy factor of MG calculated by CTE maintained below 1.02, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic property of MG is beneficial for the integrity and safety of the graphite used in the spherical fuel element for a MSR.
Effects of Oxidation on Oxidation-Resistant Graphite

Energy Technology Data Exchange (ETDEWEB)

Windes, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Rebecca [Idaho National Lab. (INL), Idaho Falls, ID (United States); Carroll, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-05-01

The Advanced Reactor Technology (ART) Graphite Research and Development Program is investigating doped nuclear graphite grades that exhibit oxidation resistance through the formation of protective oxides on the surface of the graphite material. In the unlikely event of an oxygen ingress accident, graphite components within the VHTR core region are anticipated to oxidize so long as the oxygen continues to enter the hot core region and the core temperatures remain above 400°C. For the most serious air-ingress accident which persists over several hours or days the continued oxidation can result in significant structural damage to the core. Reducing the oxidation rate of the graphite core material during any air-ingress accident would mitigate the structural effects and keep the core intact. Previous air oxidation testing of nuclear-grade graphite doped with varying levels of boron-carbide (B4C) at a nominal 739°C was conducted for a limited number of doped specimens demonstrating a dramatic reduction in oxidation rate for the boronated graphite grade. This report summarizes the conclusions from this small scoping study by determining the effects of oxidation on the mechanical strength resulting from oxidation of boronated and unboronated graphite to a 10% mass loss level. While the B4C additive did reduce mechanical strength loss during oxidation, adding B4C dopants to a level of 3.5% or more reduced the as-fabricated compressive strength nearly 50%. This effectively minimized any benefits realized from the protective film formed on the boronated grades. Future work to infuse different graphite grades with silicon- and boron-doped material as a post-machining conditioning step for nuclear components is discussed as a potential solution for these challenges in this report.
Research on the reactor physics using the Kyoto University Critical Assembly (KUCA)

International Nuclear Information System (INIS)

1986-10-01

The Kyoto University Critical Assembly [KUCA] is a multi-core type critical assembly established in 1974, as a facility for the joint use study by researchers of all universities in Japan. Thereafter, many reactor physics experiments have been carried out using three cores (A-, B-, and C-cores) in the KUCA. In the A- and B-cores, solid moderator such as polyethylene or graphite is used, whereas light-water is utilized as moderator in the C-core. The A-core has been employed mainly in connection with the Cockcroft-Walton type accelerator installed in the KUCA, to measure (1) the subcriticality by the pulsed neutron technique for the critical safety research and (2) the neutron spectrum by the time-of-flight technique. Recently, a basic study on the tight lattice core has also launched using the A-core. The B-core has been employed for the research on the thorium fuel cycle ever since. The C-core has been employed (1) for the basic studies on the nuclear characteristics of light-water moderated high-flux research reactors, including coupled-cores, and (2) for a research related to reducing enrichment of uranium fuel used in research reactors. The C-core is being utilized in the reactor laboratory course experiment for students of ten universities in Japan. The data base of the KUCA critical experiments is generated so far on the basis of approximately 350 experimental reports accumulated in the KUCA. Besides, the assessed KUCA code system has been established through analyses on the various KUCA experiments. In addition to the KUCA itself, both of them are provided for the joint use study by researchers of all universities in Japan. (author)
Nuclear Reactor Sharing Program

International Nuclear Information System (INIS)

1994-01-01

The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system
Brookhaven National Laboratory Institutional Plan FY2001--FY2005

Energy Technology Data Exchange (ETDEWEB)

Davis, S.

2000-10-01

Brookhaven National Laboratory is a multidisciplinary laboratory in the Department of Energy National Laboratory system and plays a lead role in the DOE Science and Technology mission. The Laboratory also contributes to the DOE missions in Energy Resources, Environmental Quality, and National Security. Brookhaven strives for excellence in its science research and in facility operations and manages its activities with particular sensitivity to environmental and community issues. The Laboratory's programs are aligned continuously with the goals and objectives of the DOE through an Integrated Planning Process. This Institutional Plan summarizes the portfolio of research and capabilities that will assure success in the Laboratory's mission in the future. It also sets forth BNL strategies for our programs and for management of the Laboratory. The Department of Energy national laboratory system provides extensive capabilities in both world class research expertise and unique facilities that cannot exist without federal support. Through these national resources, which are available to researchers from industry, universities, other government agencies and other nations, the Department advances the energy, environmental, economic and national security well being of the US, provides for the international advancement of science, and educates future scientists and engineers.

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

International Nuclear Information System (INIS)

2016-02-01

The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)
Immobilization of carbon-14 from reactor graphite waste by use of self-sustaining reaction in the C-Al-TiO2 system

International Nuclear Information System (INIS)

Karlina, O.K.; Klimov, V.L.; Ojovan, M.I.; Pavlova, G.Yu.; Dmitriev, S.A.; Yurchenko, A.Yu.

2005-01-01

As a result of long-term neutron irradiation, the long-lived 14 C is produced in the reactor graphite. The exothermic self-sustaining reaction 3C(graphite) + 4Al + 3TiO 2 = 3TiC + 2Al 2 O 3 was proposed for processing of such waste. In doing so, the carbon, including the 14 C, is chemically bound in the stable TiC. The reaction products in the C-Al-TiO 2 system were investigated both by thermodynamic simulation and experimentally in the course of this work
Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community
Investigation on structural integrity of graphite component during high temperature 950degC continuous operation of HTTR

International Nuclear Information System (INIS)

Sumita, Junya; Shimazaki, Yosuke; Shibata, Taiju

2014-01-01

Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950degC continuous operation, a high temperature continuous operation with reactor outlet temperature of 950degC for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950degC continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000degC. In addition, the programs of surveillance test and ISI using TV camera were introduced. (author)
Development and engineering plan for graphite spent fuels conditioning program

International Nuclear Information System (INIS)

Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

1993-09-01

Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated
Database activities at Brookhaven National Laboratory

International Nuclear Information System (INIS)

Trahern, C.G.

1995-01-01

Brookhaven National Laboratory is a multi-disciplinary lab in the DOE system of research laboratories. Database activities are correspondingly diverse within the restrictions imposed by the dominant relational database paradigm. The authors discuss related activities and tools used in RHIC and in the other major projects at BNL. The others are the Protein Data Bank being maintained by the Chemistry department, and a Geographical Information System (GIS)--a Superfund sponsored environmental monitoring project under development in the Office of Environmental Restoration
Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and
Critical experiments on enriched uranium graphite moderated cores

International Nuclear Information System (INIS)

Kaneko, Yoshihiko; Akino, Fujiyoshi; Kitadate, Kenji; Kurokawa, Ryosuke

1978-07-01

A variety of 20 % enriched uranium loaded and graphite-moderated cores consisting of the different lattice cells in a wide range of the carbon to uranium atomic ratio have been built at Semi-Homogeneous Critical Experimental Assembly (SHE) to perform the critical experiments systematically. In the present report, the experimental results for homogeneously or heterogeneously fuel loaded cores and for simulation core of the experimental reactor for a multi-purpose high temperature reactor are filed so as to be utilized for evaluating the accuracy of core design calculation for the experimental reactor. The filed experimental data are composed of critical masses of uranium, kinetic parameters, reactivity worths of the experimental control rods and power distributions in the cores with those rods. Theoretical analyses are made for the experimental data by adopting a simple ''homogenized cylindrical core model'' using the nuclear data of ENDF/B-III, which treats the neutron behaviour after smearing the lattice cell structure. It is made clear from a comparison between the measurement and the calculation that the group constants and fundamental methods of calculations, based on this theoretical model, are valid for the homogeneously fuel loaded cores, but not for both of the heterogeneously fuel loaded cores and the core for simulation of the experimental reactor. Then, it is pointed out that consideration to semi-homogeneous property of the lattice cells for reactor neutrons is essential for high temperature graphite-moderated reactors using dispersion fuel elements of graphite and uranium. (author)
{sup 36}Cl and {sup 14}C behaviour in UNGG graphite during leaching experiments

Energy Technology Data Exchange (ETDEWEB)

Pichon, C.; Guy, C.; Comte, J. [Commissariat a l' Energie Atomique - C.E.A., Laboratoire d' Analyses Radiochimiques et Chimiques (L.A.R.C.) 13108 Saint Paul lez Durance (France)

2008-07-01

Graphite has been used as a moderator in Natural Uranium Graphite Gas reactors. Among the radionuclides, the long-lived activation product {sup 36}Cl and {sup 14}C, which are abundant in graphite after irradiation can be the main contributors to the dose during disposal. This paper deals with the first results obtained on irradiated graphite from French G2 reactor. Both leaching and diffusion experiments have been performed in order to understand and quantify the radionuclides behaviour. Chlorine leaching seems to be controlled by diffusion transport through graphite matrix. On the contrary {sup 14}C leaching is very low, probably because after irradiation, the remaining {sup 14}C was produced from {sup 13}C activation in the crystalline structure of graphite. (authors)
Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project

International Nuclear Information System (INIS)

2016-05-01

Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants
Some economic aspects of natural uranium graphite gas reactor types. Present status and trends of costs in France; Quelques aspects economiques de la filiere uranium naturel - Graphite - gaz. Etat actuel et tendance des couts en France

Energy Technology Data Exchange (ETDEWEB)

Gaussens, J; Tanguy, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Leo, B [Electricite de France (EDF), 75 - Paris (France)

1964-07-01

The first part of this report defines the economic advantages of natural uranium fuels, which are as follows: the restricted number and relatively simple fabrication processes of the fuel elements, the low cost per kWh of the finished product and the reasonable capital investments involved in this type of fuel cycle as compared to that of enriched uranium. All these factors combine to reduce the arbitrary nature of cost estimates, which is particularly marked in the case of enriched uranium due to the complexity of its cycle and the uncertainties of plutonium prices). Finally, the wide availability of yellowcake, as opposed to the present day virtual monopoly of isotope separation, and the low cost of natural uranium stockpiling, offer appreciable guarantees in the way of security of supply and economic and political independence as compared with the use of enriched uranium. As far as overall capital investments are concerned, it is shown that, although graphite-gas reactor costs are higher than those of light water reactors in certain capacity ranges, the situation becomes far less clear when we start taking into account, in the interest of national independence, the cost of nuclear fuel production equipment in the case of each of these types of reactor. Finally, the marginal cost of the power capacity of a graphite-gas reactor is low and its technological limitations have receded (owing particularly to the use of prestressed concrete). It is a well known fact that the trend is now towards larger power station units, which means that the rentability of natural uranium graphite reactors as compared to other types of reactors will become more and more pronounced. The second section aims at presenting a realistic short and medium term view of the fuel, running, and investment costs of French natural uranium graphite gas, reactors. Finally, the economic goals which this type of reactor can reach in the very near future are given. It is thus shown that considerable
Application of 3D coupled code ATHLET-QUABOX/CUBBOX for RBMK-1000 transients after graphite block modernization

Energy Technology Data Exchange (ETDEWEB)

Samokhin, Aleksei [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS), Moscow (Russian Federation); Zilly, Matias [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

2016-11-15

This work describes the application and the results of transient calculations for the RBMK-1000 with the coupled code system ATHLET 2.2A-QUABOX/CUBBOX which was developed in GRS. Within these studies the planned modernization of the graphite blocks of the RBMK-1000 reactor is taken into account. During the long-term operation of the uranium-graphite reactors RBMK-1000, a change of physical and mechanical properties of the reactor graphite blocks is observed due to the impact of radiation and temperature effects. These have led to a deformation of the reactor graphite columns and, as a result, a deformation of the control and protection system (CPS) and of fuel channels. Potentially, this deformation can lead to problems affecting the smooth movement of the control rods in the CPS channels and problems during the loading and unloading of fuel assemblies. The present paper analyzes two reactivity insertion transients, each taking into account three graphite removal scenarios. The presented work is directly connected with the modernization program of the RBMK- 1000 reactors and has an important contribution to the assessment of the safety-relevant parameters after the modification of the core graphite blocks.
Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs) Volume 5: Graphite PIRTs

International Nuclear Information System (INIS)

Burchell, Timothy D.; Bratton, Rob; Marsden, Barry; Srinivasan, Makuteswara; Penfield, Scott; Mitchell, Mark; Windes, Will

2008-01-01

Here we report the outcome of the application of the Nuclear Regulatory Commission (NRC) Phenomena Identification and Ranking Table (PIRT) process to the issue of nuclear-grade graphite for the moderator and structural components of a next generation nuclear plant (NGNP), considering both routine (normal operation) and postulated accident conditions for the NGNP. The NGNP is assumed to be a modular high-temperature gas-cooled reactor (HTGR), either a gas-turbine modular helium reactor (GTMHR) version (a prismatic-core modular reactor (PMR)] or a pebble-bed modular reactor (PBMR) version (a pebble bed reactor (PBR)] design, with either a direct- or indirect-cycle gas turbine (Brayton cycle) system for electric power production, and an indirect-cycle component for hydrogen production. NGNP design options with a high-pressure steam generator (Rankine cycle) in the primary loop are not considered in this PIRT. This graphite PIRT was conducted in parallel with four other NRC PIRT activities, taking advantage of the relationships and overlaps in subject matter. The graphite PIRT panel identified numerous phenomena, five of which were ranked high importance-low knowledge. A further nine were ranked with high importance and medium knowledge rank. Two phenomena were ranked with medium importance and low knowledge, and a further 14 were ranked medium importance and medium knowledge rank. The last 12 phenomena were ranked with low importance and high knowledge rank (or similar combinations suggesting they have low priority). The ranking/scoring rationale for the reported graphite phenomena is discussed. Much has been learned about the behavior of graphite in reactor environments in the 60-plus years since the first graphite rectors went into service. The extensive list of references in the Bibliography is plainly testament to this fact. Our current knowledge base is well developed. Although data are lacking for the specific grades being considered for Generation IV (Gen IV
Defuelling of the UTR-300 research reactor

International Nuclear Information System (INIS)

Scott, R.D.; Banford, H.M.; East, B.W.

1997-01-01

The UTR-300 reactor at the Scottish Universities Research and Reactor Centre was based on the original Argonaut design with two aluminium core tanks set in a graphite reflector each containing six fuel elements cooled and moderated by water flowing up through the tanks in a closed primary circuit. The fuel plates in the original 13-plate elements were uranium oxide-aluminium with a 22g loading of 90% 235 U. After 7 years of operation at 100 kW (10 kW average), the maximum power was increased to 300 kW (30 kW average) and, in order to maintain the operational excess reactivity, it was necessary to add another plate to each element progressively over the years until they all contained 14 plates. These extra plates were uranium metal-aluminium with 24.5 g of 90% 235 U. No further modification of the elements was possible and so, with reactivity steadily decreasing, and for a variety of other reasons, a decision was taken to cease operation in September 1995. This paper describes the procedures whereby the fuel was unloaded from the core into a UNIFETCH flask equipped with a specially designed rotating gamma ray shield and then transported on two separate loads to Dounreay for reprocessing. (author)
Preliminary experiment design of graphite dust emission measurement under accident conditions for HTGR

Energy Technology Data Exchange (ETDEWEB)

Peng, Wei, E-mail: pengwei@tsinghua.edu.cn [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Chen, Tao; Sun, Qi; Wang, Jie [Institute of Nuclear and New Energy Technology of Tsinghua University, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 100084 (China); Yu, Suyuan, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Education, Tsinghua University, Beijing 100084 (China)

2017-05-15

Highlights: • A theoretical analysis is used to predict the total graphite dust release for an AVR LOCA. • Similarity criteria must be satisfied between the experiment and the actual HTGR system. • Model experiments should be conducted to predict the graphite dust resuspension rate. - Abstract: The graphite dust movement behavior is significant for the safety analyses of high-temperature gas cooled reactor (HTGR). The graphite dust release for accident conditions is an important source term for HTGR safety analyses. Depressurization release tests are not practical in HTGR because of a radioactivity release to the environment. Thus, a theoretical analysis and similarity principles were used to design a group of modeling experiments. Modeling experiments for fan start-up and depressurization process and actual experiments of helium circulator start-up in an HTGR were used to predict the rate of graphite dust resuspension and the graphite dust concentration, which can be used to predict the graphite dust release during accidents. The modeling experiments are easy to realize and the helium circulator start-up test does not harm the reactor system or the environment, so this experiment program is easily achieved. The revised Rock’n’Roll model was then used to calculate the AVR reactor release. The calculation results indicate that the total graphite dust releases during a LOCA will be about 0.65 g in AVR.
Innovative approaches to the Management of Irradiated Nuclear Graphite Wastes: Addressing the Challenges through International Collaboration with Project 'GRAPA'

International Nuclear Information System (INIS)

Wickham, A.J.; Ojovan, M.; O'Sullivan, P.; )

2017-01-01

There exists more than 250.000 tonnes of irradiated (and therefore radioactive) nuclear graphite (i-graphite) in the world, primarily as a result of the development of graphite-moderated power-reactor systems, initially for defence and subsequently for commercial purposes. Only a very small number of such plants have been dismantled and, for most cases, the final destiny of the irradiated graphite remains unresolved. Future high-temperature reactor programmes, such as the Chinese HTR-PM development, will produce more graphite and carbonaceous wastes from both structural components and the fuel pebbles (which are approximately 96% carbonaceous), the latter producing a continuous stream of so-called 'operational waste'. The problem of dismantling irradiated graphite reactor stacks, possibly distorted through neutron damage and in some cases degraded further by radiation-chemical attack by gaseous coolants, and then finding the appropriate treatments and final destiny of the material, has exercised both the European Union and the International Atomic Energy Agency for more than 25 years, seeking to address the different issues and available disposal solutions in different IAEA Member States. An IAEA collaborative research programme on treatment options has recently been completed, and an active group of international specialists in this area has now been established as part of the IAEA International Decommissioning Network under the envelope of Project 'GRAPA' (Irradiated Graphite Processing Approaches), which includes representatives from Belgium, China, France, Germany, India, Italy, Lithuania, Rep. of Korea, Romania, Spain, Switzerland, Ukraine and the Russian Federation with direct responsibilities for various parts of the decommissioning and graphite-disposal process in a variety of reactor designs. Interest has also been expressed by colleagues from Sweden and Japan. Work is in progress on a number of topic areas where weaknesses in the
Multi purpose research reactor

International Nuclear Information System (INIS)

Raina, V.K.; Sasidharan, K.; Sengupta, Samiran; Singh, Tej

2006-01-01

At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor
Review of ageing management of NPPs - Experience feed back form research reactors

International Nuclear Information System (INIS)

Bhatnagar, A.; Gujarathi, R.I.; Chowdhury, R.; Tikku, A.C.

2002-01-01

Ageing of Systems, Structures and Components (SSCs) is a natural process and sets in along with the construction and commissioning of plants in spite of best design provisions and maintenance practices. Plant operators and maintainers need to plan and take measures against ageing degradation of SSCs to maintain the high standards of safety. As safety is a continuously evolving phenomenon, incorporating safety upgrades from time to time and carrying out ageing management towards improved safety for research and power reactors is very important. Cirus research reactor which was commissioned in 1960 and Tarapur Atomic power station which was commissioned in 1969 are two such examples of older generation nuclear plants in India which are presently undergoing extensive refurbishment towards implementation of ageing management programme. The 40 MWt Cirus Research Reactor located at the Bhabha Atomic Research Centre, Mumbai, is a vertical closed tank type reactor with natural uranium as fuel, demineralised light water as primary coolant, heavy water as moderator and graphite as reflector. The reflector and the thermal shields are cooled by reactor building ventilation system. Sea water is used as secondary coolant. The reactor vessel is made of aluminium and has 199 lattice tubes rolled into top and bottom tube sheets. It has an expansion joint between the top tube sheet and the shell to allow for thermal expansion. The reactor operated very efficiently till early nineties after which the ageing degradation of SSCs started affecting the reactor operation. Plant availability factor showed a declining trend due to frequent breakdown of equipment. Detailed performance review was carried out for various equipment and a list of equipment that needed replacement was prepared. Equipment, for which availability of spares was becoming difficult due to obsolescence, were also included in this list. Detailed ageing studies were then taken up on various SSCs. The SSCs were
Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)
Applications: fission, nuclear reactors. Fission: the various ways for reactors and cycles

International Nuclear Information System (INIS)

Bacher, P.

1997-01-01

A historical review is presented concerning the various nuclear reactor systems developed in France by the CEA: the UNGG (graphite-gas) system with higher CO 2 pressures, bigger fuel assemblies and powers higher than 500 MW e, allowed by studies on reactor physics, cladding material developments and reactor optimization; the fast neutron reactor system, following the graphite-gas development, led to the Superphenix reactor and important progress in simulation based on experiment and return of experience; and the PWR system, based on the american license, which has been successfully accommodated to the french industry and generates up to 75% of the electric power in France

Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

International Nuclear Information System (INIS)

Higinbotham, W.A.

1994-01-01

The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 ± 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF
Reactor safety research program. A description of current and planned reactor safety research sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research

International Nuclear Information System (INIS)

1975-06-01

The reactor safety research program, sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, is described in terms of its program objectives, current status, and future plans. Elements of safety research work applicable to water reactors, fast reactors, and gas cooled reactors are presented together with brief descriptions of current and planned test facilities. (U.S.)
Study on graphite samples for nuclear usage

International Nuclear Information System (INIS)

Suarez, J.C.M.; Silva Roseira, M. da

1994-01-01

Available as short communication only. The graphite, due to its properties (mechanical strength, thermal conductivity, high-temperature stability, machinability etc.) have many industrial applications, and consequently, an important strategic value. In the nuclear area, it has been used as moderator and reflector of neutrons in the fission process of uranium. The graphite can be produced from many types of carbonaceous materials by a variety of process dominated by the manufactures. This is the reason why there are in the world market a lot of graphite types with different physical and mechanical properties. The present investigation studies some physical characteristics of the graphite samples destined to use in a nuclear reactor. (author). 8 refs, 1 fig, 1 tab
A three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition in graphite components of advanced gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Morgan, D.O.; Robinson, A.T.; Allen, D.A.; Picton, D.J.; Thornton, D.A. [TCS, Serco, Rutherford House, Olympus Park, Quedgeley, Gloucester, Gloucestershire GL2 4NF (United Kingdom); Shaw, S.E. [EDF Energy, Barnet Way, Barnwood, Gloucester GL4 3RS (United Kingdom)

2011-07-01

This paper describes the development of a three-dimensional methodology for the assessment of neutron damage and nuclear energy deposition (or nuclear heating) throughout the graphite cores of the UK's Advanced Gas-cooled Reactors. Advances in the development of the Monte Carlo radiation transport code MCBEND have enabled the efficient production of detailed fully three-dimensional models that utilise three-dimensional source distributions obtained from Core Follow data supplied by the reactor physics code PANTHER. The calculational approach can be simplified to reduce both the requisite number of intensive radiation transport calculations, as well as the quantity of data output. These simplifications have been qualified by comparison with explicit calculations and they have been shown not to introduce significant systematic uncertainties. Simple calculational approaches are described that allow users of the data to address the effects on neutron damage and nuclear energy deposition predictions of the feedback resulting from the mutual dependencies of graphite weight loss and nuclear energy deposition. (authors)
Research reactor DHRUVA

International Nuclear Information System (INIS)

Veeraraghaven, N.

1990-01-01

DHRUVA, a 100 MWt research reactor located at the Bhabha Atomic Research Centre, Bombay, attained first criticality during August, 1985. The reactor is fuelled with natural uranium and is cooled, moderated and reflected by heavy water. Maximum thermal neutron flux obtained in the reactor is 1.8 X 10 14 n/cm 2 /sec. Some of the salient design features of the reactor are discussed in this paper. Some important features of the reactor coolant system, regulation and protection systems and experimental facilities are presented. A short account of the engineered safety features is provided. Some of the problems that were faced during commissioning and the initial phase of power operation are also dealt upon
CER. Research reactors in France

International Nuclear Information System (INIS)

Estrade, Jerome

2012-01-01

Networking and the establishment of coalitions between research reactors are important to guarantee a high technical quality of the facility, to assure well educated and trained personnel, to harmonize the codes of standards and the know-ledge of the personnel as well as to enhance research reactor utilization. In addition to the European co-operation, country-specific working groups have been established for many years, such as the French research reactor Club d'Exploitants des Reacteurs (CER). It is the association of French research reactors representing all types of research reactors from zero power up to high flux reactors. CER was founded in 1990 and today a number of 14 research reactors meet twice a year for an exchange of experience. (orig.)
Design of a graphite-moderated {sup 241}Am-Li neutron field to simulate reactor spectra

Energy Technology Data Exchange (ETDEWEB)

Tsujimura, N., E-mail: tsujimura.norio@jaea.go.j [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan); Yoshida, T. [Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, 4-33, Tokai-mura, Ibaraki-ken, 319-1194 (Japan)

2010-12-15

A neutron calibration field using {sup 241}Am-Li sources and a moderator was designed to simulate the neutron fields found outside a reactor. The moderating assembly selected for the design calculation consists of a cube of graphite blocks with dimensions of 50 cm by 50 cm by 50 cm, in which the {sup 241}Am-Li sources are placed. Monte Carlo calculations revealed the optimal depth of the source to be 15 cm. This moderated neutron source can be used to provide a test field that has a large number of intermediate energy neutrons with a small portion of MeV component.
Experiences in the emptying of waste silos containing solid nuclear waste from graphite- moderated reactors

International Nuclear Information System (INIS)

Wall, S.; Schwarz, T.

2003-01-01

Before reactor sites can be handed over for ultimate decommissioning, at some sites silos containing waste from operations need to be emptied. The form and physical condition of the waste demands sophisticated retrieval technologies taking into account the onsite situation in terms of infrastructure and silo geometry. Furthermore, in the case of graphite moderated reactors, this waste usually includes several tonnes of graphite waste requiring special HVAC and dust handling measures. RWE NUKEM Group has already performed several contracts dealing with such emptying tasks. Of particular interest for the upcoming decommissioning projects in France might be the activities at Vandellos, Spain and Trawsfynnyd, UK. Retrieval System for Vandellos NPP is discussed. Following an international competitive tender exercise, RWE NUKEM won the contract to provide a turn-key retrieval system. This involved the design, manufacture and installation of a system built around the modules of a 200 kg capacity version of the ARTISAN manipulator system. The ARTISAN 200 manipulator, with remote slave arm detach facility, was deployed on a telescopic mast inserted into the silos through the roof penetrations. The manipulator deployed a range of tools to gather the waste and load it into a transfer basket, deployed through an adjacent penetration. After commissioning, the system cleared the vaults in less than the scheduled period with no failures. At the Trawsfynnyd Magnox plants two types of intermediate level waste (ILW) accumulated on site; namely Miscellaneous Activated Components (MAC) and Fuel Element Debris (FED). MAC is predominantly components that have been activated by the reactor core and then discharged. FED mainly consists of fuel cladding produced when fuel elements were prepared for dispatch to the reprocessing facility. RWE NUKEM Ltd. was awarded a contract to design, supply, commission and operate equipment to retrieve, pack and immobilize the two waste streams. Major
Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2008

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment and practice in the nuclear reactor training, Irradiation for activation analyses, RI productions, fission tracks etc. In the fiscal year 2008, the research reactor JRR-3 was operated for 7 cycles (cycle operation : 26days/cycle) for utilization sharing of facility. The research reactor JRR-4 was not operated in 2008. Because a crack was found on the weld of the aluminum cladding of a graphite reflector element. JRR-4 has remained shutdown until the reflector elements were replaced. The volume contains 250activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, and others submitted by the users in JAEA and other Organizations. (author)
Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2013

International Nuclear Information System (INIS)

2014-12-01

The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30MW in December 2001 and achieved the 950degC of outlet coolant temperature at the outside the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2013, we started to prepare the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 when the Pacific coast of Tohoku Earthquake (2011.3.11) occurred. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2013. (author)
Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)
Computerized reactor monitor and control for research reactors

International Nuclear Information System (INIS)

Buerger, L.; Vegh, E.

1981-09-01

The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)
Assessment of structural materials inside the reactor pool of the Dalat research reactor

International Nuclear Information System (INIS)

Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

2010-01-01

Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of
Nuclear reactor instrumentation at research reactor renewal

International Nuclear Information System (INIS)

Baers, B.; Pellionisz, P.

1981-10-01

The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)
Brookhaven National Laboratory site environmental report for calendar year 1994

International Nuclear Information System (INIS)

Naidu, J.R.; Royce, B.A.

1995-05-01

This report documents the results of the Environmental Monitoring Program at Brookhaven National Laboratory and presents summary information about environmental compliance for 1994. To evaluate the effect of Brookhaven National Laboratory's operations on the local environment, measurements of direct radiation, and a variety of radionuclides and chemical compounds in ambient air, soil, sewage effluent, surface water, groundwater, fauna and vegetation were made at the Brookhaven National Laboratory site and at sites adjacent to the Laboratory
Strategic planning for research reactors. Guidance for reactor managers

International Nuclear Information System (INIS)

2001-04-01

The purpose of this publication is to provide guidance on how to develop a strategic plan for a research reactor. The IAEA is convinced of the need for research reactors to have strategic plans and is issuing a series of publications to help owners and operators in this regard. One of these covers the applications of research reactors. That report brings together all of the current uses of research reactors and enables a reactor owner or operator to evaluate which applications might be possible with a particular facility. An analysis of research reactor capabilities is an early phase in the strategic planning process. The current document provides the rationale for a strategic plan, outlines the methodology of developing such a plan and then gives a model that may be followed. While there are many purposes for research reactor strategic plans, this report emphasizes the use of strategic planning in order to increase utilization. A number of examples are given in order to clearly illustrate this function
Thermogravimetric and Differential Scanning Calorimetric Behavior of Ball-Milled Nuclear Graphite

Energy Technology Data Exchange (ETDEWEB)

Kim, Eung Seon; Kim, Min Hwan; Kim, Yong Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi Hyun; Cho, Seung Yon [National Fusion Research Institute, Daejeon (Korea, Republic of)

2013-10-15

An examination was made to characterize the oxidation behavior of ball-milled nuclear graphite powder through a TG-DSC analysis. With the ball milling time, the BET surface area increased with the reduction of particle size, but decreased with the chemisorptions of O{sub 2} on the activated surface. The enhancement of the oxidation after the ball milling is attributed to both increases in the specific surface area and atomic scale defects in the graphite structure. In a high temperature gas-cooled reactor, nuclear graphite has been widely used as fuel elements, moderator or reflector blocks, and core support structures owing to its excellent moderating power, mechanical properties and machinability. For the same reason, it will be used in a helium cooled ceramic reflector test blanket module for the ITER. Each submodule has a seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebbles packed tritium breeder layers, and a reflector layer packed with 1 mm diameter graphite pebbles to reduce the volume of beryllium. The abrasion of graphite structures owing to relative motion or thermal cycle during operation may produce graphite dust. It is expected that graphite dust will be more oxidative than bulk graphite, and thus the oxidation behavior of graphite dust must be examined to analyze the safety of the reactors during an air ingress accident. In this study, the thermal stability of ball-milled graphite powder was investigated using a simultaneous thermogravimeter-differential scanning calorimeter.
Ageing of research reactors

International Nuclear Information System (INIS)

Ciocanescu, M.

2001-01-01

Historically, many of the research institutions were centred on a research reactor facility as main technological asset and major source of neutrons for research. Important achievements were made in time in these research institutions for development of nuclear materials technology and nuclear safety for nuclear energy. At present, ageing of nuclear research facilities among these research reactors and ageing of staff are considerable factors of reduction of competence in research centres. The safe way of mitigation of this trend deals with ageing management by so called, for power reactors, Plant Life Management and new investments in staff as investments in research, or in future resources of competence. A programmatic approach of ageing of research reactors in correlation with their actual and future utilisation, will be used as a basis for safety evaluation and future spending. (author)
The radioactivity estimation of 14C and 3H in graphite waste samples of the KRR-2.

Science.gov (United States)

Reyoung Kim, Hee

2013-09-01

The radioactivity of (14)C and (3)H in graphite samples from the dismantled Korea Research Reactor-2 (the KRR-2) site was analyzed by high-temperature oxidation and liquid scintillation counting, and the graphite waste was suggested to be disposed of as a low-level radioactive waste. The graphite samples were oxidized at a high temperature of 800 °C, and their counting rates were measured by using a liquid scintillation counter (LSC). The combustion ratio of the graphite was about 99% on the sample with a maximum weight of 1g. The recoveries from the combustion furnace were around 100% and 90% in (14)C and (3)H, respectively. The minimum detectable activity was 0.04-0.05 Bq/g for the (14)C and 0.13-0.15 Bq/g for the (3)H at the same background counting time. The activity of (14)C was higher than that of (3)H over all samples with the activity ratios of the (14)C to (3)H, (14)C/(3)H, being between 2.8 and 25. The dose calculation was carried out from its radioactivity analysis results. The dose estimation gave a higher annual dose than the domestic legal limit for a clearance. It was thought that the sampled graphite waste from the dismantled research reactor was not available for reuse or recycling and should be monitored as low-level radioactive waste. Copyright © 2013 Elsevier Ltd. All rights reserved.
High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

International Nuclear Information System (INIS)

1981-08-01

Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies

High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

Energy Technology Data Exchange (ETDEWEB)

1981-08-01

Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.
Research on plasma core reactors

International Nuclear Information System (INIS)

Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

1977-01-01

Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF 6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm 3 aluminum canister in the central region was fueled with UF 6 gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation
Study on the properties of the fuel compact for High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho; Lee, Young-woo; Cho, Moon-sung

2015-01-01

High Temperature Gas-cooled Reactors (HTGR), one of the Gen-IV reactors, have been using the fuel element which is manufactured by the graphite matrix, surrounding Tristructural-isotropic (TRISO)-coated Uranium particles. Factors with these characteristics effecting on the matrix of fuel compact are chosen and their impacts on the properties are studied. The fuel elements are considered with two types of concepts for HTGR, which are the block type reactor and the pebble bed reactor. In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength with the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and the two kinds of candidate binder (Phenol and Polyvinyl butyral) were chosen and mixed with each other, formed and heated to measure mechanical properties. The objective of this research is to optimize the materials and composition of the mixture and the forming process by evaluating the mechanical properties before/after carbonization and heat treatment. From the mechanical test results, the mechanical properties of graphite pellets was related to the various conditions such as the contents and kinds of binder, the kinds of graphite and the heat treatments. In the result of the compressive strength and Vicker's hardness, the 10 wt% phenol binder added R+S graphite pellet was relatively higher mechanical properties than other pellets. The contents of Phenol binder, the kinds of graphite powder and the temperature of carbonization and heat treatment are considered important factors for the properties. To optimize the mechanical properties of fuel elements, the role of binders and the properties of graphites will be investigated as
Brookhaven National Laboratory site environmental report for calendar year 1994

Energy Technology Data Exchange (ETDEWEB)

Naidu, J.R.; Royce, B.A. [eds.

1995-05-01

This report documents the results of the Environmental Monitoring Program at Brookhaven National Laboratory and presents summary information about environmental compliance for 1994. To evaluate the effect of Brookhaven National Laboratory`s operations on the local environment, measurements of direct radiation, and a variety of radionuclides and chemical compounds in ambient air, soil, sewage effluent, surface water, groundwater, fauna and vegetation were made at the Brookhaven National Laboratory site and at sites adjacent to the Laboratory.
Beryllium and graphite performance in ITER during a disruption

International Nuclear Information System (INIS)

Hassanein, A.; Ehst, D.A.; Gahl, J.

1994-01-01

Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed. ((orig.))
Beryllium and graphite performance in ITER during a disruption

International Nuclear Information System (INIS)

Hassanein, A.; Ehst, D.A.; Gahl, J.

1993-09-01

Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed
Siting of research reactors

International Nuclear Information System (INIS)

1987-01-01

The purpose of this document is to develop criteria for siting and the site-related design basis for research reactors. The concepts presented in this document are intended as recommendations for new reactors and are not suggested for backfitting purposes for facilities already in existence. In siting research reactors serious consideration is given to minimizing the effects of the site on the reactor and the reactor on the site and the potential impact of the reactor on the environment. In this document guidance is first provided on the evaluation of the radiological impact of the installation under normal reactor operation and accident conditions. A classification of research reactors in groups is then proposed, together with a different approach for each group, to take into account the relevant safety problems associated with facilities of different characteristics. Guidance is also provided for both extreme natural events and for man-induced external events which could affect the safe operation of the reactor. Extreme natural events include earthquakes, flooding for river or coastal sites and extreme meteorological phenomena. The feasibility of emergency planning is finally considered for each group of reactors
Graphite Isotope Ratio Method Development Report: Irradiation Test Demonstration of Uranium as a Low Fluence Indicator

International Nuclear Information System (INIS)

Reid, B.D.; Gerlach, D.C.; Love, E.F.; McNeece, J.P.; Livingston, J.V.; Greenwood, L.R.; Petersen, S.L.; Morgan, W.C.

1999-01-01

This report describes an irradiation test designed to investigate the suitability of uranium as a graphite isotope ratio method (GIRM) low fluence indicator. GIRM is a demonstrated concept that gives a graphite-moderated reactor's lifetime production based on measuring changes in the isotopic ratio of elements known to exist in trace quantities within reactor-grade graphite. Appendix I of this report provides a tutorial on the GIRM concept
Digital control of research reactors

International Nuclear Information System (INIS)

Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

1991-01-01

Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind
A microstructural study of dynamic crack propagation in nuclear graphites

International Nuclear Information System (INIS)

Burchell, T.D.; McEnaney, B.; Tucker, M.O.; Rose, A.P.G.

1986-01-01

This paper reports a new microstructural study of dynamic crack propagation in three nuclear graphites: (i) PGA, the moderator material in UK Magnox reactors; (ii) IMl-24, the moderator material in UK Advanced gas cooled reactors (AGR); and (iii) a pitch coke graphite, which is used in the fabrication of AGR fuel sleeves. The fracture mechanisms in nuclear graphites are initiated by microcrack formation at low stresses. Typically, microcracks form in regions of well-aligned binder or at favourably-oriented pores, where stress is concentrated. With increasing applied loads, microcracks propagate taking advantage of easy cleavage paths or linking with pores. Eventually, coalescence of such cracks and inherent porosity produces a crack of critical length for fast fracture. (orig./MM)
Molten salt reactors: reactor cores

International Nuclear Information System (INIS)

1983-01-01

In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr
MEQALAC development at Brookhaven

International Nuclear Information System (INIS)

Gammel, G.; Brodowski, J.; Keane, J.; Maschke, A.; Meier, E.; Mobley, R.; Sanders, R.

1981-01-01

A novel method of transporting and accelerating high brightness ion beams, called MEQALAC, has been developed at Brookhaven. The concept and its motivation will be described first, with reference to other sources for detail, and then the performance of two operating MEQALAC's will be presented
A study of the coefficient of thermal expansion of nuclear graphites

International Nuclear Information System (INIS)

Hacker, P.J.

2001-02-01

This thesis presents the results of a study of the Coefficient of Thermal Expansion (CTE) of two grades of nuclear graphite that are used as the moderator in the Magnox and Advanced Gas-Cooled reactors operated in the UK. This work has two main aims, the first is to characterise those elements of the graphite microstructure that control CTE within these materials and to relate these to the effects induced within the reactor. The second is to develop a microstructural model, of general applicability, that can initially be applied to model the CTE changes within the graphites under reactor conditions (neutron irradiation and radiolytic oxidation). These aims have been met by study in three interlinked areas, theoretical, experimental and modelling. Previous to this study, a loose assembly of single crystals together with changes in small scale nanometric porosity (Mrozowski cracks) were used to describe CTE behaviour of nuclear graphite both as-received and under reactor conditions. Within the experimental part of this thesis the graphite nanostructure was studied using, primarily, Transmission Electron Microscopy (TEM). This work concluded that structure on this scale was complex and that the loose assembly of single crystals was a poor microstructural approximation for modelling the CTE of these materials. Other experimental programmes measured the CTE of highly oxidised samples and simulated the effects of irradiation. The former discovered that CTE remained largely unaffected to high weight losses. This insensitivity was explained by ''The Continuous Network Hypothesis'' that was also related to classical percolation theory. The final part of the thesis modelled an abstraction of the key microstructural features identified in the previous parts of the thesis. This approach has been applied to AGR moderator graphite where it has successfully modelled the thermal expansion behaviour of the as-received, irradiated and oxidised material. (author)
Overview of EU research activities in transmutation and innovative reactor systems within the Euratom framework programmes

International Nuclear Information System (INIS)

Bhatnagar, V.

2009-01-01

European Community (EC) (currently 27 Member States) shared-cost research has been organised in Framework Programmes (FP) of durations of 4 - 5 years since 1984. The 6th European Atomic Energy Community (EURATOM) Framework Programme (2002 - 06) and the current 7th FP (2007 - 11) have been allocated a fission research budget respectively of 209 and 287 Million Euro from the EC. There are 10 projects (total budget 70 M Euro, EC contribution 38 M Euro) in all aspects of transmutation ranging from road-mapping exercise to large integrated projects on accelerator driven systems, lead-cooled fast critical systems for waste transmutation, technology, fuel, accelerator facilities for nuclear data etc. In Innovative Reactor concepts, there are about half-a-dozen projects (total budget 30 M Euro, EC contribution 16 M Euro) including High Temperature Reactors, Gas-cooled Fast reactors, road-mapping exercise on sodium fast reactors etc. The main research and training activities in FP7 are: management of radioactive waste, reactor systems, radiation protection, infrastructures, human resources and mobility and training. In the two call for proposals (2007 and 2008) in FP7, 8 projects have been accepted in transmutation and innovative reactor concepts (total budget 53 M Euro, EC contribution 32 M Euro). These research projects cover activities ranging from materials, fuels, treatment of irradiated graphite waste, European sodium fast reactor to the establishment of a Central Design Team of a fast-spectrum transmutation device in Europe. The third call for proposals is underway requesting proposals on nuclear data, thermal hydraulics, gas and lead-cooled fast reactor systems with a total EC budget of 20 M Euro. International collaboration is an important element of the EU research policy. This overview paper will present elements of the strategy of EURATOM research and training in waste management including accelerator driven transmutation systems and Innovative reactor concepts
Fission Product Sorptivity in Graphite

Energy Technology Data Exchange (ETDEWEB)

Tompson, Jr., Robert V. [Univ. of Missouri, Columbia, MO (United States); Loyalka, Sudarshan [Univ. of Missouri, Columbia, MO (United States); Ghosh, Tushar [Univ. of Missouri, Columbia, MO (United States); Viswanath, Dabir [Univ. of Missouri, Columbia, MO (United States); Walton, Kyle [Univ. of Missouri, Columbia, MO (United States); Haffner, Robert [Univ. of Missouri, Columbia, MO (United States)

2015-04-01

Both adsorption and absorption (sorption) of fission product (FP) gases on/into graphite are issues of interest in very high temperature reactors (VHTRs). In the original proposal, we proposed to use packed beds of graphite particles to measure sorption at a variety of temperatures and to use an electrodynamic balance (EDB) to measure sorption onto single graphite particles (a few μm in diameter) at room temperature. The use of packed beds at elevated temperature is not an issue. However, the TPOC requested revision of this initial proposal to included single particle measurements at elevated temperatures up to 1100 °C. To accommodate the desire of NEUP to extend the single particle EDB measurements to elevated temperatures it was necessary to significantly revise the plan and the budget. These revisions were approved. In the EDB method, we levitate a single graphite particle (the size, surface characteristics, morphology, purity, and composition of the particle can be varied) or agglomerate in the balance and measure the sorption of species by observing the changes in mass. This process involves the use of an electron stepping technique to measure the total charge on a particle which, in conjunction with the measured suspension voltages for the particle, allows for determinations of mass and, hence, of mass changes which then correspond to measurements of sorption. Accommodating elevated temperatures with this type of system required a significant system redesign and required additional time that ultimately was not available. These constraints also meant that the grant had to focus on fewer species as a result. Overall, the extension of the original proposed single particle work to elevated temperatures added greatly to the complexity of the proposed project and added greatly to the time that would eventually be required as well. This means that the bulk of the experimental progress was made using the packed bed sorption systems. Only being able to recruit one
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
Investigation on wear behavior of graphite baII under different pneumatic conveying environments

International Nuclear Information System (INIS)

Chen Zhipeng; Zheng Yanhua; Shi Lei; Yu Suyuan

2014-01-01

An experimental platform was built in the Institute of Nuclear and New Energy Technology (INET) to investigate the wear behavior of the graphite ball under the operational condition of the high temperature gas-cooled reactor (HTGR) fuel handling system. In this experimental platform, a series of experiments were carried out under different pneumatic conveying environments with the graphite balls, which were made of the material same as the fuel element matrix graphite (A3) of the 10 MW high temperature gas cooled reactor (HTR-10). The effect of the pneumatic conveying condition on the wear rate of graphite ball has been investigated, and the results include: (1) There is an obvious linear relationship between the wear rate and the feeding velocity of graphite ball elevated in the stainless steel elevating tube, and the wear rate will increase with the increase of the feeding velocity. (2) The wear rate of graphite ball under helium environment is significantly greater than that under air and nitrogen environments, which is caused by the different effects of various gas environments on mechanical properties of graphite. (author)
IAEA safety standards for research reactors

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

The general structure of the IAEA Safety Standards and the process for their development and revision are briefly presented and discussed together with the progress achieved in the development of Safety Standards for research reactor. These documents provide the safety requirements and the key technical recommendations to achieve enhanced safety. They are intended for use by all organizations involved in safety of research reactors and developed in a way that allows them to be incorporated into national laws and regulations. The author reviews the safety standards for research reactors and details their specificities. There are 4 published safety standards: 1) Safety assessment of research reactors and preparation of the safety analysis report (35-G1), 2) Safety in the utilization and modification of research reactors (35-G2), 3) Commissioning of research reactors (NS-G-4.1), and 4) Maintenance, periodic testing and inspection of research reactors (NS-G-4.2). There 5 draft safety standards: 1) Operational limits and conditions and operating procedures for research reactors (DS261), 2) The operating organization and the recruitment, training and qualification of personnel for research reactors (DS325), 3) Radiation protection and radioactive waste management in the design and operation of research reactors (DS340), 4) Core management and fuel handling at research reactors (DS350), and 5) Grading the application of safety requirements for research reactors (DS351). There are 2 planned safety standards, one concerning the ageing management for research reactor and the second deals with the control and instrumentation of research reactors
Research reactors and materials testing

International Nuclear Information System (INIS)

Vidal, H.

1986-01-01

Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr
Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

Interactions of D-T neutrons in graphite and lithium blankets of fusion reactors

International Nuclear Information System (INIS)

Ofek, R.

1986-05-01

The present study deals with integral experiment and calculation of neutron energy spectra in bulks of graphite which is used as a reflector in blankets of fusion reactors, and lithium, the material of the blanket on which lithium is bred due to neutron interactions. The collimated beam configuration enables - due to the almost monoenergeticity and unidirectionality of the neutrons impinging on the target - to identify fine details in the measured spectra, and also facilitates the absolute normalization of the spectra. The measured and calculated spectra are generally in a good agreement and in a very good agreement at mesh points close to the system axis. A few conclusions may be drawn: a) the collimated beam source configuration is a sensitive tool for measuring neutron energy spectra with a high resolution, b) the method of unfolding proton-recoil spectra measured with a NE-213 scintillator should be improved, c) MCNP and DOT 4.2 may be used as complementary codes for neutron transport calculations of fusion blankets and deep-penetration problems, d) the updating of the cross-section libraries and checking by integral experiments is highly important for the design of fusion blankets. The present study may be regarded as an important course in the research and development of tools for the design of fusion blankets
Physical security at research reactors

International Nuclear Information System (INIS)

Clark, R.A.

1977-01-01

Of the 84 non-power research facilities licensed under 10 CFR Part 50, 73 are active (two test reactors, 68 research reactors and three critical facilities) and are required by 10 CFR Part 73.40 to provide physical protection against theft of SNM and against industrial sabotage. Each licensee has developed a security plan required by 10 CFR Part 50.34(c) to demonstrate the means of compliance with the applicable requirements of 10 CFR Part 73. In 1974, the Commission provided interim guidance for the organization and content of security plans for (a) test reactors, (b) medium power research and training reactors, and (c) low power research and training reactors. Eleven TRIGA reactors, with power levels greater than 250 kW and all other research and training reactors with power levels greater than 100 kW and less than or equal to 5,000 kW are designated as medium power research and training reactors. Thirteen TRIGA reactors with authorized power levels less than 250 kW are considered to be low power research and training reactors. Additional guidance for complying with the requirements of 73.50 and 73.60, if applicable, is provided in the Commission's Regulatory Guides. The Commission's Office of Inspection and Enforcement inspects each licensed facility to assure that an approved security plan is properly implemented with appropriate procedures and physical protection systems
Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

International Nuclear Information System (INIS)

Karahan, Aydın; Kazimi, Mujid S.

2013-01-01

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors
Using graphitic foam as the bonding material in metal fuel pins for sodium fast reactors

Energy Technology Data Exchange (ETDEWEB)

Karahan, Aydın, E-mail: karahan@alum.mit.edu; Kazimi, Mujid S.

2013-10-15

The study evaluates the possible use of graphite foam as the bonding material between U–Pu–Zr metallic fuel and steel clad for sodium fast reactor applications using FEAST-METAL fuel performance code. Furthermore, the applicability of FEAST-METAL to the advanced fuel designs is demonstrated. Replacing the sodium bond with a chemically stable foam material would eliminate fuel clad metallurgical interactions, and allow for fuel swelling under low external stress. Hence, a significant improvement is expected for the steady state and transient performance. FEAST-METAL was used to assess the thermo-mechanical behavior of the new fuel form and a reference metallic fuel pin. Nearly unity conversion ratio, 75% smear density U–15Pu–6Zr metallic fuel pin with sodium bond, and T91 cladding was selected as a reference case. It was found that operating the reference case at high clad temperatures (600–660 °C) results in (1) excessive clad wastage formation/clad thinning due to lanthanide migration and formation of brittle phases at clad inner surface, and (2) excessive clad hoop strain at the upper axial section due mainly to the occurrence of thermal creep. The combination of these two factors may lead to cladding breach. The work concludes that replacing the sodium bond with 80% porous graphite foam and reducing the fuel smear density to 70%, it is likely that the fuel clad metallurgical interaction would be eliminated while the fuel swelling is allowed without excessive fuel clad mechanical interaction. The suggested design appears as an alternative for a high performance metallic fuel design for sodium fast reactors.
Ion irradiation of {sup 37}Cl implanted nuclear graphite: Effect of the energy deposition on the chlorine behavior and consequences for the mobility of {sup 36}Cl in irradiated graphite

Energy Technology Data Exchange (ETDEWEB)

Toulhoat, N., E-mail: nelly.toulhoat@univ-lyon1.fr [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); CEA/DEN, Centre de Saclay (France); Moncoffre, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Bérerd, N.; Pipon, Y. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Université de Lyon, Université Lyon, IUT Lyon-1 département chimie (France); Blondel, A. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Andra, Châtenay-Malabry (France); Galy, N. [Université de Lyon, Université Lyon 1, CNRS/IN2P3, UMR5822, Institut de Physique Nucléaire de Lyon (IPNL) (France); Sainsot, P. [Université de Lyon, Université Lyon 1, LaMCoS, INSA-Lyon, CNRS UMR5259 (France); Rouzaud, J.-N.; Deldicque, D. [Laboratoire de Géologie de l’Ecole Normale Supérieure (ENS), Paris, UMR CNRS-ENS 8538 (France)

2015-09-15

Graphite is used in many types of nuclear reactors due to its ability to slow down fast neutrons without capturing them. Whatever the reactor design, the irradiated graphite waste management has to be faced sooner or later regarding the production of long lived or dose determining radioactive species such as {sup 14}C, {sup 3}H or {sup 36}Cl. The first carbon dioxide cooled, graphite moderated nuclear reactors resulted in a huge quantity of irradiated graphite waste for which the management needs a previous assessment of the radioactive inventory and the radionuclide’s location and speciation. As the detection limits of usual spectroscopic methods are generally not adequate to detect the low concentration levels (<1 ppm) of the radionuclides, we used an indirect approach based on the implantation of {sup 37}Cl, to simulate the presence of {sup 36}Cl. Our previous studies show that temperature is one of the main factors to be considered regarding the structural evolution of nuclear graphite and chlorine mobility during reactor operation. However, thermal release of chlorine cannot be solely responsible for the depletion of the {sup 36}Cl inventory. We propose in this paper to study the impact of irradiation and its synergetic effects with temperature on chlorine release. Indeed, the collision of the impinging neutrons with the graphite matrix carbon atoms induces mainly ballistic collisions. However, a small part of the recoil carbon atom energy is also transferred to the lattice through electronic excitation. This paper aims at elucidating the effects of the different irradiation regimes (ballistic and electronic) using ion irradiation, on the mobility of implanted {sup 37}Cl, taking into account the initial disorder level of the nuclear graphite.
Studies of the role of molten materials in interactions with UO2 and graphite

International Nuclear Information System (INIS)

Fink, J.K.; Heiberger, J.J.; Leibowitz, L.

1979-01-01

Graphite, which is being considered as a lower reactor shield in gas-cooled fast reactors, would be contacted by core debris during a core disruptive accident. Information on the interaction of graphite, UO 2 , and stainless steel is needed in assessing the safety of the GCFR. In an ongoing study of the interaction of graphite, UO 2 , and stainless steel, the effects of the steel components have been investigated by electron microprobe scans, x-ray diffraction, and reaction-rate measurements. Experiments to study the role of the reaction product, FeUC 2 , in the interaction suggested that FeUC 2 promotes the interaction by acting as a carrier to bring graphite to the reaction site. Additional experiments using pyrolytic graphite show that while the reaction rate is decreased at 2400 K, at higher temperatures the rate is similar to that using other grades of graphite
Management of research reactor ageing

International Nuclear Information System (INIS)

1995-03-01

As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs
Management of research reactor ageing

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-03-01

As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs.
Measurement and Calculation of Gamma Radiation from HWZPR Reactor

International Nuclear Information System (INIS)

Jalali, Majid

2006-01-01

HWZPR is a research reactor with natural uranium fuel, D 2 O moderator and graphite reflector with maximum power of 100 W. It is a suitable means for theoretical research and heavy water reactor experiments. Neutrons from the core participate in different nuclear reactions by interactions with fuel, moderator, graphite and the concrete around the reactor. The results of these interactions are the production of prompt gammas in the environment. Useful information is gained by the reactor gamma spectrum measurement from point of view of relative quantity and energy distribution of direct and scattered radiations. Reactor gamma ray spectrum has been gathered in different places around the reactor by HPGe detector. In analysis of these spectra, 1 H(n,γ) 2 H, 16 O(n,n'γ) 16 O, 2 H(n,γ) 3 H and 238 U(n,γ) 239 U reactions occurring in reactor moderator and fuel, are important. The measured spectrum has been primarily estimated by the MCNP code. There is agreement between the code and the experiments in some points. The scattered gamma rays from 27 Al (n,γ) 28 Al reaction in the reactor tank, are the most among the gammas scattered in the reactor environment. Also the dose calculations by MCNP code show that 72% of gamma dose belongs to the energy range 3-11 MeV from reactor gamma spectrum and the danger of exposure from the reactor high-energy photons is serious. (author)
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.)
Neutronic reactor

International Nuclear Information System (INIS)

Wende, C.W.J.

1976-01-01

The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures
Reactor Materials Research

International Nuclear Information System (INIS)

Van Walle, E.

2001-01-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)
Reactor Materials Research

Energy Technology Data Exchange (ETDEWEB)

Van Walle, E

2001-04-01

The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)
Kinetics of Chronic Oxidation of NBG-17 Nuclear Graphite by Water Vapor

Energy Technology Data Exchange (ETDEWEB)

Contescu, Cristian I [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Burchell, Timothy D [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mee, Robert [Univ. of Tennessee, Knoxville, TN (United States)

2015-05-01

This report presents the results of kinetic measurements during accelerated oxidation tests of NBG-17 nuclear graphite by low concentration of water vapor and hydrogen in ultra-high purity helium. The objective is to determine the parameters in the Langmuir-Hinshelwood (L-H) equation describing the oxidation kinetics of nuclear graphite in the helium coolant of high temperature gas-cooled reactors (HTGR). Although the helium coolant chemistry is strictly controlled during normal operating conditions, trace amounts of moisture (predictably < 0.2 ppm) cannot be avoided. Prolonged exposure of graphite components to water vapor at high temperature will cause very slow (chronic) oxidation over the lifetime of graphite components. This behavior must be understood and predicted for the design and safe operation of gas-cooled nuclear reactors. The results reported here show that, in general, oxidation by water of graphite NBG-17 obeys the L-H mechanism, previously documented for other graphite grades. However, the characteristic kinetic parameters that best describe oxidation rates measured for graphite NBG-17 are different than those reported previously for grades H-451 (General Atomics, 1978) and PCEA (ORNL, 2013). In some specific conditions, certain deviations from the generally accepted L-H model were observed for graphite NBG-17. This graphite is manufactured in Germany by SGL Carbon Group and is a possible candidate for the fuel elements and reflector blocks of HTGR.
Vapour pressure of caesium over nuclear graphite

International Nuclear Information System (INIS)

Faircloth, R.L.; Pummery, F.C.W.

1976-01-01

The vapour pressure of caesium over a fine-grained isotropic moulded gilsocarbon nuclear graphite intended for use in the manufacture of fuel tubes for the high temperature reactor has been determined as a function of temperature and concentration by means of the Knudsen effusion technique. The concentration range 0 to 10 μg caesium/g graphite was investigated and it was concluded that a Langmuir adsorption situation exists under these conditions. (author)
The research reactors their contribution to the reactors physics

International Nuclear Information System (INIS)

Barral, J.C.; Zaetta, A.; Johner, J.; Mathoniere, G.

2000-01-01

The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)
Improvement of research reactor sustainability

International Nuclear Information System (INIS)

Ciocanescu, M.; Paunoiu, C.; Toma, C.; Preda, M.; Ionila, M.

2010-01-01

The Research Reactors as is well known have numerous applications in a wide range of science technology, nuclear power development, medicine, to enumerate only the most important. The requirements of clients and stack-holders are fluctuating for the reasons out of control of Research Reactor Operating Organization, which may ensure with priority the safety of facility and nuclear installation. Sustainability of Research Reactor encompasses several aspects which finally are concentrated on safety of Research Reactor and economical aspects concerning operational expenses and income from external resources. Ensuring sustainability is a continuous, permanent activity and also it requests a strategic approach. The TRIGA - 14 MW Research Reactor detains a 30 years experience of safe utilization with good performance indicators. In the last 4 years the reactor benefited of a large investment project for modernization, thus ensuring the previous performances and opening new perspectives for power increase and for new applications. The previous core conversion from LEU to HEU fuel accomplished in 2006 ensures the utilization of reactor based on new qualified European supplier of TRIGA LEU fuel. Due to reduction of number of performed research reactors, the 14 MW TRIGA modernized reactor will play a significant role for the following two decades. (author)
Helium generation and diffusion in graphite and some carbides

International Nuclear Information System (INIS)

Holt, J.B.; Guinan, M.W.; Hosmer, D.W.; Condit, R.H.; Borg, R.J.

1976-01-01

The cross section for the generation of helium in neutron irradiated carbon was found to be 654 mb at 14.4 MeV and 744 mb at 14.9 MeV. Extrapolating to 14.1 MeV (the fusion reactor spectrum) gives 615 mb. The diffusion of helium in dense polycrystalline graphite and in pyrographite was measured and found to be D = 7.2 x 10 -7 m 2 s -1 exp (-80 kJ/RT). It is assumed that diffusion is primarily in the basal plane direction in crystals of the graphite. In polycrystalline graphite the path length is a factor of √2 longer than the measured distance due to the random orientation mismatch between successive grains. Isochronal anneals (measured helium release as the specimen is steadily heated) were run and maximum release rates were found at 200 0 C in polycrystalline graphite, 1000 0 C in pyrographite, 1350 0 C in boron carbide, and 1350 0 and 2400 0 C (two peaks) in silicon carbide. It is concluded that in these candidates for curtain materials in fusion reactors the helium releases can probably occur without bubble formation in graphites, may occur in boron carbide, but will probably cause bubble formation in silicon carbide. 7 figures
Strategy for Handling and Treatment of INPP RBMK-1500 Irradiated Graphite

International Nuclear Information System (INIS)

Oryšaka, A.

2016-01-01

There are two RBMK-1500 water-cooled graphite-moderated channel-type power reactors at Ignalina NPP. After the final shutdown of the INPP, radioactive i-graphite dismantling, handling, conditioning, storage and disposal is an important part of the decommissioning activities. The core of the INPP unit 1 and 2 contains about 3600 tons of i-graphite. Formation of activation products strongly depends on the contents of impurities, operational mode and concentration of impurities in the graphite. The case study for INPP envisages the analysis of possibilities of graphite handling and treatment in the context of immediate decommissioning. (author)
Special graphites; Graphites speciaux

Energy Technology Data Exchange (ETDEWEB)

Leveque, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

A large fraction of the work undertaken jointly by the Commissariat a l'Energie Atomique (CEA) and the Pechiney Company has been the improvement of the properties of nuclear pile graphite and the opening up of new fields of graphite application. New processes for the manufacture of carbons and special graphites have been developed: forged graphite, pyro-carbons, high density graphite agglomeration of graphite powders by cracking of natural gas, impervious graphites. The physical properties of these products and their reaction with various oxidising gases are described. The first irradiation results are also given. (authors) [French] Ameliorer les proprietes du graphite nucleaire pour empilements et ouvrir de nouveaux domaines d'application au graphite constituent une part importante de l'effort entrepris en commun par le Commissariat a l'Energie Atomique (CEA) et la compagnie PECHINEY. Des procedes nouveaux de fabrication de carbones et graphites speciaux ont ete mis au point: graphite forge, pyrocarbone, graphite de haute densite, agglomeration de poudres de graphite par craquage de gaz naturel, graphites impermeables. Les proprietes physiques de ces produits ainsi que leur reaction avec differents gaz oxydants sont decrites. Les premiers resultats d'irradiation sont aussi donnes. (auteurs)

Operation, test, research and development of the high temperature engineering test reactor (HTTR). (FY2005)

International Nuclear Information System (INIS)

2007-03-01

The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power. The full power operation of 30 MW was attained in December, 2001, and then JAERI (JAEA) received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. In fiscal 2005 year, periodical inspection and overhaul of reactivity control system were conducted, and safety demonstration tests were promoted. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2005. (author)
Status of Chronic Oxidation Studies of Graphite

Energy Technology Data Exchange (ETDEWEB)

Contescu, Cristian I. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Mee, Robert W. [Univ. of Tennessee, Knoxville, TN (United States)

2016-05-01

Graphite will undergo extremely slow, but continuous oxidation by traces of moisture that will be present, albeit at very low levels, in the helium coolant of HTGR. This chronic oxidation may cause degradation of mechanical strength and thermal properties of graphite components if a porous oxidation layer penetrates deep enough in the bulk of graphite components during the lifetime of the reactor. The current research on graphite chronic oxidation is motivated by the acute need to understand the behavior of each graphite grade during prolonged exposure to high temperature chemical attack by moisture. The goal is to provide the elements needed to develop predictive models for long-time oxidation behavior of graphite components in the cooling helium of HTGR. The tasks derived from this goal are: (1) Oxidation rate measurements in order to determine and validate a comprehensive kinetic model suitable for prediction of intrinsic oxidation rates as a function of temperature and oxidant gas composition; (2) Characterization of effective diffusivity of water vapor in the graphite pore system in order to account for the in-pore transport of moisture; and (3) Development and validation of a predictive model for the penetration depth of the oxidized layer, in order to assess the risk of oxidation caused damage of particular graphite grades after prolonged exposure to the environment of helium coolant in HTGR. The most important and most time consuming of these tasks is the measurement of oxidation rates in accelerated oxidation tests (but still under kinetic control) and the development of a reliable kinetic model. This report summarizes the status of chronic oxidation studies on graphite, and then focuses on model development activities, progress of kinetic measurements, validation of results, and improvement of the kinetic models. Analysis of current and past results obtained with three grades of showed that the classical Langmuir-Hinshelwood model cannot reproduce all
PALLADIUM-FACILITATED ELECTROLYTIC DECHLORINATION OF 2-CHLOROBIPHENYL USING A GRANULAR-GRAPHITE ELECTRODE.

Science.gov (United States)

Palladium-assisted electrocatalytic dechlorination of 2-chlorobiphenyl (2-Cl BP) in aqueous solutions was conducted in a membrane-separated electrochemical reactor with granular-graphite packed electrodes. The dechlorination took place at a granular-graphite cathode while Pd was ...
Planning the Decommissioning of Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Podlaha, J., E-mail: pod@ujv.cz [Nuclear Research Institute Rez, 25068 Rez (Czech Republic)

2013-08-15

In the Czech Republic, three research nuclear reactors are in operation. According to the valid legislation, preliminary decommissioning plans have been prepared for all research reactors in the Czech Republic. The decommissioning plans shall be updated at least every 5 years. Decommissioning funds have been established and financial resources are regularly deposited. Current situation in planning of decommissioning of research reactors in the Czech Republic, especially planning of decommissioning of the LVR-15 research reactor is described in this paper. There appeared new circumstances having wide impact on the decommissioning planning of the LVR-15 research reactor: (1) Shipment of spent fuel to the Russian Federation for reprocessing and (2) preparation of processing of radioactive waste from reconstruction of the VVR-S research reactor (now LVR-15 research reactor). The experience from spent fuel shipment to the Russian Federation and from the process of radiological characterization and processing of radioactive waste from reconstruction of the VVR-S research reactor (now the LVR-15 research reactor) and the impact on the decommissioning planning is described in this paper. (author)
Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

International Nuclear Information System (INIS)

Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo

2011-01-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the
Ageing implementation and refurbishment development at the IEA-R1 nuclear research reactor: a 15 years experience

Energy Technology Data Exchange (ETDEWEB)

Cardenas, Jose Patricio N.; Ricci Filho, Walter; Carvalho, Marcos R. de; Berretta, Jose Roberto; Marra Neto, Adolfo, E-mail: ahiru@ipen.b, E-mail: wricci@ipen.b, E-mail: carvalho@ipen.b, E-mail: jrretta@ipen.b, E-mail: amneto@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

IPEN (Instituto de Pesquisas Energeticas e Nucleares) is a nuclear research center established into the Secretary of Science and Technology from the government of the state of Sao Paulo, and administered both technically and financially by Comissao Nacional de Energia Nuclear (CNEN), a federal government organization under the Ministry of Science and Technology. The institute is located inside the campus of the University of Sao Paulo, Sao Paulo city, Brazil. One of major nuclear facilities at IPEN is the IEA-R1 nuclear research reactor. It is the unique Brazilian research reactor with substantial power level suitable for application with research in physics, chemistry, biology and engineering, as well as radioisotope production for medical and other applications. Designed and built by Babcok-Wilcox, in accordance with technical specifications established by the Brazilian Nuclear Energy Commission, and financed by the US Atoms for Peace Program, it is a swimming pool type reactor, moderated and cooled by light water and uses graphite and beryllium as reflector elements. The first criticality was achieved on September 16, 1957 and the reactor is currently operating at 4.0 MW on a 64h per week cycle. Since 1996, an IEA-R1 reactor ageing study was established at the Research Reactor Center (CRPq) related with general deterioration of components belonging to some operational systems, as cooling towers from secondary cooling system, piping and pumps, sample irradiation devices, radiation monitoring system, fuel elements, rod drive mechanisms, nuclear and process instrumentation and safety operational system. Although basic structures are almost the same as the original design, several improvements and modifications in components, systems and structures had been made along reactor life. This work aims to show the development of the ageing program in the IEA-R1 reactor and the upgrading (modernization) that was carried out, concerning several equipment and system in the
On the Thermal Conductivity Change of Matrix Graphite Materials after Neutron Irradiation

Energy Technology Data Exchange (ETDEWEB)

Lee, Young-Woo; Yeo, Seunghwan; Kim, Eung-Seon; Sah, Injin; Park, Daegyu; Kim, Youngjun; Cho, Moon Sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this work, the variations of the thermal conductivity of the A3-3 matrix graphite after neutron irradiation is discussed as well as of the IG-110 graphite for comparison. Neutron irradiation of the graphite specimens was carried out as a part of the first irradiation test of KAERI's coated particle fuel specimens by use of Hanaro research reactor. This work can be summarized as follows: 1) In the evaluation of the specific heat of the graphite materials, various literature data were used and the variations of the specific heat data of all the graphite specimens are observed well agreed, irrespectively of the difference in specimens (graphite and matrix graphite and irradiated and un-irradiated). 2) This implies that it should be reasonable that for both structural graphite and fuel matrix graphite, and even for the neuron-irradiated graphite, any of these specific heat data set be used in the calculation of the thermal conductivity. 3) For the irradiated A3-3 matrix graphite specimens, the thermal conductivity decreased on both directions. On the radial direction, the tendency of variation upon temperature is similar to that of unirradiated specimen, i.e., decreasing as the temperature increases. 4) In the German irradiation experiments with A3-27 matrix graphite specimens, the thermal conductivity of the un-irradiated specimen shows a decrease and that of irradiated specimen is nearly constant as the temperature increases. 5) The thermal conductivity of the irradiated IG-110 was considerably decreased compared with that of un-irradiated specimens The difference of the thermal conductivity of un-irradiated and irradiated IG-110 graphite specimens is much larger than that of un-irradiated and irradiated A3-3 matrix graphite specimens.
Nuclear research reactors in Brazil

Energy Technology Data Exchange (ETDEWEB)

Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)
Nuclear research reactors in Brazil

International Nuclear Information System (INIS)

Cota, Anna Paula Leite; Mesquita, Amir Zacarias

2011-01-01

The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)
Industrial structure at research reactor suppliers

International Nuclear Information System (INIS)

Roegler, H.-J.; Bogusch, E.; Friebe, T.

2001-01-01

Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community
Draft pre-application safety evaluation report for the modular high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Williams, P.M.; King, T.L.; Wilson, J.N.

1989-03-01

This draft safety evaluation report (SER) presents the preliminary results of a pre-application design review for the standard modular high-temperature gas-cooled reactor (MHTGR) (Project 672). The MHTGR conceptual design was submitted by the U.S. Department of Energy (DOE) in accordance with the U.S. Nuclear Regulatory Commission(NRC) 'Statement of Policy for the Regulation of Advanced Nuclear Power Plants' (51 FR 24643), which provides for early Commission review and interaction. The standard MHTGR consists of four identical reactor modules, each with a thermal output of 350 MWt, coupled with two steam turbine-generator sets to produce a total plant electrical output of 540 MWe. The reactors are helium cooled and graphite moderated and utilize ceramically coated particle-type nuclear fuel. The design includes passive reactor-shutdown and decay-heat-removal features. The staff and its contractors at the Oak Ridge National Laboratory and the Brookhaven National Laboratory have reviewed this design with emphasis on those unique provisions in the design that accomplish the key safety functions of reactor shutdown, decay-heat removal, and containment of radioactive material. This report presents the NRC staff's technical evaluation of those features in the MHTGR design important to safety, including their proposed research and testing needs. In addition this report presents the criteria proposed by the NRC staff to judge the acceptability of the MHTGR design and, where possible, includes statements on the potential of the MHTGR to meet these criteria. However, it should be recognized that final conclusions in all matters discussed in this report require approval by the Commission. Final determination on the acceptability of the MHTGR standard design is contingent on receipt and evaluation of additional information requested from DOE pertaining to the adequacy of the containment design and on the following: (1) satisfactory resolution of open safety issues identified
Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

International Nuclear Information System (INIS)

2014-01-01

Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.
Understanding Creep Mechanisms in Graphite with Experiments, Multiscale Simulations, and Modeling

Energy Technology Data Exchange (ETDEWEB)

Eapen, Jacob [North Carolina State Univ., Raleigh, NC (United States); Murty, Korukonda [North Carolina State Univ., Raleigh, NC (United States); Burchell, Timothy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2014-06-02

Disordering mechanisms in graphite have a long history with conflicting viewpoints. Using Raman and x-ray photon spectroscopy, electron microscopy, x-ray diffraction experiments and atomistic modeling and simulations, the current project has developed a fundamental understanding of early-to-late state radiation damage mechanisms in nuclear reactor grade graphite (NBG-18 and PCEA). We show that the topological defects in graphite play an important role under neutron and ion irradiation.
Isotropic nuclear graphites; the effect of neutron irradiation

International Nuclear Information System (INIS)

Lore, J.; Buscaillon, A.; Mottet, P.; Micaud, G.

1977-01-01

Several isotropic graphites have been manufactured using different forming processes and fillers such as needle coke, regular coke, or pitch coke. Their properties are described in this paper. Specimens of these products have been irradiated in the fast reactor Rapsodie between 400 to 1400 0 C, at fluences up to 1,7.10 21 n.cm -2 PHI.FG. The results show an isotropic behavior under neutron irradiation, but the induced dimensional changes are higher than those of isotropic coke graphites although they are lower than those of conventional extruded graphites made with the same coke
BROOKHAVEN: High energy gold

International Nuclear Information System (INIS)

Bleser, Ed

1992-01-01

On April 24, Brookhaven's Alternating Gradient Synchrotron (AGS) started to deliver gold ions at 11.4 GeV per nucleon (2,000 GeV per ion) to experimenters who were delighted not only to receive the world's highest energy gold beam but also to receive it on schedule
Research reactor`s role in Korea

Energy Technology Data Exchange (ETDEWEB)

Choi, C-O [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

1996-12-31

After a TRIGA MARK-II was constructed in 1962, new research activity of a general nature, utilizing neutrons, prevailed in Korea. Radioisotopes produced from the MARK-II played a good role in the 1960`s in educating people as to what could be achieved by a neutron source. Because the research reactor had implanted neutron science in the country, another TRIGA MARK-III had to be constructed within 10 years after importing the first reactor, due to increased neutron demand from the nuclear community. With the sudden growth of nuclear power, however, the emphasis of research changed. For a while research activities were almost all oriented to nuclear power plant technology. However, the specifics of nuclear power plant technology created a need for a more highly capable research reactor like HANARO 30MWt. HANARO will perform well with irradiation testing and other nuclear programs in the future, including: production of key radioisotopes, doping of silicon by transmutation, neutron activation analysis, neutron beam experiments, cold neutron source. 3 tabs., 2 figs.
BROOKHAVEN: Glueballs, hybrids and exotics

Energy Technology Data Exchange (ETDEWEB)

Chung, S. -U.

1988-12-15

A workshop at Brookhaven from August 29 to September 1 looked at the current status of hadron spectroscopy beyond the realm of states conventionally built up from quarks and discussed future experimental effort to explore such exotic states.
Oxidation of PCEA nuclear graphite by low water concentrations in helium

Energy Technology Data Exchange (ETDEWEB)

Contescu, Cristian I., E-mail: ContescuCI@ornl.gov [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6087 (United States); Mee, Robert W. [Department of Business Analytics and Statistics, University of Tennessee, Knoxville, TN 37996-0525 (United States); Wang, Peng [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831-6087 (United States); Romanova, Anna V.; Burchell, Timothy D. [Department of Business Analytics and Statistics, University of Tennessee, Knoxville, TN 37996-0525 (United States)

2014-10-15

Accelerated oxidation tests were performed to determine kinetic parameters of the chronic oxidation reaction (i.e. slow, continuous, and persistent) of PCEA graphite in contact with helium coolant containing low moisture concentrations in high temperature gas-cooled reactors. To the authors’ knowledge such a study has not been done since the detailed analysis of reaction of H-451 graphite with steam (Velasquez, Hightower, Burnette, 1978). Since that H-451 graphite is now unavailable, it is urgently needed to characterize chronic oxidation behavior of new graphite grades that are being considered for use in gas-cooled reactors. The Langmuir–Hinshelwood mechanism of carbon oxidation by water results in a non-linear reaction rate expression, with at least six different parameters. They were determined in accelerated oxidation experiments that covered a large range of temperatures (800–1100 °C), and partial pressures of water (15–850 Pa) and hydrogen (30–150 Pa) and used graphite specimens thin enough (4 mm) in order to avoid diffusion effects. Data analysis employed a statistical method based on multiple likelihood estimation of parameters and simultaneous fitting of non-linear equations. The results show significant material-specific differences between graphite grades PCEA and H-451 which were attributed to microstructural dissimilarity between the two materials. It is concluded that kinetic data cannot be transferred from one graphite grade to another.
The effective neutron temperature in heated graphite sleeves

Energy Technology Data Exchange (ETDEWEB)

Shaw, J A; Small, V G [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

1963-08-15

In a series of oscillator measurements carried out in the reactor NERO the variation of the relative reaction rates of cadmium and boron absorbers has been used to determine the effective neutron temperature inside heated graphite sleeves. This work extends the scope of similar oscillator measurements previously carried out in DIMPLE, in that the bulk moderator is now graphite as opposed to D{sub 2}O in the former case. (author)
Leaching of 14C and 36Cl from irradiated French graphite

International Nuclear Information System (INIS)

Gray, W.J.; Morgan, W.C.

1989-08-01

The leach rates of 14 C and 36 Cl were measured on solid cylindrical samples prepared from irradiated graphite blocks supplied by the French Commissariat a l'Energie Atomique (CEA). Static leach tests were conducted in deionized water at 20 degree C for 13 weeks. The graphite samples were completely submerged in the water, and the entire volume of water was changed and analyzed at weekly intervals for the first three weeks and biweekly thereafter. Large differences in the leach rates of both 14 C and 36 Cl were observed between samples machined from the different blocks. In general, the leach rates were much higher than those measured in an earlier study with graphite obtained from a block removed from one of the Hanford reactors. The data from this study are compared with those from the previous study using the Hanford-reactor graphite. Implications of the data from both studies regarding possible rate-limiting mechanisms are discussed. 4 refs., 8 figs., 3 tabs

Changes in the physical and mechanical properties of graphite on irradiation in ditolylmethane

International Nuclear Information System (INIS)

Gavrilin, A.I.; Lebedev, I.G.; Sudakova, N.V.; Rizvanov, V.K.

1987-01-01

Results are presented from the irradiation and mechanical and structural testing of four grades of graphite - GMZ, VPG, MPG-6, and PG-50 - for use as moderator materials in organic cooled and graphite moderated reactors. Irradiation was carried out in the ARBUS-AST-1 reactor. Photomicrography was used to determine pore structure and ultimate strength in bending and compression was determined mechanically. Irradiation was found to increase the strength of GMZ, PMG-6, and PG-50 considerably, due to the healing of microdefects as a result of the pores filling with radiolysis products from the coolant, ditolylmethane. Conversely, VPG graphite, which has closed porosity, lost strength on irradiation
Australian research reactor studies

International Nuclear Information System (INIS)

McCulloch, D.B.

1978-01-01

The Australian AEC has two research reactors at the Lucas Heights Research Establishment, a 10 HW DIDO class materials testing reactor, HIFAR, and a smaller 100kW reactor MOATA, which was recently upgraded from 10kW power level. Because of the HIFAR being some 20 years old, major renewal and repair programmes are necessary to keep it operational. To enable meeting projected increases in demand for radioisotopes, plans for a new reactor to replace the HIFAR have been made and the design criteria are described in the paper. (author)
Thermal Properties of G-348 Graphite

Energy Technology Data Exchange (ETDEWEB)

McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Swank, W. David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cottle, David L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Valentin, Francisco I. [City Univ. (CUNY), NY (United States)

2017-04-01

Fundamental measurements have been obtained in the INL Graphite Characterization Laboratory to deduce the temperature dependence of thermal conductivity for G-348 isotropic graphite, which has been used by City College of New York in thermal experiments related to gas-cooled nuclear reactors. Measurements of thermal diffusivity, mass, volume and thermal expansion were converted to thermal conductivity in accordance with ASTM Standard Practice C781-08 (R-2014). Data are tabulated and a preliminary correlation for the thermal conductivity is presented as a function of temperature from laboratory temperature to 1000C.
Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

International Nuclear Information System (INIS)

Bozoki, G.; Kohut, P.; Fitzpatrick, R.

1989-02-01

This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs
Effect of reacting surface density on the overall graphite oxidation rate

International Nuclear Information System (INIS)

Oh, Chang; Kim, Eung; Lim, Jong; Schultz, Richard; Petti, David

2009-01-01

Graphite oxidation in an air-ingress accident is presently a very important issue for the reactor safety of the very high temperature gas cooled-reactor (VHTR), the concept of the next generation nuclear plant (NGNP) because of its potential problems such as mechanical degradation of the supporting graphite in the lower plenum of the VHTR might lead to core collapse if the countermeasure is taken carefully. The oxidation process of graphite has known to be affected by various factors, including temperature, pressure, oxygen concentration, types of graphite, graphite shape and size, flow distribution, etc. However, our recent study reveals that the internal pore characteristics play very important roles in the overall graphite oxidation rate. One of the main issues regarding graphite oxidation is the potential core collapse problem that may occur following the degradation of graphite mechanical strength. In analyzing this phenomenon, it is very important to understand the relationship between the degree of oxidization and strength degradation. In addition, the change of oxidation rate by graphite oxidation degree characterization by burn-off (ratio of the oxidized graphite density to the original density) should be quantified because graphite strength degradation is followed by graphite density decrease, which highly affects oxidation rates and patterns. Because the density change is proportional to the internal pore surface area, they should be quantified in advance. In order to understand the above issues, the following experiments were performed: (1) Experiment on the fracture of the oxidized graphite and validation of the previous correlations, (2) Experiment on the change of oxidation rate using graphite density and data collection, (3) Measure the BET surface area of the graphite. The experiments were performed using H451 (Great Lakes Carbon Corporation) and IG-110 (Toyo Tanso Co., Ltd) graphite. The reason for the use of those graphite materials is because
2012 review of French research reactors

International Nuclear Information System (INIS)

Estrade, Jerome

2013-01-01

Proposed by the French Reactor Operators' Club (CER), the meeting and discussion forum for operators of French research reactors, this report first gives a brief presentation of these reactors and of their scope of application, and a summary of highlights in 2012 for each of them. Then, it proposes more detailed presentations and reviews of characteristics, activities, highlights, objectives and results for the different types of reactors: neutron beam reactors (Orphee, High flux reactor-Laue-Langevin Institute or HFR-ILL), technological irradiation reactors (Osiris and Phenix), training reactors (Isis and Azur), reactors for safety research purposes (Cabri and Phebus), reactors for neutronic studies (Caliban, Prospero, Eole, Minerve and Masurca), and new research reactors (the RES facility and the Jules Horowitz reactor or JHR)
Brookhaven National Laboratory site environmental report for calendar year 1996

Energy Technology Data Exchange (ETDEWEB)

Schroeder, G.L.; Paquette, D.E.; Naidu, J.R.; Lee, R.J.; Briggs, S.L.K.

1998-01-01

This report documents the results of the Environmental Monitoring Program at Brookhaven National Laboratory and summarizes information about environmental compliance for 1996. To evaluate the effect of Brookhaven National Laboratory`s operations on the local environment, measurements of direct radiation, and of a variety of radionuclides and chemical compounds in the ambient air, soil, sewage effluent, surface water, groundwater, fauna, and vegetation were made at the Brookhaven National Laboratory site and at adjacent sites. The report also evaluates the Laboratory`s compliance with all applicable guides, standards, and limits for radiological and non-radiological emissions and effluents to the environment.
Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

International Nuclear Information System (INIS)

Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

2013-06-01

(Nuclear and Energy Research Institute) was founded in 1956 with the main purpose of doing research and development in the field of nuclear energy and its applications. It is located at the campus of University of Sao Paulo (USP), in the city of Sao Paulo, in an area of nearly 500, 000 m2. It has over 1.000 employees and 40% of them have qualification at master or doctor level The institute is recognized as a national leader institution in research and development (R and D) in the areas of radiopharmaceuticals, industrial applications of radiation, basic nuclear research, nuclear reactor operation and nuclear applications, materials science and technology, laser technology and applications. Along with the R and D, it has a strong educational activity, having a graduate program in Nuclear Technology, in association with the University of Sao Paulo, ranked as the best university in the country. The Federal Government Evaluation institution CAPES, granted to this course grade 6, considering it a program of Excellence. This program started at 1976 and has awarded 458 Ph.D. degrees and 937 master degrees since them. The actual graduate enrollment is around 400 students. One of major nuclear installation at IPEN is the IEA-R1 research reactor; it is the only Brazilian research reactor with substantial power level suitable for its utilization in researches concerning physics, chemistry, biology and engineering as well as for producing some useful radioisotopes for medical and other applications. IEA-R1 reactor is a swimming pool type reactor moderated and cooled by light water and uses graphite and beryllium as reflectors. The first criticality was achieved on September 16, 1957. The reactor is currently operating at 4.5 MW power level with an operational schedule of continuous 64 hours a week. In 1996 a Modernization Program was started to establish recommendations in order to mitigate equipment and structures ageing effects in the reactor components, detect and evaluate
Radiolytic graphite oxidation revisited

International Nuclear Information System (INIS)

Minshall, P.C.; Sadler, I.A.; Wickham, A.J.

1996-01-01

The importance of radiolytic oxidation in graphite-moderated CO 2 -cooled reactors has long been recognised, especially in the Advanced Gas-Cooled Reactors where potential rates are higher because of the higher gas pressure and ratings than the earlier Magnox designs. In all such reactors, the rate of oxidation is partly inhibited by the CO produced in the reaction and, in the AGR, further reduced by the deliberate addition of CH 4 . Significant roles are also played by H 2 and H 2 O. This paper reviews briefly the mechanisms of these processes and the data on which they are based. However, operational experience has demonstrated that these basic principles are unsatisfactory in a number of respects. Gilsocarbon graphites produced by different manufacturers have demonstrated a significant difference in oxidation rate despite a similar specification and apparent equivalence in their pore size and distribution, considered to be the dominant influence on oxidation rate for a given coolant-gas composition. Separately, the inhibiting influence of CH 4 , which for many years had been considered to arise from the formation of a sacrificial deposit on the pore walls, cannot adequately be explained by the actual quantities of such deposits found in monitoring samples which frequently contain far less deposited carbon than do samples from Magnox reactors where the only source of such deposits is the CO. The paper also describes the current status of moderator weight-loss predictions for Magnox and AGR Moderators and the validation of the POGO and DIFFUSE6 codes respectively. 2 refs, 5 figs
Investigation on Conversion of I-Graphite from Decommissioning of Chernobyl NPP into a Stable Waste Form Acceptable for Long Term Storage and Disposal

International Nuclear Information System (INIS)

Zlobenko, Borys; Fedorenko, Yriy; Yatzenko, Victor; Shabalin, Borys; Skripkin, Vadim

2016-01-01

For Ukraine, the main radiocarbon ( 14 C) source is irradiated graphite from Chernobyl Nuclear Power Plant. The ChNPP is a decommissioned nuclear power station about 14 km northwest of the city of Chernobyl, and 110 km north of Kyiv. The ChNPP had four RBMK reactor units. The commissioning of the first reactor in 1977 was followed by reactor No. 2 (1978), No. 3 (1981), and No.4 (1983). Reactors No.3 and 4 were second generation units, whereas Nos.1 and 2 were first-generation units. RBMK is an acronym for ''High Power Channel-type Reactor'' of a class of graphite-moderated nuclear power reactor with individual fuel channels that uses ordinary water as its coolant and graphite as its moderator. The combination of graphite moderator and water coolant is found in no other type of nuclear reactor
Retrospective review of the clinical BNCT trial at Brookhaven National Laboratory

International Nuclear Information System (INIS)

Diaz, A.Z.; Chanana, A.D.; Coderre, J.A.; Ma, R.

2000-01-01

The primary objective of the phase I/II dose escalation studies was to evaluate the safety of the boronophenylalanine-fructose (BPA-F) mediated boron neutron capture therapy (BNCT) in subjects with glioblastoma multiforme (GBM). A secondary objective was to retrospectively assess the palliation of GBM by BNCT. Fifty-three subjects with GBM were treated under multiple dose escalation protocols at the Brookhaven Medical Research Reactor (BMRR). Twenty-six subjects were treated using one field, 17 subjects were treated using 2 fields and 10 subjects were treated using 3 fields. BPA-F related toxicity was not observed. The maximum radiation dose to a volume of approximately 1 cc of the normal brain varied from 8.9 to 15.9 gray-equivalent (Gy-Eq). The volume-weighted average radiation dose to normal brain varied from 1.9 to 9.5 Gy-Eq. Six RTOG (Radiation Therapy Oncology Group) grade 3 or 4 toxicities were attributed to BNCT. Four of the 53 subjects are still alive with 3 of them free of recurrent disease with over two years follow-up. The median times to progression and median survival time from diagnosis were 28.4 weeks and 12.8 months respectively. (author)
Brookhaven: RHIC magnets

International Nuclear Information System (INIS)

Heppelman, Steve

1990-01-01

Last year, Brookhaven's proposal for a Relativistic Heavy Ion Collider - RHIC - was scrutinized by the US Department of Energy and deemed to be ready for construction funding. The hope is that the money will be voted soon so that construction can get underway at the start of the new US financial year in October. The 3.8 kilometre RHIC tunnel was completed ten years ago for the doomed Isabelle/CBA proton collider project
Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-05-01

The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)
Processing of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal. Results of a Coordinated Research Project. Companion CD-ROM

International Nuclear Information System (INIS)

2016-05-01

Graphite is widely used in the nuclear industry and in research facilities and this has led to increasing amounts of irradiated graphite residing in temporary storage facilities pending disposal. This publication arises from a coordinated research project (CRP) on the processing of irradiated graphite to meet acceptance criteria for waste disposal. It presents the findings of the CRP, the general conclusions and recommendations. The topics covered include, graphite management issues, characterization of irradiated graphite, processing and treatment, immobilization and disposal. Included on the attached CD-ROM are formal reports from the participants
Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors; Etude de la tenue de la gaine interne pour-element combustible a refroidissement interne et externe d'un reacteur graphite-gaz

Energy Technology Data Exchange (ETDEWEB)

Boudouresque, B; Courcon, P; Lestiboubois, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm{sup 2} gas pressure, should remain in contact with the fuel. (authors) [French] La cartouche d'un element combustible annulaire, a refroidissement interne et externe pour reacteur graphite-gaz, est composee d'un tube combustible en uranium, d'une gaine externe et d'une gaine interne en alliage de magnesium. Pour que l'echange thermique entre la gaine interne et le combustible soit bon, il faut que la gaine reste appliquee sur l'uranium quel que soit le regime de temperature. Cette note a pour but de montrer comment, d'apres une etude theorique, le jeu combustible-gaine interne varie au cours des operations de gainage, de chargement dans le reacteur, et des cyclages thermiques. Les parametres suivants sont etudies: diametres de tube, pression du gaz caloporteur, temperature d'entree du gaz, plasticite de l'alliage de gaine. Il est montre que, quel que soit le regime de fonctionnement, la gaine interne d'un element 77 x 95, en projet pour un reacteur graphite-gaz sous pression de 40 kg/cm{sup 2}, doit rester appliquee sur le combustible. (auteurs)
Calculation of reactivity of control rods in graphite moderated reactors

International Nuclear Information System (INIS)

Nakata, H.

1978-01-01

A study about the method of calculation for the reactivity of control rods in graphite-moderated critical assemblies, is presented. The result of theoretical calculation, developed by super celles and Nordheim-Scalettar methods are compared with experimental results for the critical Assembly of General Atomic. The two methods are then applicable to reactivity calculation of the control rods of graphite moderated critical assemblies [pt
Clinical trials of boron neutron capture therapy [in humans] [at Beth Israel Deaconess Medical Center][at Brookhaven National Laboratory

International Nuclear Information System (INIS)

Wallace, Christine

2001-01-01

Assessment of research records of Boron Neutron Capture Therapy was conducted at Brookhaven National Laboratory and Beth Israel Deaconess Medical Center using the Code of Federal Regulations, FDA Regulations and Good Clinical Practice Guidelines. Clinical data were collected FR-om subjects' research charts, and differences in conduct of studies at both centers were examined. Records maintained at Brookhaven National Laboratory were not in compliance with regulatory standards. Beth Israel's records followed federal regulations. Deficiencies discovered at both sites are discussed in the reports
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
Research reactor decommissioning planning - It is never too early to start

International Nuclear Information System (INIS)

Eby, R.S.; Buttram, C.; Ervin, P.; Lundberg, L.; Hertel, N.; Marske, S.G.

2003-01-01

Whether an organization is in the process of designing, constructing or operating nuclear research reactors, past experiences prove it is never too early to start planning for the eventual decontamination, dismantlement and decommissioning (DD and D) of the reactor. If one waits until writing the Decommissioning Plan to seriously think about the DD and D activities, they have lost a key opportunity to be able to efficiently and effectively carry out the DD and D activities and will end up spending large sums of unnecessary funds during the DD and D. This paper will review all phases of research reactor decommissioning from characterization through planning, to eventual DD and D and license termination and highlight areas where early planning can significantly reduce the financial, safety and schedule risks associated with the DD and D activities. CH2M HILL served as the Executive Engineer for the Georgia Institute of Technology and the State of Georgia to oversee the successful DD and D of their tank type research reactor. CH2M HILL is currently serving as the DD and D contractor for the University of Virginia pool type UVAR and the low power CAVALIER research reactors and as the characterization and Decommissioning Planning contractor for the University of Michigan Ford Nuclear Reactor. Through these activities, an array of lessons learned have been compiled that will prove invaluable to the research reactor owner when they eventually face the DD and D challenge. As an example, in almost every case CH2M HILL has been involved in reactor DD and D, less than adequate up-front characterization has significantly impacted the ultimate DD and D process cost and schedule. Due to regulatory reasons, intrusive characterization may not always be possible prior to DD and D. However, a thorough understanding of the materials of construction and the quantities of additives or impurities present in those materials; e.g., cobalt in stainless steel, rare earth elements or
Replacement research reactor for Australia

International Nuclear Information System (INIS)

Miller, Ross

1998-01-01

In 1992, the Australian Government commissioned a review into the need for a replacement research reactor. That review concluded that in about years, if certain conditions were met, the Government could make a decision in favour of a replacement reactor. A major milestone was achieved when, on 3 September 1997, the Australian Government announced the construction of a replacement research reactor at the site of Australia's existing research reactor HIFAR, subject to the satisfactory outcome of an environmental assessment process. The reactor will be have the dual purpose of providing a first class facility for neutron beam research as well as providing irradiation facilities for both medical isotope production and commercial irradiations. The project is scheduled for completion before the end of 2005. (author)

Brazing graphite to graphite

International Nuclear Information System (INIS)

Peterson, G.R.

1976-01-01

Graphite is joined to graphite by employing both fine molybdenum powder as the brazing material and an annealing step that together produce a virtually metal-free joint exhibiting properties similar to those found in the parent graphite. Molybdenum powder is placed between the faying surfaces of two graphite parts and melted to form molybdenum carbide. The joint area is thereafter subjected to an annealing operation which diffuses the carbide away from the joint and into the graphite parts. Graphite dissolved by the dispersed molybdenum carbide precipitates into the joint area, replacing the molybdenum carbide to provide a joint of graphite
International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)
Measurements of impurity migration in graphite at high temperatures using a proton microprobe

International Nuclear Information System (INIS)

Shroy, R.E.; Soo, P.; Sastre, C.A.; Schweiter, D.G.; Kraner, H.W.; Jones, K.W.

1978-01-01

The migration of fission products and other impurities through the graphite core of a High Temperature Gas Cooled Reactor is of prime importance in studies of reactor safety. Work in this area is being carried out in which graphite specimens are heated to temperatures up to 3800 0 C to induce migration of trace elements whose local concentrations are then measured with a proton microprobe. This instrument is a powerful device for such work because of its ability to determine concentrations at a part per million (ppm) level in a circular area as small as 10 μm while operating in an air environment. Studies show that Si, Ca, Cl, and Fe impurities in graphite migrate from hotter to cooler regions. Also Si, S, Cl, Ca, Fe, Mn, and Cr are observed to escape from the graphite and be deposited on cooler surfaces
New research possibilities at the Budapest research reactor

International Nuclear Information System (INIS)

Hargitai, T.; Vidovszky, I.

2001-01-01

The Budapest Research Reactor is the first nuclear facility of Hungary. It was commissioned in 1959, reconstructed and upgraded in 1967 and 1986-92. The main purpose of the reactor is to serve neutron research. The reactor was extended by a liquid hydrogen type cold neutron source in 2000. The research possibilities are much improved by the CNS both in neutron scattering and neutron activation. (author)
Decommissioning of the Dragon High Temperature Reactor (HTR) Located at the Former United Kingdom Atomic Energy Authority (UKAEA) Research Site at Winfrith - 13180

Energy Technology Data Exchange (ETDEWEB)

Smith, Anthony A. [Research Sites Restoration Ltd, Winfrith, Dorset (United Kingdom)

2013-07-01

The Dragon Reactor was constructed at the United Kingdom Atomic Energy Research Establishment at Winfrith in Dorset through the late 1950's and into the early 1960's. It was a High Temperature Gas Cooled Reactor (HTR) with helium gas coolant and graphite moderation. It operated as a fuel testing and demonstration reactor at up to 20 MW (Thermal) from 1964 until 1975, when international funding for this project was terminated. The fuel was removed from the core in 1976 and the reactor was put into Safestore. To meet the UK's Nuclear Decommissioning Authority (NDA) objective to 'drive hazard reduction' [1] it is necessary to decommission and remediate all the Research Sites Restoration Ltd (RSRL) facilities. This includes the Dragon Reactor where the activated core, pressure vessel and control rods and the contaminated primary circuit (including a {sup 90}Sr source) still remain. It is essential to remove these hazards at the appropriate time and return the area occupied by the reactor to a safe condition. (author)
Rules for design of nuclear graphite core components - some considerations and approaches

International Nuclear Information System (INIS)

Svalbonas, V.; Stilwell, T.C.; Zudans, Z.

1978-01-01

The use of graphite as a structural element presents unusual problems both for the designer and stress analysist. When the structure happens to be a nuclear reactor core, these problems are significantly magnified both by the environment and the attendant safety requirements. In the high temperature gas reactor (HTGR) core a large number of elements are constructed of nuclear graphite. This paper discusses the attendant difficulties, and presents some approaches, for ASME code safety-consistent design and analysis. The statistical scatter of material properties, which complicates even the definitions of allowable stress, as well as the brittle, anisotropic, inhomogeneous nature of the graphite was considered. The study of this subject was undertaken under contract to the U.S. Nuclear Regulatory Commission. (Auth.)
Criticality calculations for a critical assembly, graphite moderate, using 20% enriched uranium

International Nuclear Information System (INIS)

Almeida Ferreira, A.C. de; Hukai, R.Y.

1975-01-01

The construction of a Zero Power Reactor (ZPR) at the Instituto de Energia Atomica in order to measure the neutron characteristics (parameters) of HTGR reactors is proposed. The necessary quantity fissile uranium for these measurements has been calculed. Criticality studies of graphite moderated critical assemblies containing thorium have been made and the critical mass of each of several typical commercial HTGR compositions has been calculated using computer codes HAMMER and CITATION. Assemblies investigated contained a central cylindrical core region, simulating a typical commercial HTGR composition, a uranium-graphite driver region and a outer pure graphite reflector region. It is concluded that a 10Kg inventory of fissile uranium will be required for a program of measurements utilizing each of the several calculated assemblies
Melting of contaminated steel scrap from the dismantling of the CO2 systems of gas cooled, graphite moderated nuclear reactors

International Nuclear Information System (INIS)

Feaugas, J.; Jeanjacques, M.; Peulve, J.

1994-01-01

G2 and G3 are the natural Uranium cooled reactors Graphite/Gas. The two reactors were designed for both plutonium and electricity production (45 MWe). The dismantling of the reactors at stage 2 has produced more than 4 000 tonnes of contaminated scrap. Because of their large mass and low residual contamination level, the French Atomic Energy Commission (CEA) considered various possibilities for the processing of these metallic products in order to reduce the volume of waste going to be stored. After different studies and tests of several processes and the evaluation of their results, the choice to melt the dismantled pipeworks was taken. It was decided to build the Nuclear Steel Melting Facility known as INFANTE, in cooperation with a steelmaker (AHL). The realization time schedule for the INFANTE lasted 20 months. It included studies, construction and the licensing procedure. (authors). 2 tabs., 3 figs
Development of fracture toughness test method for nuclear grade graphite

Energy Technology Data Exchange (ETDEWEB)

Chang, C. H.; Lee, J. S.; Cho, H. C.; Kim, D. J.; Lee, D. J. [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2006-02-15

Because of its high strength and stability at very high temperature, as well as very low thermal neutron absorption cross-section, graphite has been widely used as a structural material in Gas Cooled Reactors (GCR). Recently, many countries are developing the Very High Temperature gas cooled Reactor (VHTR) because of the potentials of hydrogen production, as well as its safety and viable economics. In VHTR, helium gas serves as the primary coolant. Graphite will be used as a reflector, moderator and core structural materials. The life time of graphite is determined from dimensional changes due to neutron irradiation, which closely relates to the changes of crystal structure. The changes of both lattice parameter and crystallite size can be easily measured by X-ray diffraction method. However, due to high cost and long time of neutron irradiation test, ion irradiation test is being performed instead in KAERI. Therefore, it is essential to develop the technique for measurement of ion irradiation damage of nuclear graphite. Fracture toughness of nuclear grade graphite is one of the key properties in the design and development of VHTR. It is important not only to evaluate the various properties of candidate graphite but also to assess the integrity of nuclear grade graphite during operation. Although fracture toughness tests on graphite have been performed in many laboratories, there have been wide variations in values of the calculated fracture toughness, due to the differences in the geometry of specimens and test conditions. Hence, standard test method for nuclear graphite is required to obtain the reliable fracture toughness values. Crack growth behavior of nuclear grade graphite shows rising R-curve which means the increase in crack growth resistance as the crack length increases. Crack bridging and microcracking have been proposed to be the dominant mechanisms of rising R-curve behavior. In this paper, the technique to measure the changes of crystallite size and
High and very high temperature reactor research for multipurpose energy applications (RAPHAEL)

International Nuclear Information System (INIS)

Hittner, Dominique; Bogusch, Edgar; Fuetterer, Michal; De Groot, Sander; Ruer, Jacques

2010-01-01

The sections of the contribution are as follows: Achievements and future R and D needs in baseline technologies (Fuel and fuel cycle; Materials and components - reactor vessel, intermediate heat exchanger, graphite internals; Safety; Computer code qualification; Waste management; Education and training; International dimension of the European R and D); Towards application; and The path forward (P.A.)
Nondestructive testing on graphite structures for high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Ishihara, Masahiro; Kambe, Mamoru; Tsuji, Nobumasa.

1994-01-01

The application of ultrasonic (for internal defects) and eddy current testing (for surface defects) were investigated on the structures of nuclear-grade IG-110 and PGX graphite for the HTTR. The equipment were developed in order to detect the specific configuration of graphite blocks and the testing conditions were defined as the practical testing methods. The established testing methods are being used for the acceptance tests of graphite structures in the HTTR. (author)
Analysis of electrochemical disintegration process of graphite matrix

International Nuclear Information System (INIS)

Tian Lifang; Wen Mingfen; Chen Jing

2010-01-01

The electrochemical method with ammonium nitrate as electrolyte was studied to disintegrate the graphite matrix from the simulative fuel elements for high temperature gas-cooled reactor. The influences of process parameters, including salt concentration, system temperature and current density, on the disintegration rate of graphite fragments were investigated in the present work. The experimental results showed that the disintegration rate depended slightly on the temperature and salt concentration. The current density strongly affected the disintegration rate of graphite fragments. Furthermore, the content of introduced oxygen in final graphite fragments was independent of the current density and the concentration of electrolyte. Moreover, the structural evolution of graphite was analyzed based on the microstructural parameters determined by X-ray diffraction profile fitting analysis using MAUD (material analysis using diffraction) before and after the disintegration process. It may safely be concluded that the graphite disintegration can be ascribed to the influences of the intercalation of foreign molecules in between crystal planes and the partial oxidation involved. The disintegration process was described deeply composed of intercalate part and further oxidation part of carbon which effected together to lead to the collapse of graphite crystals.
Utilization of nuclear research reactors

International Nuclear Information System (INIS)

1980-01-01

Full text: Report on an IAEA interregional training course, Budapest, Hungary, 5-30 November 1979. The course was attended by 19 participants from 16 Member States. Among the 28 training courses which the International Atomic Energy Agency organized within its 1979 programme of technical assistance was the Interregional Training Course on the Utilization of Nuclear Research Reactors. This course was held at the Nuclear Training Reactor (a low-power pool-type reactor) of the Technical University, Budapest, Hungary, from 5 to 30 November 1979 and it was complemented by a one-week Study Tour to the Nuclear Research Centre in Rossendorf near Dresden, German Democratic Republic. The training course was very successful, with 19 participants attending from 16 Member States - Bangladesh, Bolivia, Czechoslovakia, Ecuador, Egypt, India, Iraq, Korean Democratic People's Republic, Morocco, Peru, Philippines, Spain, Thailand, Turkey, Vietnam and Yugoslavia. Selected invited lecturers were recruited from the USA and Finland, as well as local scientists from Hungarian institutions. During the past two decades or so, many research reactors have been put into operation around the world, and the demand for well qualified personnel to run and fully utilize these facilities has increased accordingly. Several developing countries have already acquired small- and medium-size research reactors mainly for isotope production, research in various fields, and training, while others are presently at different stages of planning and installation. Through different sources of information, such as requests to the IAEA for fellowship awards and experts, it became apparent that many research reactors and their associated facilities are not being utilized to their full potential in many of the developing countries. One reason for this is the lack of a sufficient number of trained professionals who are well acquainted with all the capabilities that a research reactor can offer, both in research and
Surface area-burnoff correlation for the steam--graphite reaction

International Nuclear Information System (INIS)

Stark, W.A. Jr.; Malinauskas, A.P.

1977-01-01

The oxidation of core graphite by steam of air represents a problem area of significant concern in safety analyses for the high temperature gas cooled reactor (HTGR). Core and core-support graphite integrity and strength deteriorate with oxidation of the graphite, and oxidation furthermore could affect the rate of fission product release under upset conditions. Consequently, modeling of core response during steam or air ingress conditions requires an expression for the rate of graphite interaction with those impurities. The steam--graphite reaction in particular is a complex interaction of mass transport within the graphite with chemi-sorption and reaction on accessible surfaces; experimental results from graphite to graphite are highly variable, and the description of the reaction is not yet completely consistent. A simple etch pit model relating surface area to burnoff has been proposed and shown to provide reasonable correlation with experimental data obtained from steam oxidation studies of nuclear grade H-327 graphite. Unaccounted differences between theory and experiment arise at burnoffs exceeding 3 to 5 percent. The model, while not complete nor comprehensive, is consistent with experimental observations of graphite oxidation by O 2 (air), CO 2 , or H 2 O, and could have some utility in safety analysis
High-temperature process heat reactor with solid coolant and radiant heat exchange

International Nuclear Information System (INIS)

Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

1984-01-01

The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion
Heat Transfer During Evaporation of Cesium From Graphite Surface in an Argon Environment

Directory of Open Access Journals (Sweden)

Bespala Evgeny

2016-01-01

Full Text Available The article focuses on discussion of problem of graphite radioactive waste formation and accumulation. It is shown that irradiated nuclear graphite being inalienable part of uranium-graphite reactor may contain fission and activation products. Much attention is given to the process of formation of radioactive cesium on the graphite element surface. It is described a process of plasma decontamination of irradiated graphite in inert argon atmosphere. Quasi-one mathematical model is offered, it describes heat transfer process in graphite-cesium-argon system. Article shows results of calculation of temperature field inside the unit cell. Authors determined the factors which influence on temperature change.
Porosity effects in the neutron total cross section of graphite

International Nuclear Information System (INIS)

Santisteban, J. R; Dawidowski, J; Petriw, S. N

2009-01-01

Graphite has been used in nuclear reactors since the birth of the nuclear industry due to its good performance as a neutron moderator material. Graphite is still an option as moderator for generation IV reactors due to its good mechanical and thermal properties at high operation temperatures. So, there has been renewed interest in a revision of the computer libraries used to describe the neutron cross section of graphite. For sub-thermal neutron energies, polycrystalline graphite shows a larger total cross section (between 4 and 8 barns) than predicted by existing theoretical models (0.2 barns). In order to investigate the origin of this discrepancy we measured the total cross section of graphite samples of three different origins, in the energy range from 0.001 eV to 10 eV. Different experimental arrangements and sample treatments were explored, to identify the effect of various experimental parameters on the total cross section measurement. The experiments showed that the increase in total cross section is due to neutrons scattered around the forward direction. We associate these small-angle scattered neutrons (SANS) to the porous structure of graphite, and formulate a very simple model to compute its contribution to the total cross section of the material. This results in an analytic expression that explicitly depends on the density and mean size of the pores, which can be easily incorporated in nuclear library codes. [es
Preliminary scoping study of some neutronic aspects of new shim safety rods for a typical 5 MW research reactor by Monte Carlo simulation

Energy Technology Data Exchange (ETDEWEB)

Shoushtari, M.K.; Kakavand, T. [Faculty of Science, University of Zanjan, P.O. BOX 1415, Zanjan (Iran, Islamic Republic of); Ghaforian, H. [Faculty of Science and Technology of Marine, P.O. BOX 212 Tehran (Iran, Islamic Republic of); Kiai, S.M. Sadat [Nuclear Science and Technology Research Institute (NSTR), Nuclear Science Research, A.E.O.I. P.O. BOX 14155-1339, Tehran (Iran, Islamic Republic of)], E-mail: sadatkiai@yahoo.com

2009-02-15

A Monte Carlo simulation of a typical 5 MW research reactor (TRR) was carried out using MCNP4C code. The geometry of the reactor core was modeled including the details of all fuel elements, control rods, all irradiation channels, graphite reflectors, reactor pool and thermal column. The model predicted neutron flux distributions within the core, control rod (CR) worth, core reactivity ({rho}), shutdown margin, and some kinetic parameters when the control rod insert or withdraw. This study was carried out to reduce blockage probability of shim safety rod (SSR)s of the TRR. Two introduced more blackness SSRs were chosen and made thinner in a way adequate blackness, in comparison to the present rods, achieved.
Mimic of OSU research reactor

International Nuclear Information System (INIS)

Lu, Hong; Miller, D.W.

1991-01-01

The Ohio State University research reactor (OSURR) is undergoing improvements in its research and educational capabilities. A computer-based digital data acquisition system, including a reactor system mimic, will be installed as part of these improvements. The system will monitor the reactor system parameters available to the reactor operator either in digital parameters available to the reactor operator either in digital or analog form. The system includes two computers. All the signals are sent to computer 1, which processes the data and sends the data through a serial port to computer 2 with a video graphics array VGA monitor, which is utilized to display the mimic system of the reactor
MWPC data acquisition in the Brookhaven FASTBUS

International Nuclear Information System (INIS)

Black, J.K.; Blatt, S.R.; Campbell, M.C.

1983-01-01

For Brookhaven AGS Experiment No. 749, a data acquisition system to accommodate 12, 256 wire multiwire proportional chambers (MWPCs) has been built in the context of the Brookhaven FASTBUS. Information about wires hit or continuous clusters of wires hit is encoded as the centroids of the clusters and number of wires in those clusters. The encoded information is stored in a stack memory in a FASTBUS module where it can be accessed by a FASTBUS Master. Encoding time is less than 4 microseconds. Also, information (in the form of front panel outputs) as to the nature of the data is available in less than 200 nanoseconds

Computational prediction of dust production in graphite moderated pebble bed reactors

Science.gov (United States)

Rostamian, Maziar

The scope of the work reported here, which is the computational study of graphite wear behavior, supports the Nuclear Engineering University Programs project "Experimental Study and Computational Simulations of Key Pebble Bed Thermomechanics Issues for Design and Safety" funded by the US Department of Energy. In this work, modeling and simulating the contact mechanics, as anticipated in a PBR configuration, is carried out for the purpose of assessing the amount of dust generated during a full power operation year of a PBR. A methodology that encompasses finite element analysis (FEA) and micromechanics of wear is developed to address the issue of dust production and its quantification. Particularly, the phenomenon of wear and change of its rate with sliding length is the main focus of this dissertation. This work studies the wear properties of graphite by simulating pebble motion and interactions of a specific type of nuclear grade graphite, IG-11. This study consists of two perspectives: macroscale stress analysis and microscale analysis of wear mechanisms. The first is a set of FEA simulations considering pebble-pebble frictional contact. In these simulations, the mass of generated graphite particulates due to frictional contact is calculated by incorporating FEA results into Archard's equation, which is a linear correlation between wear mass and wear length. However, the experimental data by Johnson, University of Idaho, revealed that the wear rate of graphite decreases with sliding length. This is because the surfaces of the graphite pebbles become smoother over time, which results in a gradual decrease in wear rate. In order to address the change in wear rate, a more detailed analysis of wear mechanisms at room temperature is presented. In this microscale study, the wear behavior of graphite at the asperity level is studied by simulating the contact between asperities of facing surfaces. By introducing the effect of asperity removal on wear rate, a nonlinear
The effect of oxidizing atmosphere on strength loss in HTGR graphites

International Nuclear Information System (INIS)

Heiser, J.H.; Finfrock, C.C.; Lees, B.S.

1983-01-01

This paper reports on studies involving various reactor grade graphites and the possible mechanisms leading to strength loss differences. Compressive and tensile specimens of six reactor grade graphites were oxidized. The compressive or tensile strengths were then determined using a Timus-Olsen Universal testing machine following ASTM standard test specifications. Two possible mechanisms are proposed to explain the differences in strength loss given the same mass loss but different oxidants. One mechanism has the impurity iron located primarily in the filler particles and the second mechanism arranges the iron either uniformly throughout the binder or inhomogeneously dispersed in large pockets in the binder
Barrier Cavities in the Brookhaven AGS

International Nuclear Information System (INIS)

Blaskiewicz, M.; Brennan, J. M.; Roser, T.; Smith, K.; Spitz, R.; Zaltsman, A.; Fujieda, M.; Iwashita, Y.; Noda, A.; Yoshii, M.; Mori, Y.; Ohmori, C.; Sato, Y.

1999-01-01

In collaboration with KEK two barrier cavities, each generating 40 kV per turn have been installed in the Brookhaven AGS. Machine studies are described and their implications for high intensity operations are discussed
Measurement of graphite and aluminium absorption cross sections via reactor period by danger coefficient method; Merenje apsorpcionih preseka grafita i aluminijuma preko periode reaktora metodom koeficijenta opasnosti

Energy Technology Data Exchange (ETDEWEB)

Petrovic, M; Markovic, V; Velickovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

1963-07-01

Full text: This activity is a logical continuation of the experiment at the RA reactor during 1962 which was based on compensating the effect by means of control rod. Since results are given with significant errors, new method for measuring the absorption cross sections via reactor period. Experiment was done at the RB reactor which was particularly prepared for this type of experiments. Reactor power was from 50 mW to 2 W. Absorption cross sections were measured for two types of material: domestic graphite No.3 and French graphite 'Pachiney', and two types of aluminium. Total errors in applying this method are {+-} 5%, where the source of major part of error comes from uncertainty of the standard absorption power (previous method gave {+-} 10 do 55% ). Comparison of French graphite absorption cross section obtained via reactor period and via control rod showed approximate agreement with discrepancy of 5.4% which is considered within the precision of this method. Considering the accuracy of measurement results and reactor economy it is concluded that measuring absorption cross sections of samples via period of RB reactor is more favourable than measurements by control rod at the RA reactor. Pun tekst: Ovaj rad predstavlja logican nastavak eksperimenta na reaktoru RA u toku 1962. godine, koji je bazirao na kompenzaciji efekta pomocu kontrolne sipke. Kako su rezultati dati sa velikim greskama, to se prislo novom nacinu merenja apsorpsionih preseka preko periode reaktora. Eksperiment je radjen na reaktoru RB koji je specijalno pripremljen za ovu vrstu eksperimenta. Snaga reaktora se kretala od 50 mW do 2 W. Preko periode reaktora RB odredjeni su apsorpcioni preseci za dve vrste materijala i to: domaci grafit No.3 i francuski 'Pachiney', i dve vrste aluminijuma. Ukupne greske pri ovom nacimu merenja iznose oko {+-} 5%, gde glavni deo greske nosi neodredjenost apsorpcione moci standarda (ranija metoda je dala {+-} 10 do 55% ). Poredjenjem vrednosti apsorpcionih preseka
Nuclear Capacity Building through Research Reactors

International Nuclear Information System (INIS)

2017-01-01

Four Instruments: •The IAEA has recently developed a specific scheme of services for Nuclear Capacity Building in support of the Member States cooperating research reactors (RR) willing to use RRs as a primary facility to develop nuclear competences as a supporting step to embark into a national nuclear programme. •The scheme is composed of four complementary instruments, each of them being targeted to specific objective and audience: Distance Training: Internet Reactor Laboratory (IRL); Basic Training: Regional Research Reactor Schools; Intermediate Training: East European Research Reactor Initiative (EERRI); Group Fellowship Course Advanced Training: International Centres based on Research Reactors (ICERR)
Utilizing the slowing-down-time technique for benchmarking neutron thermalization in graphite

International Nuclear Information System (INIS)

Zhou, T.; Hawari, A. I.; Wehring, B. W.

2007-01-01

Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in 'reactor grade' graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70 cm x 70 cm x 70 cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multichannel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured
Use of thorium for high temperature gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

Guimarães, Cláudio Q., E-mail: claudio_guimaraes@usp.br [Universidade de São Paulo (USP), SP (Brazil). Instituto de Física; Stefani, Giovanni L. de, E-mail: giovanni.stefani@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil); Santos, Thiago A. dos, E-mail: thiago.santos@ufabc.edu.br [Universidade Federal do ABC (UFABC), Santo André, SP (Brazil)

2017-07-01

The HTGR ( High Temperature Gas-cooled Reactor) is a 4{sup th} generation nuclear reactor and is fuelled by a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type: The German “pebble bed” mode, which is a system that uses spherical fuel elements, containing a graphite-and-fuel mixture coated in a graphite shell; and the American version, whose fuel is loaded into precisely located graphite hexagonal prisms that interlock to create the core of the vessel. In both variants, the coolant consists of helium pressurised. The HTGR system operates most efficiently with the thorium fuel cycle, however, so relatively little development has been carried out in this country on that cycle for HTGRs. In the Nuclear Engineering Centre of IPEN (Instituto de Pesquisas Energéticas e Nucleares), a study group is being formed linked to thorium reactors, whose proposal is to investigate reactors using thorium for {sup 233}U production and rejects burning. The present work intends to show the use of thorium in HTGRs, their advantages and disadvantages and its feasibility. (author)
Wear Behavior of Selected Nuclear Grade Graphites at Room Temperature in Ambient Air Environment

International Nuclear Information System (INIS)

Kim, Eung-Seon; Park, Kwang-Seok; Kim, Yong-Wan

2008-01-01

In a very high temperature reactor (VHTR), graphite will be used not only for as a moderator and reflector but also as a major structural component due to its excellent neutronic, thermal and mechanical properties. In the VHTR, wear of graphite components is inevitable due to a neutron irradiation-induced dimensional change, thermal gradient, relative motions of graphite components and a shock load such as an earthquake. Large wear particles accumulated at the bottom of a reactor can influence the cooling of the lower part and small wear particles accumulated on the primary circuit and heat exchanger tube can make it difficult to inspect the equipment, and also decrease the heat exchange rate. In the present work, preliminary wear tests were performed at room temperature in ambient air environment to understand the basic wear characteristics of selected nuclear grade graphites for the VHTR
Upgrade of the controls for the Brookhaven linac

International Nuclear Information System (INIS)

Buxton, W.E.

1995-01-01

The control of the magnets, rf system, and other components at the Brookhaven Linac uses a system that was developed at Brookhaven in the late 1960's. This system will be retired in the summer of 1995. The Linac controls are being upgraded using modem VME-based hardware compatible with RHIC generation controls, and an existing serial field bus. The timing for the Linac will also be upgraded and will use components developed for RHIC. The controls in general, the timing for the Linac, and the modules developed will be described
Biaxial testing for nuclear grade graphite by ball on three balls assessment

International Nuclear Information System (INIS)

Mohd Reusmaazran Yusof; Yusof Abdullah

2012-01-01

Nuclear grade (high-purity) graphite for fuel element and moderator material in Advanced Gas Cooling Reactors (AGR) displays large scatter in strength and a non-linear stress-strain response from the damage accumulation. These responses can be characterized as quasi-brittle behaviour. Current assessments of fracture in core graphite components are based on the linear elastic approximation and thus represent a major assumption. The quasi-brittle behaviour gives challenge to assess the real nuclear graphite component. The selected test method would help to bridge the gap between microscale to macro-scale in real reactor component. The small scale tests presented here can contribute some statistical data to manifests the failure in real component. The evaluation and choice of different solution design of biaxial test will be discussed in this paper. The ball on-three ball test method was used for assessment test follows by numerous of analytical method. The results shown that biaxial strength of the EY9 grade graphite depends on the method used for evaluation. Some of the analytical methods use to calculate biaxial strength were found not to be valid and therefore should not be used to assess the mechanical properties of nuclear graphite. (author)
TRISTAN-II at Brookhaven

International Nuclear Information System (INIS)

Chrien, R.E.; Stelts, M.L.; Manzella, V.; Gill, R.; Wohn, F.; Hill, J.C.

1979-01-01

The mass separator TRISTAN has been moved to Brookhaven HFBR. A flux of up to 10 11 neutrons/cm 2 /sec will be available on target, and improvements in the ion source and beam optics have been made. A computer system with a CAMAC interface allows data collection from eight independent sources, while a PDP-11/34 computer is available for extensive off-line analysis
Development of integrated waste management options for irradiated graphite

Energy Technology Data Exchange (ETDEWEB)

Wareing, Alan; Abrahamsen-Mills, Liam; Fowler, Linda; Jarvis, Richard; Banford, Anthony William [National Nuclear Laboratory, Warrington (United Kingdom); Grave, Michael [Doosan Babcock, Gateshead (United Kingdom); Metcalfe, Martin [National Nuclear Laboratory, Gloucestershire (United Kingdom); Norris, Simon [Radioactive Waste Management Limited, Oxon (United Kingdom)

2017-08-15

The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.
Fort St. Vrain graphite site mechanical separation concept selection

International Nuclear Information System (INIS)

Berry, S.M.

1993-09-01

One of the alternatives to the disposal of the Fort St. Vrain (FSV) reactor spent nuclear fuel involves the separation of the fuel rods composed of compacts from the graphite fuel block assembly. After the separation of these two components, the empty graphite fuel blocks would be disposed of as a low level waste (provided the appropriate requirements are met) and the fuel compacts would be treated as high level waste material. This report deals with the mechanical separation aspects concerning physical disassembly of the FSV graphite fuel element into the empty graphite fuel blocks and fuel compacts. This report recommends that a drilling technique is the preferred choice for accessing the, fuel channel holes and that each hole is drilled separately. This report does not cover any techniques or methods to separate the triso fuel particles from the graphite matrix of the fuel compacts
Safety of research reactors (Design and Operation)

International Nuclear Information System (INIS)

Dirar, H. M.

2012-06-01

The primary objective of this thesis is to conduct a comprehensive up-to-date literature review on the current status of safety of research reactor both in design and operation providing the future trends in safety of research reactors. Data and technical information of variety selected historical research reactors were thoroughly reviewed and evaluated, furthermore illustrations of the material of fuel, control rods, shielding, moderators and coolants used were discussed. Insight study of some historical research reactors was carried with considering sample cases such as Chicago Pile-1, F-1 reactor, Chalk River Laboratories,. The National Research Experimental Reactor and others. The current status of research reactors and their geographical distribution, reactor category and utilization is also covered. Examples of some recent advanced reactors were studied like safety barriers of HANARO of Korea including safety doors of the hall and building entrance and finger print identification which prevent the reactor from sabotage. On the basis of the results of this research, it is apparent that a high quality of safety of nuclear reactors can be attained by achieving enough robust construction, designing components of high levels of efficiency, replacing the compounds of the reactor in order to avoid corrosion and degradation with age, coupled with experienced scientists and technical staffs to operate nuclear research facilities.(Author)
Commissioning of research reactors. Safety guide

International Nuclear Information System (INIS)

2006-01-01

The objective of this Safety Guide is to provide recommendations on meeting the requirements for the commissioning of research reactors on the basis of international best practices. Specifically, it provides recommendations on fulfilling the requirements established in paras 6.44 and 7.42-7.50 of International Atomic Energy Agency, Safety of Research Reactors, IAEA Safety Standards Series No. NS-R-4, IAEA, Vienna (2005) and guidance and specific and consequential recommendations relating to the recommendations presented in paras 615-621 of International Atomic Energy Agency, Safety in the Utilization and Modification of Research Reactors, Safety Series No. 35-G2, IAEA, Vienna (1994) and paras 228-229 of International Atomic Energy Agency, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, Safety Series No. 35-G1, IAEA, Vienna (1994). This Safety Guide is intended for use by all organizations involved in commissioning for a research reactor, including the operating organization, the regulatory body and other organizations involved in the research reactor project
Development of visual inspection technology for HTTR core support graphite structure

International Nuclear Information System (INIS)

Maruyama, So; Iyoku, Tatsuo; Inagaki, Yoshiyuki; Shiozawa, Shusaku; Masuma, Yoshitaka; Miki, Toshiya.

1996-01-01

The Japan Atomic Energy Research Institute is now constructing the High Temperature Engineering Test Reactor (HTTR), which employs a visual inspection of core support graphite structure, as an inservice inspection (ISI). In this inspection, TV camera will be used to investigate the alignment and integrity of the structure. Therefore, the ISI system, a combination of radiation tolerant TV camera and graphic processing system, is developed and examined its detectability and viewing angles using a simulated hot plenum of HTTR, which has artificial defects. As a result of a series of tests, it was confirmed that this system satisfied the requirements and was quite applicable for the ISI system of HTTR core support graphite structure. In addition, further improvement of the system, like a remote control procedure, will be investigated. (author)
IAEA activities on research reactor safety

International Nuclear Information System (INIS)

Alcala-Ruiz, F.

1995-01-01

Since its inception in 1957, the International Atomic Energy Agency (IAEA) has included activities in its programme to address aspects of research reactors such as safety, utilization and fuel cycle considerations. These activities were based on statutory functions and responsibilities, and on the current situation of research reactors in operation around the world; they responded to IAEA Member States' general or specific demands. At present, the IAEA activities on research reactors cover the above aspects and respond to specific and current issues, amongst which safety-related are of major concern to Member States. The present IAEA Research Reactor Safety Programme (RRSP) is a response to the current situation of about 300 research reactors in operation in 59 countries around the world. (orig.)
AGC-2 Graphite Preirradiation Data Analysis Report

Energy Technology Data Exchange (ETDEWEB)

William Windes; W. David Swank; David Rohrbaugh; Joseph Lord

2013-08-01

This report described the specimen loading order and documents all pre-irradiation examination material property measurement data for the graphite specimens contained within the second Advanced Graphite Capsule (AGC-2) irradiation capsule. The AGC-2 capsule is the second in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. The AGC test series is used to irradiate graphite specimens allowing quantitative data necessary for predicting the irradiation behavior and operating performance of new nuclear graphite grades to be generated which will ascertain the in-service behavior of the graphite for pebble bed and prismatic Very High Temperature Reactor (VHTR) designs. Similar to the AGC-1 specimen pre-irradiation examination report, material property tests were conducted on specimens from 18 nuclear graphite types but on an increased number of specimens (512) prior to loading into the AGC-2 irradiation assembly. All AGC-2 specimen testing was conducted at Idaho National Laboratory (INL) from October 2009 to August 2010. This report also details the specimen loading methodology for the graphite specimens inside the AGC-2 irradiation capsule. The AGC-2 capsule design requires “matched pair” creep specimens that have similar dose levels above and below the neutron flux profile mid-plane to provide similar specimens with and without an applied load. This document utilized the neutron flux profile calculated for the AGC-2 capsule design, the capsule dimensions, and the size (length) of the selected graphite and silicon carbide samples to create a stacking order that can produce “matched pairs” of graphite samples above and below the AGC-2 capsule elevation mid-point to provide specimens with similar neutron dose levels.
On disruption of reactor core of the Chernobylsk-4 reactor (retrospective analysis of experiments and facts)

International Nuclear Information System (INIS)

Platonov, P.A.

2007-01-01

Fragments of graphite blocks from the damaged Chernobyl NPP, unit 4 are studied, the results are analyzed. The temperature of the graphite blocks at the moment of accident release from the reactor is evaluated. Results of studying the fragments of fuel channel and fuel dispersion are considered. The fuel heat content at the moment of the explosion is evaluated and some conclusions are made about the character of the reactor core destruction [ru
Decontamination of nuclear graphite by thermal processing; Dekontamination von Nukleargraphit durch thermische Behandlung

Energy Technology Data Exchange (ETDEWEB)

Florjan, Monika W.

2010-04-15

The main problem in view of the direct disposal of the nuclear graphite is its large volume. This waste contains long-lived and short-lived radionuclides which determine the waste strategy. The irradiated graphite possess high amount of the {sup 14}C isotope. The main object of the present work was the selective separation of {sup 14}C isotope from the isotope {sup 12}C by thermal treatment (pyrolysis, partial oxidation). A successful separation could reduce the radiotoxicity and offer a different disposal strategy. Three different graphite types were investigated. The samples originate from the reflector and from the flaking of spherical fuel elements of the high-temperature reactor (AVR) Juelich. The samples from the thermal column of the research reactor (Merlin, Juelich) were also investigated. The maximum tritium releases were obtained both in inert gas atmosphere (N{sub 2}) and under water vapour-oxidizing conditions at 1280 C and 900 C. Furthermore it could be shown that 28% of {sup 14}C could be released under inert gas conditions at a 1280 C. By additive of oxidizing agent such as water vapour and oxygen the {sup 14}C release could be increased. Under water vapour-oxidizing conditions at a temperature of 1280 C up to 93% of the {sup 14}C was separated from the graphite. The matrix corrosion of 5.4% was obtained. The selective separation of the {sup 14}C is possible, because a substantial part of the radiocarbon is bound near the grain boundary surfaces. (orig.)

Characterization of un-irradiated and irradiated reactor graphite; Karakterizacija neozracenog i ozracenog reaktorskog grafita

Energy Technology Data Exchange (ETDEWEB)

Marinkovic, S [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1965-11-15

This report contains three parts: characterization of Yugoslav nuclear graphite development of methods and obtained results, characterization of un-irradiated and irradiated domestic nuclear graphite; calculation of electrical conductivity changes due to vacancies in the graphite crystal lattice.
Design of a multipurpose research reactor

International Nuclear Information System (INIS)

Sanchez Rios, A.A.

1990-01-01

The availability of a research reactor is essential in any endeavor to improve the execution of a nuclear programme, since it is a very versatile tool which can make a decisive contribution to a country's scientific and technological development. Because of their design, however, many existing research reactors are poorly adapted to certain uses. In some nuclear research centres, especially in the advanced countries, changes have been made in the original designs or new research prototypes have been designed for specific purposes. These modifications have proven very costly and therefore beyond the reach of developing countries. For this reason, what the research institutes in such countries need is a single sufficiently versatile nuclear plant capable of meeting the requirements of a nuclear research programme at a reasonable cost. This is precisely what a multipurpose reactor does. The Mexican National Nuclear Research Institute (ININ) plans to design and build a multipurpose research reactor capable at the same time of being used for the development of reactor design skills and for testing nuclear materials and fuels, for radioisotopes production, for nuclear power studies and basic scientific research, for specialized training, and so on. For this design work on the ININ Multipurpose Research Reactor, collaborative relations have been established with various international organizations possessing experience in nuclear reactor design: Atomehnergoeksport of the USSR: Atomic Energy of Canada Limited (AECL); General Atomics (GA) of the USA; and Japan Atomic Energy Research Institute
Differences in the irradiation effects of IG-110 and IG-430 nuclear graphites : effects of coke difference

International Nuclear Information System (INIS)

Chi, Se Hwan; Kim, Gen Chan; Kim, Eung Seon; Hong, Jin Ki; Chang, Jong Hwa

2005-01-01

In the high temperature gas cooled reactors (HTGRs), graphite acts as a moderator and reflector as well as a major structural component that may provide channels for the fuel and coolant gas, channels for control and shut down, and thermal and neutron shielding. During a reactor operation, many of the physical, chemical and mechanical properties of these graphite components are significantly modified as a function of the temperature, environment, and an irradiation. On the other hand, currently, all the nuclear graphites are being manufactured from two types of cokes, i.e., petroleum and coal-tar pitch coke, and it has been understood that the type of coke plays the most critical role determining the properties of a specific graphite grade. To investigate the effects of coke types on the irradiation response of a graphite, two graphites of different cokes were irradiated by 3 MeV C+ ions and the differences in the response of ion-irradiation were investigated
Graphite-based detectors of alkali metals for nuclear power plants

International Nuclear Information System (INIS)

Kalandarishvili, A.G.; Kuchukhidze, V.A.; Sordiya, T.D.; Shartava, Sh.Sh.; Stepennov, B.S.

1993-01-01

The coolants most commonly used in today's fast reactors are alkali metals or their alloys. A major problem in nuclear plant design is leakproofing of the liquid-metal cooling system, and many leak detection methods and safety specifications have been developed as a result. Whatever the safety standards adopted for nuclear plants in different countries, they all rely on the basic fact that control of the contamination and radiation hazards involved requires reliable monitoring equipment. Results are presented of trials with some leak detectors for the alkali-metal circuits of nuclear reactors. The principal component affecting the detector performance is the sensing element. In the detectors graphite was employed, whose laminar structure enables it to absorb efficiently alkali-metal vapors at high temperatures (320--500 K). This produces a continuous series of alkali-metal-graphite solid solutions with distinct electrical, thermal, and other physical properties. The principle of operation of the detectors resides in the characteristic reactions of the metal-graphite system. One detector type uses the change of electrical conductivity of the graphite-film sensor when it is exposed to alkali-metal vapor. In order to minimize the effect of temperature on the resistance the authors prepared composite layers of graphite intercalated with a donor impurity (cesium or barium), and a graphite-nickel material. The addition of a small percentage of cesium, barium, or nickel produces a material whose temperature coefficient of resistance is nearly zero. Used as a sensing element, such a material can eliminate the need for thermostatic control of the detector
The IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Abou Yehia, H.

2007-01-01

According to the research reactor database of IAEA (RRDB), 250 reactors are operating worldwide, 248 have been shut down and 170 have been decommissioned. Among the 248 reactors that do not run, some will resume their activities, others will be dismantled and the rest do not face a clear future. The analysis of reported incidents shows that the ageing process is a major cause of failures, more than two thirds of operating reactors are over 30 years old. It also appears that the lack of adequate regulations or safety standards for research reactors is an important issue concerning reactor safety particularly when reactors are facing re-starting or upgrading or modifications. The IAEA has launched a 4-axis program: 1) to set basic safety regulations and standards for research reactors, 2) to provide IAEA members with an efficient help for the application of these safety regulations to their reactors, 3) to foster international exchange of information on research reactor safety, and 4) to provide IAEA members with a help concerning safety issues linked to malicious acts or sabotage on research reactors
Improving nuclear safety at international research reactors: The Integrated Research Reactor Safety Enhancement Program (IRRSEP)

International Nuclear Information System (INIS)

Huizenga, David; Newton, Douglas; Connery, Joyce

2002-01-01

Nuclear energy continues to play a major role in the world's energy economy. Research and test reactors are an important component of a nation's nuclear power infrastructure as they provide training, experiments and operating experience vital to developing and sustaining the industry. Indeed, nations with aspirations for nuclear power development usually begin their programs with a research reactor program. Research reactors also are vital to international science and technology development. It is important to keep them safe from both accident and sabotage, not only because of our obligation to prevent human and environmental consequence but also to prevent corresponding damage to science and industry. For example, an incident at a research reactor could cause a political and public backlash that would do irreparable harm to national nuclear programs. Following the accidents at Three Mile Island and Chernobyl, considerable efforts and resources were committed to improving the safety posture of the world's nuclear power plants. Unsafe operation of research reactors will have an amplifying effect throughout a country or region's entire nuclear programs due to political, economic and nuclear infrastructure consequences. (author)
IAEA programme on research reactor safety

International Nuclear Information System (INIS)

Alcala, F.; Di Meglio, A.F.

1995-01-01

This paper describes the IAEA programme on research reactor safety and includes the safety related areas of conversions to the use of low enriched uranium (LEU) fuel. The program is based on the IAEA statutory responsibilities as they apply to the requirements of over 320 research reactors operating around the world. The programme covers four major areas: (a) the development of safety documents; (b) safety missions to research reactor facilities; (c) support of research programmes on research reactor safety; (d) support of Technical Cooperation projects on research reactor safety issues. The demand for these activities by the IAEA member states has increased substantially in recent years especially in developing countries with increasing emphasis being placed on LEU conversion matters. In response to this demand, the IAEA has undertaken an extensive programme for each of the four areas above. (author)
Brookhaven: Ready for RHIC

International Nuclear Information System (INIS)

Ludlam, Tom

1990-01-01

With its RHIC - Relativistic Heavy Ion Collider - project now part of the budget proposed by US President Bush for fiscal year 1991, Brookhaven is about to start construction of a unique kind of high energy collider. At a time when accelerators handling particles - electrons, protons and their antimatter counterparts - are boosting beam energies for microscopes to probe evershorter distances, RHIC is envisioned as a great pressure-cooker for strongly interacting matter
Brookhaven: Ready for RHIC

Energy Technology Data Exchange (ETDEWEB)

Ludlam, Tom

1990-04-15

With its RHIC - Relativistic Heavy Ion Collider - project now part of the budget proposed by US President Bush for fiscal year 1991, Brookhaven is about to start construction of a unique kind of high energy collider. At a time when accelerators handling particles - electrons, protons and their antimatter counterparts - are boosting beam energies for microscopes to probe evershorter distances, RHIC is envisioned as a great pressure-cooker for strongly interacting matter.
Measurements of anomalous neutron transport in bulk graphite

International Nuclear Information System (INIS)

Bowman, C.D.; Smith, G.A.; Vogelaar, B.; Howell, C.R.; Bilpuch, E.G.; Tornow, W.

2003-01-01

The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)
Technical development of graphite waste treatment in NUPEC

International Nuclear Information System (INIS)

Saishu, S.; Inoue, T.

2001-01-01

In Japan, Tokai Power Station, which is a Gas Cooled Reactor and uses graphite as moderator, ceased operation at the end of March in 1998 and it is planned to transfer to decommissioning stage. In this decommissioning stage it is very important to be able to treat and dispose the graphite waste in order to carry out the decommissioning safely and economically. NUPEC has been developing the graphite treatment and disposal technology since 1997 and we introduce the outline of the technical development. For the technology on high density packing into disposal container, the high density packing method and the assessment method on nuclide leaching characteristics were developed, and the cementing test for graphite powder by using Tokai spare graphite was performed and the hydrophobic characteristics between graphite and cement was grasped and the accelerator candidature for affinity was selected. From the view point of economical treatment, the incinerating technology was selected as candidature, and the methods for incinerating graphite and treating off gas are developed. The method of collecting C-14 in off gas was selected for reducing the off gas radiation level. The applicability of actual graphite treatment technology was considered from the view point of safety, economics and preparation of technical standard; the technical theme appeared, the developing planning items were established, and the detailed and actual scale tests will be carried out according to the planning. (author)
Measurements of anomalous neutron transport in bulk graphite

Energy Technology Data Exchange (ETDEWEB)

Bowman, C.D.; Smith, G.A. [ADNA Corp., Los Alamos, NM (United States); Vogelaar, B. [Virginia Tech., Blacksburg, VA (United States); Howell, C.R.; Bilpuch, E.G.; Tornow, W. [Triangle Univ. Nuclear Lab., Duke Univ., Durham, NC (United States)

2003-07-01

The neutron absorption of bulk granular graphite has been measured in a classical exponential diffusion experiment. Our first measurements of April 2002 implementing both exponential decay and pulsed die-away experiments and using the TUNL pulsed accelerator at Duke University as a neutron source indicated a capture cross section for graphite a striking factor of three lower than the measured value for carbon of 3.4 millibarns. Therefore a new exponential experiment with an improved geometry enabling greater accuracy has been performed giving an apparent cross section for carbon in the form of bulk granular graphite of less than 0.5 millibarns. This result confirms our first result and is also consistent with less than one part per million of boron in our graphite. The bulk density of the graphite is 1.02 compared with the actual particle density of 1.60 indicating a packing fraction of 0.64 or a void fraction of 0.36. We suspect that the apparent suppression of absorption in bulk graphite may be associated with the strong coherent diffraction of neutrons that dominates neutron transport in graphite. Coherent diffraction has never been taken into account in graphite reactor design and no neutron transport code including general use codes such as MCNP incorporate diffraction effects even though diffraction dominates many practical thermal neutron transport problems. (orig.)
Safety analysis for research reactors

International Nuclear Information System (INIS)

2008-01-01

The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of
Nuclear data usage for research reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Soyama, Kazuhiko; Amano, Toshio

1996-01-01

In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)
Strengthening IAEA Safeguards for Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-09-01

During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To
Design Procedure of Graphite Components by ASME HTR Codes

Energy Technology Data Exchange (ETDEWEB)

Kang, Ji-Ho; Jo, Chang Keun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet.
Design Procedure of Graphite Components by ASME HTR Codes

International Nuclear Information System (INIS)

Kang, Ji-Ho; Jo, Chang Keun

2016-01-01

In this study, the ASME B and PV Code, Subsection HH, Subpart A, design procedure for graphite components of HTRs was reviewed and the differences from metal materials were remarked. The Korean VHTR has a prismatic core which is made of multiple graphite blocks, reflectors, and core supports. One of the design issues is the assessment of the structural integrity of the graphite components because the graphite is brittle and shows quite different behaviors from metals in high temperature environment. The American Society of Mechanical Engineers (ASME) issued the latest edition of the code for the high temperature reactors (HTR) in 2015. In this study, the ASME B and PV Code, Subsection HH, Subpart A, Graphite Materials was reviewed and the special features were remarked. Due the brittleness of graphites, the damage-tolerant design procedures different from the conventional metals were adopted based on semi-probabilistic approaches. The unique additional classification, SRC, is allotted to the graphite components and the full 3-D FEM or equivalent stress analysis method is required. In specific conditions, the oxidation and viscoelasticity analysis of material are required. The fatigue damage rule has not been established yet
Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition

International Nuclear Information System (INIS)

Sharma, Uttam; Chauhan, Sachin S; Sharma, Jayshree; Sanyasi, A K; Ghosh, J; Choudhary, K K; Ghosh, S K

2016-01-01

The tokamak concept is the frontrunner for achieving controlled thermonuclear reaction on earth, an environment friendly way to solve future energy crisis. Although much progress has been made in controlling the heated fusion plasmas (temperature ∼ 150 million degrees) in tokamaks, technological issues related to plasma wall interaction topic still need focused attention. In future, reactor grade tokamak operational scenarios, the reactor wall and target plates are expected to experience a heat load of 10 MW/m 2 and even more during the unfortunate events of ELM's and disruptions. Tungsten remains a suitable choice for the wall and target plates. It can withstand high temperatures, its ductile to brittle temperature is fairly low and it has low sputtering yield and low fuel retention capabilities. However, it is difficult to machine tungsten and hence usages of tungsten coated surfaces are mostly desirable. To produce tungsten coated graphite tiles for the above-mentioned purpose, a coating reactor has been designed, developed and made operational at the SVITS, Indore. Tungsten coating on graphite has been attempted and successfully carried out by using radio frequency induced plasma enhanced chemical vapour deposition (rf -PECVD) for the first time in India. Tungsten hexa-fluoride has been used as a pre-cursor gas. Energy Dispersive X-ray spectroscopy (EDS) clearly showed the presence of tungsten coating on the graphite samples. This paper presents the details of successful operation and achievement of tungsten coating in the reactor at SVITS. (paper)
European Research Reactor Conference (RRFM) 2015: Conference Proceedings

International Nuclear Information System (INIS)

2015-01-01

In 2015 the European Research Reactor Conference, RRFM, took place in Bucharest, Romania. The conference programme resolved around a series of plenary sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions focused on all areas of the fuel cycle of research reactors, their utilisation, operation and management as well as new research reactor projects and Innovative methods in reactor physics and thermo-hydraulics. The European Research Reactor Conference also gave special attention to safety and security of research reactors
European Research Reactor Conference (RRFM) 2016: Conference Proceedings

International Nuclear Information System (INIS)

2016-01-01

The 2016 European Research Reactor Conference, RRFM, took place in Berlin, Germany. The conference programme resolved around a series of plenary sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions focused on all areas of the fuel cycle of research reactors, their utilisation, operation and management as well as new research reactor projects and Innovative methods in reactor physics and thermo-hydraulics. The European Research Reactor Conference also gave special attention to safety and security of research reactors.

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